NUREG/CR-6654 

INEEL/EXT-98-00383

A Study of Air-Operated 

Valves in U.S. Nuclear 

Power Plants 

Idaho National Engineering and Environmental Laboratory 

Bechtel BWXT Idaho, LLC

U.S. Nuclear Regulatory Commission 

Office of Nuclear Regulatory Research 

Washington, DC 20555-0001

Gp,JA R#EG'u 

do4 eoIEJ'
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NUREG/CR-6654 

INEEL/IEXT-98-00383 

A Study of Air-Operated 

Valves in U.S. Nuclear 

Power Plants 

Manuscript Completed: December 1999 

Date Published: February 2000 

Prepared by 

0. Rothberg, S. Khericha, J. Watkins, M. Holbrook 

Idaho National Engineering and Environmental Laboratory 

Bechtel BWXT Idaho, LLC 

Idaho Falls, ID 83415-3129 

H. Ornstein, NRC Project Manager 

Prepared for 

Division of Systems Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

NRC Job Code E8238
ABSTRACT 

A study of air-operated valves in nuclear power plant applications was con

ducted for the NRC Office of Research (the project was initiated by 

NRC/AEOD). The results of the study were based on visits to seven nuclear 

power plant sites, literature studies, and examinations of event records in data

bases available to the NRC. The purpose is to provide information to the NRC 

staff concerning capabilities and performance of air-operated valves (AOVs). 

Descriptions of air systems and AOVs were studied along with the support 

systems and equipment. Systems and equipment that contain AOVs and SOVs 

were studied to determine their dependencies. Applications of AOVs and SOVs 

were listed along with current NRC requirements. 

Observations and conclusions included: 

• Licensees may not know if the design basis loads or environmental 

conditions can be met with acceptable margins for the AOVs in their 

plants. AOVs may have reduced operating margins caused by such 

factors as aging, load mechanisms not understood or considered in the 

original design, or previously contaminated air. Calculations or valve 

descriptions were found that included mistakes or inaccurate 

information. 

"* Air systems and solenoid-operated valves (SOVs) have been and 

continue to be sources of common-cause failures of AOVs. 

Accumulators are potential sources of AOV failures. Air-operated 

dampers have been and continue to be potential sources of safety 

system failures. 

"* Accident sequence precursor (ASP) analyses performed by the NRC 

indicated that there have been a number of risk significant events 

involving AOVs. 

"* Generic probabilistic risk assessments performed by LNEEL indicated 

that changes in the failure probabilities of AOVs and SOVs can result 

in proportional changes in system unreliability and that AOVs and 

SOVs can have an important role in system reliability. It was also noted 

that some plants were using low generic probability values to estimate 

AOV failure probabilities in their PRA calculations. 

Examples were observed during the plant visits of events and conditions, 

including common-cause events and conditions, involving AOVs and SOVs 

which were under-reported or not reported to the NRC. This may have resulted in 

computing lower estimates of the risk and safety significance associated with 

AOV failures. 

Many of the licensees' individualized plant reviews for Maintenance 

Rule evaluations and for the plants' AOV programs included reviews of 

PRA findings by expert panels. The licensees found that by using their 

plants' operating experience and plant-specific probabilistic risk

NUREG/CR-6654 ii°
assessments, certain AOVs had high risk importance, high risk 

achievement worth, or were important to preventing large early 

releases. 

The nuclear industry and several licensees are preparing or have 

prepared AOV program plans which include design basis reviews, 

margin calculations, and use of diagnostic systems to ensure the 

operability of AOVs. 

NUREG/CR-6654 iv
CONTENTS 

A B ST R A C T ....................................................................................................................................... i.. 

EX ECUTIV E SUM M A RY ................................................................................................................ xi 

ACKN OW LED GM ENTS .................................................................................................................. xiii 

1. INTRODUCTION AND PURPOSE ..................................................................................... 1 

2. B A C K G R O U N D ..................................................................................................................... 3 

3. O B JEC T IV E S .......................................................................................................................... 5 

4 . SC O PE ..................................................................................................................................... 7 

5. DESCRIPTION OF AIR SYSTEMS AND AOVS ............................................................... 9 

6. APPLICATIONS OF AOVS IN NUCLEAR POWER PLANTS ........................................... 13 

7. CURRENT NRC EQUIPMENT QUALIFICATION, MAINTENANCE, AND 

TESTING REQUIREMENTS FOR AOVS IN NUCLEAR POWER PLANTS ................... 15 

8. AOV OPERATING EXPERIENCE AT THE PLANTS VISITED ....................................... 17 

8.1 AOV Operating Experience at Palo Verde 1, 2, and 3 (PVNGS), Trip No. 1 ............... 17 

8.1.1 Atmospheric Dump Valve Failures ................................................................. 17 

8.1.2 Letdown Containment Isolation Valve Leakage .............................................. 18 

8.1.3 Downcomer Feedwater Isolation Valve Leakage ............................................ 18 

8.1.4 Failure of a Vacuum Breaker Solenoid and Subsequent Investigations 

Involving SO V s ............................................................................................... 20 

8.1.5 AOV M argin Calculations ................................................................................ 20 

8.2 AOV Operating Experience at Enrico Fermi 2, Trip No. 2 .......................................... 21 

8.2.1 Failure of Multiple SOVs Controlling Safety-Related AOVs ......................... 21 

8.2.2 Failure of Multiple SOVs Controlling Safety-Related AOVs .......................... 22 

8.3 AOV Operating Experience at Palisades, Trip No. 3 .................................................... 23 

8.3.1 Failure of Multiple Air Regulators in the High-Pressure Air System ............. 23 

8.3.2 Failure of Shutdown Cooling Heat Exchanger Outlet Isolation Valve ........... 23 

8.4 AOV Operating Experience at LaSalle 1 and 2, Trip No. 4 ......................................... 24 

8.4.1 Pneumatic Valves with Less-Than-Designed Effective Diaphragm 

Areas Result in Inadequate Valve Closing Forces ........................................... 24 

8.4.2 Failure of MSIVs Due to Sticking Solenoid Operated Pilot Valves ............... 25 

8.4.3 Potential Failure of Safety-Related Dampers Due to Invalid Calculation 

Assumptions in the Original Builders Calculations ......................................... 25

V NUREG/CR-6654
8.5 AOV Operating Experience at Three Mile Island 1, Trip No. 5 ................................... 26 

8.5.1 Insufficient Design Margins in Aloyco Air-Operated Gate Valves ................ 26 

8.6 AOV Operating Experience at Indian Point 3, Trip No. 6 ........................................... 27 

8.6.1 Failure of Diaphragms in AOVs ....................................................................... 27 

8.6.2 Two Containment Isolation, In-Series AOVs were Inoperable Due to 

Inadequate D esign ............................................................................................ 27 

8.7 AOV Operating Experience at Turkey Point 3 and 4, Trip No. 7 ................................. 28 

8.7.1 Failure of Auxiliary Feedwater System AOVs from Air System 

C ontam ination ................................................................................................. 28 

8.7.2 Pilot Lockup Valves (POLVs) for Emergency Containment Cooler (ECC) 

V alves .............................................................................................................. 29 

8.7.3 Intake Cooling Water Header Inlet Isolation Valve Failed to Fully Close ....... 29 

8.7.4 RCS Letdown Isolation Valve Failure ............................................................. 30 

8.7.5 Revised Diaphragm Areas Furnished by AOV Vendor were Lower than 

D esigned .......................................................................................................... 30 

9. OTHER PERTINENT AOV EVENTS AND OPERATING EXPERIENCE ....................... 31 

9.1 Undersized AOV Acuators ........................................................................................... 31 

9.2 Weight and Center-of-Gravity Discrepancies in AOVs ................................................ 32 

9.3 Failure of Air Actuators Upon Gradual Loss of Air Pressure ...................................... 33 

9.4 Problems with Copes-Vulcan Pressurizer Power-Operated Relief Valves 

(PO R V s) ........................................................................................................................ 35 

9.5 Air-Operated Dampers ................................................................................................. 37 

9.6 AOVs and SOVs Used in Services Related to Emergency Diesel Generators ............. 38 

9.7 Establishment of and Conformance with the Design Basis for AOVs at 

M illstone 3 .................................................................................................................... 39 

10. AIR SYSTEMS OPERATING EXPERIENCE ....................................................................... 43 

10.1 A ir System s ................................................................................................................ 43 

10.2 Accumulators for AOVs ............................................................................................... 46 

11. SOLENOID OPERATED VALVE (SOV) OPERATING EXPERIENCE ............................ 49 

12. AOV PROGRAM PLANS AT THE SITES VISITED .......................................................... 53 

12.1 Palo Verde Nuclear Generating Station (PVNGS) AOV Program ............... 53 

12.2 Fermi 2 AOV Program .................................................................................................. 54

NUREG/CR-6654 vi
12.3 Palisades AOV Program ................................................................................................ 55 

12.4 LaSalle AOV Program .................................................................................................. 58 

12.5 Three M ile Island I AOV Program ................................................................................ 58 

12.6 Indian Point 3 (IP3) AOV Program .............................................................................. 59 

12.7 Turkey Point 3 and 4 AOV Program ............................................................................ 60 

13. INDUSTRY INITIATIVES REGARDING AOVS ................................................................. 61 

13.1 The M aintenance Rule .................................................................................................. 65 

14. DIAGNOSTIC SYSTEMS, MAINTENANCE, AND TESTING OF AOVS ........................ 67 

15. RISK ANALYSIS INVOLVING AOVS ............................................................................... 69 

15.1 Previous Risk Analyses Results ................................................................................... 69 

15.2 Accident Sequence Precursor Analysis ........................................................................ 71 

15.3 Recent AOV risk Studies .............................................................................................. 72 

15.3.1 Risk Studies in Several of the Plants Visited .................................................. 72 

15.3.2 Generic Issue 158 Draft Study and Subsequent Sensitivity Analysis ............. 73 

15.4 Common-Cause Failures ............................................................................................... 74 

15.5 AOV Failure Rates ........................................................................................................ 75 

15.6 Observations Pertinent to AOVs from Examination of Risk-Related Analyses and 

Probabilistic Safety Assessments (PSAs) for Motor-Operated Valves (MOVs) ........... 76 

15.6.1 BW R Owners Group Report NEDC-32264-A ................................................ 77 

15.6.2 W estinghouse Engineering Report V-EC-1658-A .......................................... 80 

16. CONCLUSIONS ...................................................................................................................... 83 

16.1 AOV Population ............................................................................................................ 83 

16.2 AOV Design Basis Demands and M argins .................................................................. 83 

16.3 AOV Failures ................................................................................................................. 84 

16.4 Air Systems ................................................................................................................... 84 

16.5 Accumulators ................................................................................................................. 85 

16.6 Solenoid-Operated Valves ............................................................................................. 85 

16.7 AOV Program Plans ...................................................................................................... 86

vii NUREG/CR-6654
16.8 Industry Initiatives Regarding AOVs ........................................................................... 86 

16.9 The M aintenance Rule .................................................................................................. 86 

16.10 Diagnostic Systems, Maintenance, and Testing of AOVs ............................................ 87 

16.11 Risk Analysis Involving AOVs .................................................................................... 87 

17. SAFETY SIGNIFICANCE ..................................................................................................... 89 

18. R E FE R EN C E S ........................................................................................................................ 91 

Appendix A-NRC Generic Communications Related to AOVs 

Appendix B-Air-Operated Valve Event Significance Analysis 

Appendix C-Trip Reports

NUREG/CR-6654 viii
LIST OF TABLES 

1. Populations of air-operated valves in plants visited ................................................................ 97 

2. Selected events documented by NRC Licensee Event Reports (LERs) ................................... 99 

3. Selected events documented by NRC Licensee Event Reports involving air-operated 

dampers or ventilation components that are powered from air systems .................................. 145 

4. Selected events described in reports other than Licensee Event Reports ................................ 155 

5. Licensee Event Reports cited in Tables 2, 3, and 7 listed in numerical order ......................... 169 

6. Air-operated valves considered by the licensees in the plants visited to be risk 

significan t ................................................................................................................................. 176 

7. Some recent (approximately last 5 years) events and conditions involving AOVs 

and/or air-operated components where the design basis was not met and/or not known ........ 187

ix NUREG/CR-6654
EXECUTIVE SUMMARY

The purpose of this study was to provide information to the NRC staff con

cerning air-operated valve (AOV) capabilities and performance in nuclear power 

plants, in order to help them determine if and how the NRC needs to focus addi

tional attention on the design, qualification, testing and/or maintenance of AOVs. 

One of the primary objectives of this study was to determine if AOVs can per

form their designed functions. 

Seven nuclear plant sites were visited, a literature search was conducted, and 

the NRC's databases were consulted to identify pertinent events and information. 

Descriptions of air systems and AOVs were studied along with the support sys

tems and equipment. Systems and equipment that contain AOVs and solenoid

operated valves (SOVs) were studied to determine their dependencies. Applica

tions of AOVs and SOVs were listed along with current NRC requirements. 

There are several hundred to over a thousand AOVs in each plant and typi

cally over 2000 SOVs. 

Events involving AOVs and reported conditions of AOVs indicate that many 

AOVs have reduced operating margins caused by such factors as aging, load 

mechanisms not understood or considered in the original design, and common

cause failures caused by currently or previously contaminated air or solenoid

operated valve defects. Licensees may not know that the design basis loads or 

environmental conditions can be met, with acceptable margins, for some of the 

AOVs in their plants. 

Licensees are relying on either manufacturers' information and calculations 

or, more recently, design reviews of the AOV design bases. Several manufactur

ers and nuclear plant architects/engineers' AOV calculations or valve descrip

tions were found that included mistakes or inaccurate information. The knowl

edge on the part of licensees of the design bases and margins in AOVs to meet 

the design basis demands is a matter of safety significance. 

Air systems are a source of common-cause failures of AOVs. A reliable sup

ply of clean, dry, oil-free air, at specified pressure is essential to proper function 

of AOVs and SOVs. Licensees must know that the air supply is reliable at all 

times in order to justify the assumptions that AOVs will move to or remain in 

their fail-safe positions. 

Accumulators that supply air to safety-related and important non-safety 

related AOVs can be sources of contamination, reduced capacity, and subsequent 

AOV inoperability unless it is verified that the accumulators are of sufficient 

capacity, are free of contamination, do not contain trapped water, and the accu

mulator check valves are functioning. Licensees must know that the accumula

tors are capable of functioning properly in order to justify the assumptions that 

AOVs will move or remain in their fail-safe positions. 

Air-operated dampers have been and can be a potential source of failure of 

emergency diesel generators, control room ventilation systems and other safety

related systems that they may serve.

xi NUREG/CR-6654
SOVs are a source of common-cause failure of AOVs. Root causes of AOV 

failure include potential contamination from the air system, design, qualification, 

and maintenance. These causes have been reported previously and need to be 

systematically addressed by licensees based on the safety status and risk signifi

cance of the SOVs. 

Generic probabilistic risk assessments performed by the INEEL indicated that 

changes in the failure probabilities of AOVs and SOVs can result in proportional 

changes in system unreliability and that AOVs and SOVs can have an important 

role in system reliability. It was also noted that some plants were using low 

generic probability values to estimate AOV failure probabilities in their PRA 

calculations. Individual AOVs may be found to be of low risk significance but 

the common-cause failure (CCF) of two or more AOVs performing the same 

function may have considerable risk significance. 

Many of the licensees' individualized plant reviews for Maintenance Rule 

evaluations and for the plants' AOV programs included reviews of PRA findings 

by expert panels. The licensees found that by using their plants' operating experi

ence and plant-specific probabilistic risk assessments, certain AOVs had high 

risk importance, high risk achievement worth, or were important to preventing 

large early releases. 

Among other observations, the site visits and reviews of events revealed that 

at some plants common-cause AOV failures, degradations, or precursors were 

under-reported or not reported to the NRC. This may have resulted in lower esti

mates of the risk and safety significance of AOVs in several studies. The under

reporting of common-cause AOV events could lead to lower computed beta fac

tors and lower estimates of core damage frequency and large early releases. 

The number and scope of NRC generic communications and studies of AOV 

and air systems provide an important indicator of the overall safety significance 

of these components and systems. Over 100 NRC generic communications 

related to AOV events and problems were identified. Accident Sequence Precur

sor (ASP) analyses included a number of risk significant events involving AOVs. 

The nuclear industry and several licensees are preparing or have prepared 

AOV program plans, including testing, that are in varying stages of development. 

The plans are not, as yet, fully implemented. Diagnostic systems for AOVs and 

SOVs are available but manufacturers' claims regarding accuracy and attributes 

have yet to be verified. 

Tables of pertinent events to support the conclusions are included, along with 

trip reports from the visits to the sites.

NUREG/CR-6654 xii
ACKNOWLEDGMENTS 

The authors wish to express their appreciation to Dr. Harold Ornstein of the 

United States Nuclear Regulatory Commission (NRC), Office of Nuclear Regu

latory Research staff, who provided attention, direction, support, and a wealth of 

knowledge of the subject. We also wish to thank W. A. Christenson, S. A. Eide, 

R. D. Fowler, J. W. Hudson, J. J. Jones, P. H. McCabe, and J. E. Trainer, of the 

INEEL engineering staff, who participated in, and provided engineering support 

for, this effort.

xiii NUREG/CR-6654
A Study of Air-Operated Valves 

in U.S. Nuclear Power Plants 

1. INTRODUCTION AND PURPOSE

The NRC, with assistance from the INEEL, 

studied the design, qualification, operation, 

maintenance, and testing of air-operated valves 

(AOVs) in commercial nuclear power plants. 

The study was conducted in accordance with the 

Program Plan dated October 22, 1997, (INEEL 

Letter to H. Ornstein, NRC, from J. Bryce, 

October 23, 1997, Job Code E8238, Task 

Order 15 - JHB-167-97 and INEEL Letter 

from T. Ryan to H. Ornstein, NRC, Transmittal 

of Revised Cost Estimate, Schedule, and 

Spending Plan for Revision 2 of JCN 8328, Task 

Order No. 15, Investigation of Air-Operated 

Valves, March 25, 1999, RYAN-55-99).

Seven nuclear plant sites were visited as part 

of this study. 

The purpose of this study was to provide 

information to the NRC staff concerning AOV 

capabilities and performance in nuclear power 

plants. The goal was to help the NRC determine 

if and how the NRC needs to focus additional 

attention on the design, qualification, testing 

(initial and in-service), and/or maintenance of 

AOVs in order to reduce plant vulnerabilities 

associated with individual or common-cause 

failures. Design, qualification, applications, 

maintenance, and testing of AOVs were studied 

in relation to their safety significance in nuclear 

power plant applications.

I NUREG/CR-6654
2. BACKGROUND

AOVs are used in all U.S. nuclear power 

plants. The population of AOVs in each plant 

varies widely. The number of AOVs per plant 

can be over a thousand and the number of 

safety-related AOVs per plant can be several 

hundred. Many plants have large numbers of 

"important" AOVs that are not necessarily des

ignated "safety-related." Solenoid-operated 

valves (SOVs) in most nuclear power plants 

number in the thousands. 

"Important" AOVs are those that could inter

fere with the function of safety-related equip

ment or can cause scrams or trips, or, if they 

should fail may result in release of radioactive 

contamination. For example, such a failure was 

part of the initial sequence of the TMI-2 event. 

"The loss of the main feedwater pumps which 

initiated the turbine trip followed by a reactor 

trip, has been attributed to the presence of water 

in the instrument air system that caused the con

densate polisher air outlet valves to close. It is 

postulated that water at 100 psig in the conden

sate polisher entered the service air system, 

which is at 80-100 psig, through a failed open 

check valve." (TMI Report, Vol. H, Part 2, pp. 

470-471) 

The NRC and the nuclear industry have been 

interested in the safety significance and per

formance of AOVs for many years and have 

conducted a number of studies of various appli

cations of these components and the air systems 

that provide motive power. Studies by both the

NRC and industry are referenced herein. In 

addition, a list of generic communications was 

compiled (Appendix A) related to AOVs, SOVs, 

or air systems. There are undoubtedly other 

pertinent studies or references that could also 

have been included. 

The NRC's Office for the Analysis and 

Evaluation of Operational Data (AEOD) pub

lished two case studies, NUREG-1275, Volume 

2 on air system problems and NUREG-1275, 

Volume 6, on SOV problems. These studies did 

not focus directly on AOVs but they contained 

relevant information on two of the dominant 

contributors to AOV failures and demonstrated 

plant vulnerability to common-cause failures of 

AOVs. 

Relatively recent experience in the nuclear 

industry with motor-operated valves (MOVs), 

specifically the focus on determining design 

basis demands vs. capabilities (margins) has led 

to an increased awareness by all concerned of 

the operational requirements of AOVs. Various 

groups of licensees, nuclear industry organiza

tions, and vendors are currently involved in 

efforts to ensure the reliable performance of 

AOVs in nuclear power plants. These efforts 

include pilot programs and individual plant ini

tiatives concerning diagnostic testing and design 

verification. The NRC has not, as yet, requested 

that licensees construct programs for AOVs 

similar to the efforts described in Generic Letter 

89-10 and its supplements with respect to motor

operated valves.

3 NUREG/CR-6654
3. OBJECTIVES

One of the primary objectives of this study 

was to determine if safety-related air-operated 

valves are designed, qualified, installed, main

tained, and tested so that there is a reasonable 

assurance that they can perform their design 

basis functions. Important non-safety-related 

AOVs (see the SCOPE section) were also stud

ied to determine if they could function so as not 

to cause or compound events that might com

promise safety. 

Sources of common-cause failures of AOVs, 

such as the air system and SOVs, were studied 

because of their potential impact on safety. 

Operating experience over the last 14 years 

(plus) was reviewed to help the NRC draw con

clusions about the effectiveness of current regu

lations and guidance on AOVs. The safety sig

nificance of the operating experience was 

evaluated. 

Regulatory requirements regarding AOVs 

were reviewed and assessed against the operat

ing experience. 

Domestic activities in the nuclear industry 

regarding AOVs, including development of 

AOV programs and use of diagnostic testing, 

were studied in an effort to assess their effec

tiveness in ensuring AOV operability.

Three common assumptions regarding AOV 

operability were critically reviewed as part of 

this study. These assumptions are that AOVs 

will "fail-safe," that providing clean, dry, oil

free air, at proper pressure, is all that is neces

sary to ensure the satisfactory performance of 

AOVs, and that the American Society of 

Mechanical Engineers (ASME) stroke-timing 

test (the test currently mandated by the NRC for 

safety-related AOVs) provides reasonable assur

ance of the operability of AOVs without aug

mentation. Results of the review are included in 

this study. The study results indicated that the 

"fail-safe" assumption that many AOVs are 

designed for (i.e., to move to a specified position 

in the event of a loss of air) depends on proper 

function of valves, valve operators, and compo

nents associated with AOVs or in proximity to 

them, and is not valid unless verified. Also, a 

clean, dry, oil-free air (or nitrogen) supply at 

proper pressure is necessary for the proper func

tion of AOVs; but it is not sufficient, by itself, to 

ensure the acceptable function of AOVs. Finally, 

the ASME stroke-timing test is generally per

formed as a no-load test that does not provide 

sufficient information regarding future operabil

ity of AOVs used in nuclear power plant 

applications.

5 NUREG/CR-6654
4. SCOPE

Safety-related and important non-safety

related AOVs (including directly attached com

ponents such as operators, solenoids, springs, 

and diaphragms) were included in this study. 

AOVs are defined as valves that use air or inert 

gas as the motive power source to change the 

position of valve, valve operator, or a compo

nent of the valve. Other parts or mechanisms 

such as pneumatic regulators, pneumatic con

trollers, pneumatic boosters, pneumatic reducers, 

or solenoids are considered part of the 

valve/operator if they serve a particular valve, 

even if they are not directly within its casing or 

mounted on it. 

Solenoid-operated valves (SOVs) may be 

piece parts of AOVs, serve as pilots for AOVs, 

or in some cases, directly act as valves. SOVs 

are considered to be air-operated valves in this 

study even though the device that actuates an 

SOV is an electric powered solenoid. SOVs that 

are used as pilots or piece parts of AOVs are 

closely tied to the operation of the AOVs and 

use the same air or nitrogen that operates AOVs 

as their process fluid. The process fluid is very 

closely related to the operation of an SOV and in 

some cases provides some of the force to actuate 

or control the motion of an SOV. SOV operabil

ity depends so heavily on the quality of the pro

cess fluid (air or nitrogen) that classifying SOVs 

as a category of AOVs is reasonable. The 

phrase, "quality of the process fluid (air or nitro

gen)" refers to the necessity that air or nitrogen 

be clean, dry, oil free, and delivered at pressures 

within the range specified for the SOV. 

Safety-related AOVs are valves and opera

tors that: 

"* Must remain functional during and fol

lowing design basis events; 

"* Ensure the integrity of the reactor coolant 

pressure boundary; 

"* Ensure the capability to shut down the 

reactor and maintain it in a safe-shutdown 

condition; or

Ensure the capability to prevent or miti

gate the consequences of accidents that 

could result in potential offsite exposure 

comparable to the 10 CFR 100 guidelines. 

Important non-safety-related AOVs valves 

and operators are: 

"* Relied upon to mitigate accidents or tran

sients; 

"* Used in plant emergency operating proce

dures (EOPs); 

"* Those whose failure could prevent safety

related structures, systems, and compo

nents (SSCs) from fulfilling their safety

related functions; or 

Those whose failure could cause a reactor 

scram (trip) or actuation of a safety

related system. 

The above categorizations are in accordance 

with the information in 10 CFR 50.2, 50.49, 

and 50.65. 

Safety-related and important non-safety

related AOVs are expected to perform their 

intended function under design-basis conditions, 

normal operating conditions, and, in some cases, 

shutdown conditions. 

Design, qualification, maintenance, and test

ing of AOVs were studied in nuclear power 

plant applications. 

Air or inert gas supply systems and compo

nents were included in this study because they 

are a source of common-cause failures in the 

equipment served. The air or inert gas operating 

fluid might be supplied from the plant's pneu

matic systems such as instrument air/service air 

systems, accumulators, receivers, emergency 

tanks, or from the process fluid.

7 NUREG/CR-6654
Scope

Electric power, as well as air, is usually 

required for AOVs in order to provide power for 

control components such as limit switches and 

SOVs associated with the AOVs. The electrical 

supply system for AOVs was not included in 

this study. However, electrical components that 

serve AOVs directly, such as limit switches and 

SOVs were noted in some of the events.

In addition to a review of the NRC's Licen

see Event Report (LER) database, searches for 

sources of other pertinent AOV related events 

were conducted in other available industry and 

NRC databases. Final Safety Analysis Report 

(FSAR) reviews were conducted for the seven 

plants visited in order to learn about the individ

ual air systems and plant characteristics.

PLANTS VISITED FOR AOV STUDY

Trip Number and Plant 

Name Dates Visited

Plant Description / A&E / Year Commercial 

Operations Started

1. Palo Verde 1, 2, and 3 October 28-29, 1997

2. Fermi 2 

3. Palisades 

4. LaSalle 1 and 2 

5. Three Mile Island 1 

6. Indian Point 3

November 3-4, 1997 

November 18-19, 1997 

December 17-18, 1997 

February 12-13, 1998 

March 10-11, 1998

7. Turkey Point 3 and 4 March 24-25, 1998

A trip report was prepared after each visit 

and these are included in Appendix C.

Combustion Engineering, 2 loop, System 80 (no 

PORVs) PWR/Bechtel/1986 

General Electric BWR 4/Detroit Edison/1988 

Combustion Engineering, 2 loop 

PWR/Bechtel/1971 

General Electric BWR 5/Sargent & Lundy/1984 

Babcock and Wilcox, lowered loop 

PWR/Gilbert/1974 

Westinghouse, 4 loop PWRIUnited Engineers & 

Constructors/1976 

Westinghouse, 3 loop PWR/Bechtel / 1972

This report includes consideration of public 

comments received on the draft dated 

April 26, 1999.

NUREG/CR-6654 8
5. DESCRIPTION OF AIR SYSTEMS AND AOVS

The purpose of air and/or nitrogen systems 

that supply power to AOVs is to provide clean, 

dry, oil-free air and/or nitrogen at a specified 

pressure. Generally, the instrument air (IA) sys

tem and its backups perform this task. Air sys

tems in nuclear plants are unique to each plant or 

site and no two sites have the same system. Air 

systems, in general, are designed and maintained 

as non-safety-related systems, although portions 

of air systems and separate air systems have 

been designated as safety-related in some plants. 

General descriptions of IA (and backup nitro

gen) systems are to be found in the introductory 

parts of: 

NUREG-1275, Volume 2, "Operating 

Experience Feedback Report - Air Sys

tems Problems;" 

"* NUREG/CR-2796, "Compressed-Air and 

Backup Nitrogen Systems in Nuclear 

Power Plants;" and 

"* NUREG/CR-5419, "Aging Assessment of 

Instrument Air Systems in Nuclear Power 

Plants." 

A comprehensive and detailed overview of 

nuclear power plant IA systems and their major 

components was found in the Instrument Air 

Systems Guide for Power Plant Maintenance 

Personnel, NMAC NP-7079. Descriptive remarks 

concerning the particular air/nitrogen systems, 

including IA systems, encountered in the visits to 

plant sites are included in the trip reports in 

Appendix C. 

Air-operated valves can range from "simple" 

diaphragm actuators to complicated, solenoid

piloted designs. The design options for AOVs 

and air-operators are virtually unlimited. Con

ceptually, some AOVs may appear to be 

straightforward devices but the details of design, 

operation and maintenance indicate that even the 

"simple" diaphragm AOV is a complex device. 

Those complexities include materials and mate

rial interactions, tolerances, and an almost 

unlimited variety of design details. Further, if 

the additional complexities of support systems

are considered, the challenge of ensuring reliable 

AOV operation is considerable. A number of 

vendors have provided AOVs for nuclear power 

plant applications, including Fisher, Copes

Vulcan, WKM, Anchor Darling, and Target 

Rock, to name a few. Fermi 2, for (perhaps an 

uncommon) example, has AOVs from approxi

mately 38 manufacturers. 

The following references provide general 

descriptions of AOVs and their associated con

trol mechanisms: 

"* NUREG/CR-6016, "Aging and Service 

Wear of Air-Operated Valves Used in 

Safety-Related Systems at Nuclear Power 

Plants;" 

"* Electric Power Research Institute (EPRI) 

report EPRI NP-7412, "Maintenance 

Guide for Air-Operated Valves, Pneu

matic Actuators, and Accessories;" and 

* EPRI NP-7412, Revision 1, "Air

Operated Valve Maintenance Guide." 

A description and discussion of the operation 

of main-steam isolation valves (MSIVs) are 

included in NUREG/CR-6246, "Effects of 

Aging and Service Wear on Main Steam Isola

tion Valves and Valve Operators," and power

operated relief valves (PORVs) are described in 

NUREG/CR-4692, "Operating Experience 

Review of Failures of Power Operated Relief 

Valves and Block Valves in Nuclear Power 

Plants." A detailed discussion of valve and valve 

trim choices is included in the "ISA Handbook 

of Control Valves." 

Globe, ball, and butterfly valves are com

monly used in AOV applications. Air-operated 

gate valves are less common but some are to be 

found in nuclear power plants. The two most 

common AOV actuator types are spring-and

diaphragm and piston. Valve and actuator 

assemblies can be designed for linear motion or 

for rotary valve motion using scotch yoke or 

rack-and-pinion arrangements. AOVs can also 

be classified as control valves or isolation

9 NUREG/CR-6654
Description of Air Systems and AOVs 

valves. An AOV "assembly" might be consid

ered to include a dedicated air regulator, posi

tioner, pressure controller, dedicated accumula

tor(s), accumulator check valves, and dedicated 

SOVs used as pilots for the valve operator. 

Globe AOVs can be cage guided, post 

guided, stem guided, or skirt guided. The valve 

plugs can be balanced or unbalanced designs. A 

variety of packing materials and packing con

figurations are available, each with its own 

attributes, advantages and disadvantages. 

Air regulators are devices used to provide 

uniform air pressure. An AOV positioner is a 

proportional controller that adjusts the air output 

to the actuator based on control system input and 

feedback from the valve stem position. Position

ers can be pneumatic or electro-pneumatic; they 

are used to automatically adjust the flow through 

AOVs in response to changes in the characteris

tics of the valve or fluid system. Pressure boost

ers or volume boosters are used to increase the 

pressure or volume from the common air supply 

to meet specific needs of an AOV. Pneumatic 

transducers are used in AOVs to convert current 

or voltage from the control system into pressure. 

Accumulators are used as backup pneumatic 

supplies for safety-related AOVs in order to 

avoid direct reliance on non-safety-related air 

systems and/or when multiple repositioning of a 

valve is required without reliance on common 

air systems. Accumulator check valves are 

installed to prevent the air from bleeding out of 

an accumulator. 

SOVs can be used as pilots to control the 

motion of larger AOVs, or they can be designed 

to operate equipment, for example dampers, 

directly. Conceptually, solenoid-operated valves 

may appear to be simple devices consisting of a 

coil that, when energized, causes the linear 

motion of a magnet in order to activate a valve 

(usually small) to direct a fluid (usually air or 

nitrogen). In practice, because of the number of 

SOVs, their use in many systems throughout 

nuclear power plants, and the number of design, 

operation and maintenance conditions that must 

be satisfied in order to ensure their successful 

operation, establishing and maintaining SOV 

operability represents one of the more complex

and challenging aspects of nuclear plant opera

tion. The considerable complexities regarding 

SOV operability arise in the details of design, 

fabrication, installation, maintenance, and test

ing, all of which involve satisfying numerous 

environmental conditions and operating 

requirements in order to ensure successful 

operation of each of the thousands of SOVs in 

each plant. 

The operation and performance of SOVs 

have been the subject of a number of studies by 

the NRC and the nuclear power industry. 

Detailed technical descriptions of SOVs, 

including their design, application, and mainte

nance, are to be found in the "Solenoid Valve 

Maintenance and Application Guide," NMAC 

NP-7414, and also in NUREG/CR-4819, Vol

ume 1, "Aging and Service Wear of SOVs in 

Safety Systems of Nuclear Power Plants." The 

uses and description of SOVs in nuclear power 

plant applications are also described in the 

beginning of NUREG- 1275, Volume 6, "Oper

ating Experience Feedback Report - Solenoid

Operated Valve Problems." 

Characteristics of SOVs pertinent to this 

study, and which raise the likelihood for com

mon-cause AOV failures, include: 

"* Exceeding the designed allowable maxi

mum operating pressure differential 

(MOPD), which is the pressure difference 

between the inlet port and the outlet port, 

may cause the SOV to spuriously open or 

not open, depending on the SOV design, 

and will usually result in future unreliable 

operation or failure. 

"* Small orifices make the SOVs subject to 

interference from small contamination 

particles or moisture intrusion. 

"* The design logic of SOV operation and 

control can be complex and this is a 

source of design or installation mistakes. 

"* Many piloted SOVs require a minimum 

operating pressure differential (Min OPD) 

to operate properly.

NUREG/CR-6654 10
"* Many SOVs will not function properly 

when subject to reverse pressurization or 

flow in the valve. 

"* Materials used to construct the valve 

bodies, internal parts, seals, O-rings, etc. 

may be subject to binding, creep, corro

sion, erosion, adverse material interac

tions, and/or environmental deterioration. 

"* Rather small forces, in the range of 

10 pounds or less, are produced by a 

solenoid to operate an SOV; therefore, 

opposing forces of similar small magnitude 

can interfere with successful operation. 

"* SOVs can be damaged if subjected to 

higher-than-designed pressure in the 

operating fluid.

Description of Air Systems and AOVs 

"* SOVs are subject to damage from the use 

of thread locking compounds in adjacent 

pipe or tube connections that tend to 

migrate into the working parts of the 

SOVs. 

"* Incompatible, incorrect, or unapproved 

elastomers and lubricants used in SOVs 

have been, and continue to be, a source of 

many SOV operational problems. 

According to section 4.2 of the "Solenoid 

Valve Maintenance and Application Guide," 

NMAC NP-7414, nuclear power plants each 

contain between 1000 and 2500 SOVs, and of 

these about 20 to 50% are used in safety-related 

applications.

I1I NUREG/CR-6654
6. APPLICATIONS OF AOVS IN NUCLEAR POWER PLANTS

All U.S. nuclear power plants use AOVs and 

some applications appear to be common to most 

if not all plants. For example, U.S. light water 

reactors (LWRs) use AOVs for containment 

isolation functions and for control of main 

steam. U.S. boiling water reactors (BWRs) use 

AOVs in conjunction with SOVs to control their 

scram systems. U.S. pressurized water reactors 

(PWRs) use AOVs for controlling auxiliary 

feedwater, main feedwater, and condensate sys

tems. 

The majority of AOVs at U.S. LWRs are 

non-safety-related and are generally associated 

with the non-nuclear balance-of-plant systems. 

However, from observations at the plants vis

ited, there are significant numbers of safety

related AOVs as well as important non-safety

related AOVs in U.S. LWRs. The number and 

general safety categorizations of AOVs in the 

plants visited as part of this study are described 

in Table 1 (because of the size of the tables in 

this report, all of them are presented at the end 

of the main text). 

Applications of AOVs in nuclear power 

plants include: 

"* Auxiliary feedwater discharge to steam 

generator isolation and flow control 

valve; 

"* Auxiliary/emergency feedwater turbine 

trip stop valve; 

"* Chemical and volume control system 

(CVCS) letdown isolation valves; 

"* Containment spray header isolation valve; 

"* Control rod drive scram discharge volume 

drain valve; 

"* Emergency diesel generator air start iso

lation valve; 

"* Feedwater containment isolation valve; 

"* Feedwater regulating valve bypass valve;

* High-pressure core spray emergency die

sel generator air start isolation valve; 

* Low-pressure core spray containment 

isolation valve; 

* Main feedwater regulating valve; 

* Main steam safety/automatic depressuri

zation system (ADS) valve; 

Main steam turbine bypass valve; 

* Pressurizer PORV; 

* Reactor coolant pressurizer spray valve; 

* Reactor core isolation cooling discharge 

to feedwater check valve; 

• Residual heat removal refueling water 

storage tank suction isolation valve; 

* Standby gas treatment system blower 

isolation valve; 

Standby gas treatment system filter isola

tion valve; 

* Suppression pool/torus vacuum breaker 

valve; and 

Ventilation dampers for the control room 

and diesel generator air supply applica

tions. 

Applications of SOVs in nuclear plants 

include: 

* Instrument air-pilot valves for AOVs; 

BWR control-rod drive and SCRAM 

system (over 300 per BWR plant, 

according to NUREG/CR-4819, Vol. 1); 

High-pressure/temperature (e.g., 2500 psi/ 

600'F) steam/water valves for primary 

and secondary systems; and

13 NUREG/CR-6654
Applications of AOVs in Nuclear Power Plants 

Low-pressure/temperature (e.g., 160 psi/ 

2000'F) process valves for flow control in 

miscellaneous plant systems.

Overall numbers of valve populations in 

nuclear power plant applications may include 

some SOVs as piece-parts of AOVs.

NUREG/CR-6654 14
7. CURRENT NRC EQUIPMENT QUALIFICATION, 

MAINTENANCE, AND TESTING REQUIREMENTS 

FOR AOVS IN NUCLEAR POWER PLANTS

The NRC's equipment qualification require

ments evolved over several decades. As a result, 

the criteria regarding equipment qualification, 

particularly environmental qualification of 

mechanical equipment, may vary among indi

vidual nuclear power plants. Current general 

requirements for environmental design and 

qualification of nuclear power plant equipment 

are summarized in the Acceptance Criteria sub

section of Section 3.11 of the NRC's Standard 

Review Plan (SRP), NUREG-0800, as follows: 

The equipment shall be designed to have 

the capability of performing its design 

safety functions under all normal, acci

dent, and post-accident environments for 

the length of time that its function is 

required; 

"* The equipment environmental capability 

shall be demonstrated by appropriate 

testing and analysis; and 

"* A quality assurance program meeting the 

requirements of 10 CFR 50, Appendix B, 

(Quality Assurance Criteria) shall be 

established and implemented to provide 

assurance that all requirements have been 

met. 

Criteria and reference standards for the con

sideration of "harsh" and "mild" environmental 

qualification of electrical and mechanical 

equipment are also included in Section 3.11 of 

the SRP. 

Most AOVs contain organic parts such as 

diaphragms, 0-rings, and seals that are subject 

to environmental deterioration and aging. Sev

eral licensees spoke about qualified life regard

ing diaphragms and seals during the visits and 

referred to requirements in their Quality Assur

ance programs to check such data. 

Maintenance requirements for nuclear power 

plant equipment of interest to the NRC

(including AOVs) are driven by the require

ments of 10 CFR 50, Appendix B, "Quality 

Assurance Criteria," the recent 10 CFR 50.65, 

"Requirements for Monitoring the Effectiveness 

of Maintenance at Nuclear Power Plants" (the 

Maintenance Rule), and licensee commitments 

toward the operability of systems and equipment 

described in individual plant FSARs. Preventive 

and corrective maintenance for safety-related 

and important non-safety-related AOVs is gov

erned by these requirements. Few licensees 

(based on the experience with those visited or 

consulted) have developed predictive mainte

nance programs for AOVs. 

Testing of safety-related AOVs is governed 

by each licensee's licensing-basis commitments 

toward specific editions of the valve testing 

requirements of the "American Society of 

Mechanical Engineers Boiler and Pressure Ves

sel Code" (ASME Code). Section XI of the 

ASME Code requires a stroke-timing test at 

three month intervals for safety-related power

operated valves. The test interval can be as long 

as each refueling outage, or even waived 

entirely, if operational safety considerations 

dictate. The test consists of stroking the valve in 

the safety-significant direction (or directions) 

and measuring the time required to complete the 

stroke. Licensee actions regarding re-testing or 

maintenance would depend on comparisons with 

required stroke time and/or comparisons with 

trends of previous test results. The performance 

of any corrective or follow-on surveillance 

would be dictated by safety and operational con

siderations. Generally, the valve stroke-timing 

test is performed with no fluid flow and no fluid 

pressure (or low fluid pressure) in the pipeline. 

No specific periodic testing requirements are 

imposed by the NRC for non-safety-related 

AOVs as individual components. System flow 

testing may have been conducted at startup or 

immediately after systems (containing AOVs) 

have undergone maintenance or repair. It is logi

cal to assume that all AOVs would be stroked, if

15 NUREG/CR-6654
Current NRC Equipment Qualification, 

Maintenance, and Testing Requirements 

not stroke-time tested or tested against flow or 

pressure, after repair and prior to being declared 

ready for service (post-maintenance testing). 

These tests provide useful information about the 

performance of AOVs, however, testing may be 

of limited value if the AOVs are not observed 

and/or instrumented when they are opened and

closed during such tests. These tests will also 

provide limited assurance of operability under 

design basis conditions when design basis or full 

flow or pressure conditions if there is no (or 

low) fluid pressure or flow in the line (the usual 

situation) when the test is performed.

NUREG/CR-6654 16
8. AOV OPERATING EXPERIENCE AT THE PLANTS VISITED

Information regarding AOV performance 

was gathered during the visits for this study and 

is discussed below. 

Events or conditions of particular interest 

were those involving common-cause failures or 

potential failures, such as design problems or 

environmental conditions that could adversely 

affect the performance of a number of AOVs 

simultaneously or over time. Table 6 provides a 

list of air-operated valves considered to be risk 

significant by the licensees in the plants visited. 

Each licensee's input to both the LER data

base and the nuclear industry's databases 

(NPRDS or EPIX) may vary. Events involving 

non-safety-related AOVs might or might not be 

included in either or both databases, depending 

on circumstances. Interpretations about what is or 

is not reportable to either or both databases 

appears to be somewhat variable. Some events or 

conditions covered by LERs were not reported to 

the industry's database and some events or 

conditions reported to the industry's database 

were not reported as LERs. There were many 

events (including failures) which were reported in 

licensees' condition reports but were not reported 

in LERs or otherwise reported to the NRC. For 

example, the Palisades events involving common

cause degradation and failure of air regulators 

were not covered by LERs at the time of the visit 

for this study. The LER and nuclear industry 

databases do not capture all of the information 

about AOV failures and other pneumatic 

equipment failures received during the site visits. 

The licensees interviewed indicated that their 

reporting was in compliance with their 

interpretations of current NRC requirements.

8.1 AOV Operating Experi

ence at Palo Verde 1, 2, 

and 3 (PVNGS), Trip No. 1 

8.1.1 Atmospheric Dump Valve 

Failures 

For each PVNGS plant, one pneumatic oper

ated atmospheric dump valve (ADV) is installed 

in each of the four main steam lines to allow 

cooldown of the steam generators when the 

MSIVs are closed or when the main condenser is 

not available as a heat sink. Each valve is 

designed so as to allow for an unavailable steam 

generator, concurrent with a loss of AC power 

and a single failure of one of the remaining 

ADVs. On March 3, 1989, after a plant trip, all 

four ADVs on Unit 3 failed to respond when 

given an open signal from the control room. The 

only ADV tried from the remote-shutdown panel 

also failed to respond. Refer to LER 52889005 

and NRC Information Notice 89-38. Manual 

local action was initiated to open one ADV per 

steam generator, however an actuator for one of 

the ADVs was damaged in doing so. On 

March 5, 1989, a test was conducted on the 

Unit 1 ADVs and one of the ADVs failed to 

open when given a 50% open demand signal 

from the control room. Shortly thereafter, all 

three units were shut down. 

The results of the investigation indicated that 

excessive piston ring leakage, combined with 

inadequate pilot valve relieving capacity, created 

high forces in the valve bonnet (also called the 

balance chamber) that could not be overcome by 

the actuator. Other problems were also found

17 NUREG/CR-6654
AOV Operating Experience at the Plants Visited 

that compromised the operability of the ADVs. 

These included: 

Valve oscillations caused by lower than 

required nitrogen pressure (the regulators 

exhibited excessive leakage); 

"* Positioners were not adjusted and/or 

maintained properly; 

"* Springs were left on the valve operators 

that should have been removed prior to 

startup; 

"* An actuator piston was fitted with non

qualified Buna-N 0-ring rather than a 

qualified Viton O-ring; 

"* Air and nitrogen quality was suspect 

(particulate contamination); and 

Several non-qualified pressure gauges 

were left installed on the positioners. 

Several or these problems were common

cause failure mechanisms which were applicable 

to either the valve design, valve operator design, 

or deficiencies in regulators and positioners. 

The licensee forwarded a 10 CFR Part 21 

report to the NRC. Other plants that have the 

same or similar ADV AOVs are Catawba 1 

and 2, SONGS 2 and 3, and Waterford 3. 

8.1.2 Letdown Containment Isolation 

Valve Leakage 

In 1995, PVNGS completed an investigation 

of the recurring seat leakage, over several years, 

of three letdown containment isolation valves. 

LER 52895007 (a previous LER 52894009 also 

applies) and PVNGS Condition Report 95Q028 

of May 11, 1995, described the results of the 

investigation. The most probable common 

causes of the seat leakage problems were: 

Undersized pneumatic actuators resulting 

from not accounting for the high frictional 

loads of graphite style packing during 

original sizing of the actuators.

Not maintaining the specified bench set 

on the spring-and-diaphragm actuators. 

"Bench set" is a term used for spring-and

diaphragm AOVs to denote the pressure 

range (pressure to the actuator diaphragm) 

through which the valve will stroke its 

full travel with the actuator uncoupled 

(packing friction, differential pressure, 

and seating forces are excluded) from the 

valve. Basically, the bench set determines 

the preload on the actuator spring. 

There are three letdown isolation valves in 

each unit at PVNGS. Two of the valves also 

serve as containment isolation valves. The 

valves are 2-inch Fisher globe valves. Flow 

pressure tends to open the valves. The actuators 

are Fisher pneumatic spring-and-diaphragm 

actuators. The valves are specified to be able to 

open and close against a differential pressure of 

2485 psig. 

Modifications were made to all three AOVs 

with undersized actuators. Stroke lengths were 

reduced, the actuators' springs were replaced and 

the bench sets were increased on all three 

actuators. Prior to the modification, one actuator 

had the bench set increased to provide seat load. 

Later, it was discovered that the higher bench set 

exceeded the manufacturer's maximum recom

mended safe spring load and stem load. PVNGS 

thereafter determined that the higher bench set 

was acceptable because the valve is normally 

open and rarely stroked and thus the assumed 

1000 cycles used in the calculations was 

conservative. 

8.1.3 Downcomer Feedwater Isolation 

Valve Leakage 

Three downcomer feedwater isolation valves 

(DCFWIVs) at PVNGS failed to open, follow

ing closure after a main steam isolation signal 

(MSIS) during a Unit 1 reactor trip on Novem

ber 26, 1995. LER 52895012 covered the reactor 

trip but not the DCFWIV failures. The source of 

the information concerning the DCFWIV fail

ures was a comprehensive report provided by the 

licensee during the site visit for this study. The 

report (the copy received was untitled) included

NUREG/CR-6654 18
AOV Operating Experience at the Plants Visited

a root-cause analysis and failure mode evalua

tion, description of the DCFWIVs, corrective 

actions description, Maintenance Rule consid

erations, and observations about the potential for 

related failures of other valves. 

The attempt to reopen the DCFWIVs was 

made about 14 hours after their closing in 

response to the MSIS, during which time the 

valves had cooled considerably. System pressure 

upstream of the DCFWIVs was bled off, result

ing in minimal pressure across the valve disc 

and a reduction in required actuator force. The 

three valves still would not open. Finally, an 

auxiliary feedwater pump was started and the 

motor-operated valves upstream of the 

DCFWIVs were opened, providing a slug of 

water that (it is postulated) finally dislodged the 

DCFWIV discs and allowed the valves to open 

successfully using available nitrogen pressure. 

There are four DCFWIVs in each unit, two in 

series to each of the two steam generators. The 

DCFWlVs are 8-inch flex-wedge gate AOVs. 

The actuators are Miller Fluid Power single

acting pneumatic cylinders with internal cylinder 

springs and external spring-and-stanchion set

ups. The actuators are designed to use pressur

ized nitrogen to open the valves and spring force 

to close them. The pneumatic control circuits 

include a 3-way SOV and a 3-way air switch 

valve. The valves are intended to fail closed on 

loss of the air signal to the 3-way air switch 

valve or loss of nitrogen to the operator. The 

valves are intended to fail open if electric power 

to the SOVs is lost. The DCFWIVs have three 

key safety functions: heat removal, trip initia

tion, and containment isolation. These valves are 

also Appendix R safe-shutdown components. 

The most probable root causes for the multi

ple valve failures to open were: 

* A lack of prudent actuator design margin. 

The low actuator margin resulted from 

using a nonconservative valve factor (0.3) 

in the original sizing. Recent tests on 

motor-operated gate valves in response to 

NRC Generic Letter 89-10 indicate that 

0.3 is not a conservative valve factor for 

flex-wedge gate valves.

"Not allowing for the potential effects of 

thermal binding in the original sizing of 

the actuator. Pressure locking and thermal 

binding of power-operated gate valves, 

including AOVs, is discussed in NRC 

Generic Letter 95-07. The DCFWIVs 

were excluded from the licensee's 

GL 95-07 evaluation because they are 

normally open valves. 

"Not allowing for the potential effects of 

degradation of the nitrogen supply in the 

calculation of actuator margin. 

Results from static diagnostic testing did not 

reveal why three of the DCFWlVs failed to open 

and the remaining AOV opened. A dynamic test 

program to establish the operability of these 

AOVs was conducted and valve factors of 0.64 

and 0.57 were established. 

From the perspective of this study, the sig

nificance of these AOV failures was that the 

actuators were undersized to open the valves. 

The actuators were not known to be marginally 

sized to open the valves until the calculation was 

completed as part of the Root Cause of Failure 

Investigation. This, combined with thermal 

binding in the gate valves and marginal nitrogen 

pressure, all contributed to the common-cause 

failure of the DCFWJVs. The failures of the 

vacuum breaker solenoids that caused the reac

tor trip leading up to these failures is described 

in section 8.1.4, below. 

The licensee did not (and did not have to) 

report the DCFWIV failures to the NRC. The 

key safety functions of the DCFWIVs are to 

close upon actuation of a main-steam isolation 

signal (MSIS) and to open upon demand to 

establish a flow path from one of the auxiliary 

feedwater pumps to the steam generators when 

that pump is chosen to make up steam generator 

inventory during periods of hot-standby and 

plant cooldown. These AOVs successfully 

closed on demand from the MSIS, thus satisfy

ing their safety function. However, the AOVs 

failed to open following the cooldown of the 

unit, and as such, the failures were not classified 

as safety-related failures. Also, the failures were 

considered by the licensee as not being

19 NUREG/CR-6654
AOV Operating Experience at the Plants Visited 

Maintenance Rule functional failures. The 

failure mode that occurred was not considered 

by PVNGS to be safety-significant because the 

open function was not credited in the safety 

analysis. The failure was not subject to review 

under the requirements of the Maintenance Rule 

because the system function was not required in 

the plant mode when the failure occurred. 

Further, as noted above, these valves were not 

within the scope of either NRC Generic Letters 

89-10 (these are not MOVs) or 95-07 

(DCFWIVs are normally open valves). 

8.1.4 Failure of a Vacuum Breaker 

Solenoid and Subsequent Inves

tigations Involving SOVs 

On November 26, 1995, a loss of condenser 

vacuum condition occurred at Unit 1 of PVNGS 

and caused a main turbine trip. A reactor trip 

followed shortly thereafter. Refer to 

LER 52895012 and to the failure of the 

DCFWIVs described in paragraph 8.1.3, which 

also resulted from the trip. The event was 

initiated when a condenser vacuum breaker 

inadvertently opened. Two similar SOVs on 

Unit 2 were found to have air leaks. 

The vacuum breakers are non-class, 14-inch 

Pratt butterfly valves equipped with air-to-open, 

spring-to-close actuators. The valves are nor

mally closed and require a 3-way ASCO 8321 

SOV to be energized to pressurize the actuator 

and open each valve. The SOVs are normally 

de-energized and spring force keeps each valve 

closed. 

The vacuum breaker went partially open due 

to a failure of the solenoid valve which allowed 

leakage of supply air to the cylinder and out the 

exhaust port. The exhaust port speed control 

valve was also throttled which allowed a larger 

amount of leaking air to pressurize the cylinder. 

Apparently, the condenser vacuum breaker sole

noid failed because the internal gasket material 

was worn. Other internal parts also showed signs 

of wear including the seating surfaces on the 

piston/guide sub-assembly and on the core 

assembly. The root cause of the SOV failure was 

attributed to aging. Two other SOV leaks were

detected as a result of inspections shortly after 

the event. 

PVNGS evaluated similar SOVs in other 

plant systems to ascertain their susceptibility to 

cause a plant trip, transient, or other system 

operation problem. SOVs that have a high 

potential to cause such problems were evaluated 

further to determine if there were current pre

ventative maintenance (PM) programs to replace 

them. A summary list was prepared of over 

1700 SOVs that were determined to need a new 

PM task generated and/or have the exhaust 

speed control valve removed. 

8.1.5 AOV Margin Calculations 

Among the observations made during the 

visit to PVNGS for this study, it was noted that 

several safety-related AOVs had low margins, 

according to the calculations that were furnished 

for discussion. Specifically: 

Globe valves: The licensee's engineers 

were using the port area times the differ

ential pressure (d/P) times 1.0 in their 

thrust estimate calculations. Instrument 

Society of America guidance is to use the 

port area times d/P times 1.0. Motor-oper

ated valve testing for other types of globe 

valves would indicate a valve factor of 0.9 

to 1.1. Depending on the particular valve 

design, the licensee thrust estimates may 

be as much as 10% low. 

It is important to note that the air-operated 

globe valves used at Palo Verde have not 

been dynamically tested in a Generic 

Letter 89-10 type of program to verify 

valve factors. Also the licensee has only 

found physical valve/actuator perform

ance issues with actuators (for globe 

valves) which were not sized in accor

dance with the original equipment manu

facturer's published recommendations. 

* Gate Valves: The licensee's engineers 

were using the seat area (as provided by 

the valve manufacturer) times d/P times a 

valve factor of 0.6 in their thrust estimate 

calculations. A valve factor of 0.6 is

NUREG/CR-6654 20
AOV Operating Experience at the Plants Visited

reasonable for cold water systems. INEEL 

and EPRI test results both support the use 

of a mean seat area [0.5 X (inside diame

ter + outside diameter)]. 

Palo Verde suggested the above wording 

in their comments on the draft report and 

provided the following explanation as 

justification for the gate valve discussion: 

The Palo Verde engineers had originally 

used the port area for the thrust estimates 

but had revised the calculation to use the 

seat area prior to the site visit. The calcu

lation shown to the authors of this study 

included an error in that some text 

explaining the calculation indicated the 

port area was used rather than the seat 

area. This text was missed as part of the 

calculation revision to use seat area and 

has since been revised. The quantitative 

part of the calculation reviewed by the 

authors actually used the seat area even 

though the text indicated the port area had 

been used. 

Palo Verde initiated several margin 

improvement modifications based on their AOV 

capability calculations. These modifications 

covered 15 Category I and II valves in each unit 

(45 AOVs total). 

The lessons learned in the calculation of 

valve forces and margins in motor-operated 

valves are at least partially pertinent for AOVs. 

The valves used in both AOVs and motor

operated valves are either identical or similar 

and subject to the same design considerations. 

Several licensees visited reported that they used 

the information and techniques developed for 

motor-operated valves to evaluate inadequate 

performance of AOVs or investigate events in

volving AOV failures. It is expected that further 

investigation of AOV performance would reveal 

problems involving sufficient margins similar to 

those found as a result of the studies of oper

ability of motor-operated valves performed in 

response to NRC Bulletin 85-03 and NRC 

Generic Letter 89-10.

8.2 AOV Operating Experi

ence at Enrico Fermi 2, 

Trip No. 2 

8.2.1 Failure of Multiple SOVs Con

trolling Safety-Related AOVs 

Inadequate closing force was discovered on 

four air-operated isolation valves on April 22, 

1994. The review in which the deficient condi

tion was found was prompted by two other 

events in the industry that occurred in 1993. See 

LER 34194004. 

These valves are containment isolation 

valves that isolate the reactor-coolant pressure 

boundary. According to data provided during the 

site visit, they are 3/4-inch Rockwell globe 

valves with Fisher 667, Size 34 air actuators. 

The operators are designed to use air to open and 

springs to close. 

The cause of this event was inadequate con

trol of the valve actuator settings. The original 

specification for these valves was to ensure clo

sure in the control-rod drive hydraulic flow 

direction at a pressure of 1750 psig. This 

resulted in a setting by the vendor sufficiently 

high to meet both the containment isolation as 

well as the reactor coolant pressure boundary 

isolation requirements for these valves. How

ever, the nameplate supplied by the vendor 

incorrectly identified a lower setting. The valves 

were reset during installation to this lower set

ting using the nameplate data. The resulting set

ting was too low to ensure adequate closure of 

the valves for meeting their reactor-coolant pres

sure boundary isolation function. 

Spring preload settings were adjusted to 

ensure valve closure and the design-basis docu

mentation was reviewed and revised. 

LER 34194004 indicates that four similar AOVs 

were discovered by the licensee to have "a 

similar problem." It is not known if or how 

many other similar AOVs at Fermi 2 or in other 

plants have similar incorrect name plate data that

21 NUREG/CR-6654
AOV Operating Experience at the Plants Visited 

could cause plant personnel to set up the AOVs 

incorrectly. This was a potential common-cause 

failure situation. 

8.2.2 Failure of Multiple SOVs Con

trolling Safety-Related AOVs 

Failure of three SOVs due to deposits of a 

mixture of thread locking compound and lubri

cant on working parts was documented in 

Fermi 2 DERs 97-1200 and 97-1202. The 

licensee determined that these events were not 

reportable. Previous problems of this nature on 

eight other SOVs were discussed in a Fermi 2 

internal report dated September 17, 1997 

(enclosed with DER 97-1202). Similar 

(lubricant) contamination problems going back a 

number of years are documented in 

NUREG-1275, Volume 6, "Operating 

Experience Feedback Report - SOV Problems," 

subsection 5.2.4. An acronym found in NUREG

1275, Volume 6, "FUSS" (foreign unidentified 

sticky substance) applies. Similar problems 

recur throughout the industry. For examples, 

refer to LER 26095008 for Browns Ferry, LER 

28688009 for Indian Point 3, LER 32192003 for 

Hatch, LER 37495005 for LaSalle (paragraph 

8.4.2 below), and LER 44090021 for Perry. 

Very recently (early September 1999), Clinton 

reported multiple failures of SOVs used to 

control AOVs, from a problem that appeared to 

be similar based on preliminary information. 

Fermi 2 DER 97-1202 included a compre

hensive "SOV Failure Team Report" that 

described the failures and a subsequent detailed 

investigation of SOVs at Fermi 2. A number of 

actual and potential common-cause failures 

resulted from the thread locking compound 

contamination problem. For example, the report 

indicated that for the normally energized, safety

related ASCO Model 8320 SOVs, there were 

15 failures in a population of 66 normally 

energized valves. The DER additionally reported 

the failure of three normally energized, safety

related ASCO Model 8316 SOVs. Conclusions 

in the report were: 

The common cause of the 18 SOV fail

ures was excessive application of Loctite 

PST-580 thread locking compound.

There were insufficient material controls 

and instructions in place to reduce the 

possibility of (pipe thread sealant and 

valve lubricant) contamination in the 

plants pneumatic system. 

"* Definitions of the categories used by 

Fermi 2 for restrictions on materials used 

to maintain the plant were needed. 

"* NUREG-1275, Volume 6 findings and 

recommendations were not thoroughly 

reviewed for applicability at Fermi 2 or 

compared against their (then) current 

practices. 

"* The corrective and preventive mainte

nance processes did not adequately use 

actual equipment performance data for 

SOVs for determining the correct preven

tive maintenance frequencies. This 

appeared to have been a problem for other 

components, such as pumps, valves, heat 

exchangers, etc., as well. 

To emphasize the common-cause failure 

potential of the above described conditions, the 

population of SOVs at Fermi 2 includes (as of 

the publication of DER 97-1202) 1449 QA-1 

SOVs, of which 134 are ASCO Model 8320. Of 

these 134 SOVs, 66 are normally energized. 

There are also 1040 non-QA-1 SOVs at Fermi 2, 

of which 258 are ASCO Model 8320. Of these 

258 SOVs, approximately 93 are normally 

energized. 

Eliminating the use of thread locking com

pound and substituting metallic (Grafoil) tape 

for sealing and locking was suggested to mini

mize valve contamination problems caused by 

migration of the thread locking materials or resi

dues. However, the Grafoil product was found 

by the licensee to be somewhat awkward to 

apply and difficult to install with a sufficiently 

tight seal because of the thickness and stiffness 

of the Grafoil material. Other thread-locking and 

pipe joint sealing options were being investi

gated at Fermi 2.

NUREG/CR-6654 22
8.3 AOV Operating Experi

ence at Palisades, Trip 

No. 3 

Events at Palisades involving AOVs were 

generally related to moisture/particulate con

tamination in the air systems and these events 

are also discussed briefly in Section 10.1 of this 

study on air systems operating experience. 

8.3.1 Failure of Multiple Air Regula

tors in the High-Pressure Air 

System 

On March 18, 1997, a safety injection tank 

test line redundant high pressure injection isola

tion valve, CV-3018 failed to change position on 

demand. This situation was originally reporting 

to NRC/AEOD in April 1997 but was not cov

ered by an LER. Palisades Condition Report 

C-PAL-97-0404, dated 3/18/97 refers to this 

event. The stroke test was part of a post

maintenance procedure to verify valve operation 

after adjustment of packing. CV-3018 is a 4

inch, 1500 lb. Walworth gate valve with a Miller 

Model DA-63 fail-closed piston air operator. 

High pressure air is supplied to the valve opera

tor through air regulator PCV-3018 and air to 

the valve is controlled by SOV SV-3018. Flow 

rate is controlled by a manual flow control valve 

between the SOV and the actuator. 

During the subsequent investigation, addi

tional attempts were made to stroke the valve 

and it was discovered that the air regulator was 

blocked with dirt and rust contamination from 

the carbon steel piping in the high pressure air 

system. Based on the condition of the air regu

lator, it was concluded that the high pressure air 

system did not meet the cleanliness criteria 

described in the vendor manual for the air regu

lator. A walk-down inspection of other air regu

lators in the system was conducted and 9 of 22 

air regulators were found to be contaminated. 

Significant amounts of material were found in 

three regulators. The restriction orifice was par

tially blocked on these regulators. Additional 

components are considered to have been 

affected, or potentially affected, from this com

mon-cause failure condition.

AOV Operating Experience at the Plants Visited 

A program for regular cleaning of the air 

restrictors in the air regulators, as well as regular 

cleaning of filter screens was initiated. The 

source of the contamination, i.e., the corrosion 

of the air system piping from moisture in the air, 

was not addressed in the licensee's actions. 

8.3.2 Failure of Shutdown Cooling 

Heat Exchanger Outlet Isolation 

Valve 

Two events that occurred at Palisades in 

1978 and 1981, both involving the same type of 

failure of valve CV-3025, were discussed with 

plant personnel during the recent visit as part of 

this study of AOVs. Refer to LERs 25578003 

and 25581030. CV-3025 was found to be inop

erable when called upon to open and allow the 

shutdown cooling heat exchanger to operate. 

Shutdown cooling flow was lost. 

According to section 5.1.1 of NUREG-1275, 

Volume 2, "(o)n both occasions, water in the 1A 

system filled a valve positioner, causing the 

control valve to fail closed. The 1978 event 

lasted for 45 minutes, allowing the primary 

coolant system to heat up from 130'F to 215'F. 

The 1981 event lasted over (one and one-half) 

hours, allowing the primary coolant system to 

heat up from 123 0F to 197'F." 

This valve has a design basis function to 

open during entry into the shutdown cooling 

mode after a small-break loss-of-coolant acci

dent (LOCA). This valve has a function to be 

throttled to adjust cooling flow after a small

break LOCA but the licensee considers this to be 

beyond the design basis. (Refer to Palisades 

System Level Design Basis Review Calculations 

EA-AOVSYS-ESS-01.) CV-3025 is a 10-inch, 

air-operated globe valve. The valve is powered 

from the non-safety-related IA system. 

The event was of interest because failures in 

the shutdown cooling system are not normally 

modeled in risk analyses. The potential loss of 

shutdown cooling merits attention because of the 

potential for uncovering the core in a relatively 

short time if boiling or near-boiling conditions 

occur as a result of the AOV failure.

23 NUREG/CR-6654
AOV Operating Experience at the Plants Visited 

An attempt was made to gather information 

about the condition of the plant during and after 

the events and to assess the current condition of 

the valve. The old records were retrieved by the 

licensee's engineers but detailed information 

about the condition of the reactor coolant system 

(pressure, temperature) and the plant recovery 

procedures at the time of the events couldn't be 

retrieved. 

A hand wheel was added to the AOV at some 

point after the 1981 event to allow the valve to 

be opened manually. However, the condition of 

the air system serving the AOV appeared, at the 

time of the visit, to be similar to the condition in 

1981, when the events occurred. Specifically, 

the potential for contamination of the valve by 

moisture (or corrosion products from moisture 

contamination) occurred because the single 

dryer available for the IA system was, and is, 

periodically bypassed. 

Similar diverse or common-cause failure 

events can occur, affecting important AOVs 

associated with the Shutdown Cooling and 

ECCS systems, because the root cause of these 

events (i.e., substandard air system design, 

operation, and maintenance) persists. 

8.4 AOV Operating Experi

ence at LaSalle 1 and 2, 

Trip No. 4 

8.4.1 Pneumatic Valves with Less

Than-Designed Effective Dia

phragm Areas Result in Inade

quate Valve Closing Forces 

While developing an AOV preventative 

maintenance program, inconsistent testing data 

were obtained for valves with WKM 70-13-1 

pneumatic actuators. The inconsistent results 

appeared to be related to incorrect effective dia

phragm areas (EDAs) for the AOV actuators. 

This occurred in the February-April 1996 time 

frame. Refer to LER 37396011. Two problems 

associated with the EDAs of the actuators of the 

WKM valves were identified. The first was 

related to the actual versus the manufacturer's 

published EDAs of the actuator. The second

problem was stretching of the diaphragm during 

valve travel resulting in a reduced EDA. In 

March 1996, LaSalle Station's AOV Component 

Engineer contacted Anchor/Darling Valve Com

pany regarding the published versus the actual 

EDAs and the stretching of the diaphragms. On 

September 20, 1996, after conducting testing, 

Anchor/Darling Valve Company acknowledged 

a reduction of the EDAs for the WKM 70-13-1 

pneumatic operators. 

Anchor/Darling performed a series of tests to 

determine the actual effective diaphragm area. 

These tests indicated that the actual diaphragm 

areas of the various sizes of the Model 70-13 

actuators were approximately 90% of the pub

lished values. Further testing by the licensee 

uncovered a contributing problem that further 

reduced the effective diaphragm area. The dia

phragm case consists of two dome-shape halves 

bolted together. Generally, one haft is deeper 

than the other. It was discovered that 

reverse-acting actuators assembled with a deep 

upper half case caused unintended stretching of 

the diaphragm within the casing. 

The licensee determined that the primary 

containment isolation valves would have closed 

at the design-basis-accident containment pres

sure of 40 pounds per square inch; however, 

many of the valves may not have been properly 

set up to close against the normally higher sys

tem pressure under some operating conditions. 

That is, the valves may not have closed under 

the highest expected differential pressure of the 

contained system fluid. 

There were a total of 36 (18 per unit) WKM 

AOVs addressed in the LER. Thirteen valves per 

unit are installed in systems which are part of the 

Primary Containment Isolation System (PCIS) 

and five valves per unit are in the Reactor Core 

Isolation Cooling System (RCIC). 

This condition was originally determined by 

the licensee to be not reportable but the LER 

was submitted as a voluntary report. 

The root cause of this event was the use of 

incorrect effective diaphragm areas by the

NUREG/CR-6654 24
AOV Operating Experience at the Plants Visited

Original Equipment Manufacturer (OEM) for 

actuator setup. The OEM at the time was WKM. 

LaSalle prepared and implemented design 

change packages (DCPs) to compensate for the 

reduced EDA and restore the valves to the 

design specifications. These DCPs were sched

uled to be completed prior to Unit 1 and Unit 2 

startup from the (then) recent outages. A 10 CFR 

Part 21 notification on the deficiency of WKM 

70-13-1 Pneumatic Actuators currently manu

factured by the Anchor/Darling Valve Company 

was issued on October 4, 1996. NRC Informa

tion Notice 96-68, Incorrect Effective Dia

phragm Area Values in Vendor Manual Result 

in Potential Failure of Pneumatic Diaphragm 

Actuators, was issued on December 19, 1996. 

The common-cause failure implications of the 

deficiency involve a whole class of AOVs, 

widely used in nuclear power plants in both 

safety-related and important non-safety related 

applications, the full scope of which has not yet 

been determined. 

8.4.2 Failure of MSIVs Due to Sticking 

Solenoid Operated Pilot Valves 

Unit 2 was shutting down for refueling on 

February 18, 1995, when a main steam isolation 

valve (MSIV) failed to close on signal. Refer to 

LER 37495005 and NRC Information 

Notice 95-53. The cause of the MISV's failure 

to close was later determined to be sticking of a 

solenoid pilot valve. The solenoid pilot valve did 

not change position, and thus did not allow the 

pilot air to vent. Failure of pilot air to vent 

results in air being ported to the under side of 

the MSIV operating piston, thus holding the 

MISV in the open position. Shortly thereafter, 

another MSIV failed to close under similar 

circumstances. The two Outboard MISVs, which 

failed to close, were in different Main Steam 

Lines; however, the Inboard MS1Vs in each of 

these lines operated as designed. 

The sticking solenoid pilot valves were dis

assembled and inspected. Foreign material was 

observed on several internal parts of the solenoid 

pilot valve. It was also observed that the 

interfacing surface of the core assembly and 

plug nut of one solenoid appeared to have a thin

coat of foreign material. When the core 

assembly and plug nut were pressed together, 

which is the normal configuration when the 

solenoid is energized, the film acted as an 

adhesive. This adhesive material was strong 

enough that when the plug nut was lifted from 

the bench, the core assembly adhered to it and 

was also lifted. 

The parts with foreign material were chemi

cally analyzed by the System Material Analysis 

Department, and the material was determined to 

be Nyogel 775A. According to the SOV manu

facturer, this material is used during assembly as 

a lubricant on the solenoid cover, under the star 

washer on the disc holder which is internal to the 

valve, and at the interface of the solenoid base 

and housing assembly. Corrective actions 

included replacement of the SOVs in the MSIV 

with SOVs from another manufacturer, disas

sembly and inspection of spare SOVs, and exer

cising the SOVs in the MSIVs at 30-day inter

vals in order to minimize the potential for 

sticking. 

These recurring incidents of SOV contami

nation from lubricants and/or thread locking 

compounds, described in 1990 in NUREG-1275, 

Volume 6, and in paragraph 8.2.2 above, con

tinue to be sources of common-cause failures in 

large numbers of SOVs and their associated 

AOVs throughout the nuclear industry. 

8.4.3 Potential Failure of Safety

Related Dampers Due to Invalid 

Calculation Assumptions in the 

Original Builders Calculations 

NRC Event No. 33434, dated December 19, 

1997, "Inadequate Turbine Building Vent Sys

tem Exhaust Tunnel/HPCS System Switchgear 

Room Wall," and a resulting LER 37397046, 

"Potential Pressurization of the Turbine Building 

Ventilation Exhaust Tunnel Resulting from a 

Main Steam HELB due to Calculation Error," 

refer to a transient analysis performed to predict 

pressures in the turbine building ventilation 

exhaust tunnel downstream of a postulated high 

energy line break. An investigation of the 

allowable closure time assumptions for these 

Turbine Building High-energy Line Break

25 NUREG/CR-6654
AOV Operating Experience at the Plants Visited 

Check Dampers (# I/2VT79YA/B/C) uncovered 

discrepancies in the original 1977 builder's 

calculations and assumptions. In the event of a 

Main Steam High Energy Line Break (HELB), 

the Turbine Building Ventilation Isolation 

Dampers, l(2)VT079YA, B & C, would not 

close fast enough to prevent the pressure from 

exceeding the pressure retaining capability of 

the walls, floors, and ceilings that separate the 

exhaust tunnel from the safety related High 

Pressure Core Spray (HIPCS) electrical 

switchgear room. Neither Event Report 33434 

nor LER 37397046 specifically described the 

operating mechanism of the dampers or 

indicated if the dampers were considered by the 

licensee to be safety-related. 

The design discrepancies led the licensee to 

reconsider the original builder's assumptions 

and calculations for the Reactor Building 

Exhaust Isolation Dampers (1VR05YA and B). 

LER 37398007 was generated based on the 

licensee's investigation. It was found that the 

originally assumed closure time for 1VR05YA 

and B would not be quick enough to prevent 

over-pressurization of the Reactor Building 

exhaust ductwork and downstream masonry ple

num. The concern was that there was 

safety-related Control Room HVAC equipment 

immediately adjacent to the plenum's masonry 

walls, which could be damaged by failure of the 

walls from sudden over-pressurization. Although 

it is not specifically stated in LER 37398007, it 

is assumed that dampers 1VR05YA and B and 

their associated SOVs are safety-related 

components. 

Dampers 1VR05YA and B had malfunc

tioned and were subsequently rebuilt in 1985 

(see LER37385008 and LER 37385011). 

Instrument air keeps the dampers open during 

normal plant operation. A high pressure signal 

should de-energize the SOVs in the event of a 

main steam line break, allowing IA to bleed off 

and the damper to close. In LER 37398007, the 

licensee indicated that they had revised the cal

culations and assumed a 0.4 second instrument 

time delay, and a 0.075 second solenoid valve 

response time, to correct the original assumption 

of instantaneous closure.

8.5 AOV Operating Experi

ence at Three Mile 

Island 1, Trip No. 5 

8.5.1 Insufficient Design Margins in 

Aloyco Air-Operated Gate Valves 

Potentially insufficient margins in two AOVs 

were described by the TMI engineers during the 

visit for this AOV study. Calculations for five 

Crane-Aloyco, 2.5 to 6 inch, 1500 and 

150 pound class, flex-wedge and split wedge 

gate AOVs with Miller DA-63-B and A-63-B 

cylinder actuators were reviewed by one of the 

TMI engineers. The architect/engineer (AE) 

requested that the valve manufacturer (Crane

Aloyco) perform thrust calculations on the five 

AOVs, as part of a limit switch upgrade 

modification, in order to verify that limit switch 

installation would not affect valve operability. 

The resultant thrust calculations, using "present 

day" methodology, indicated that two of the five 

valves had negative closing margins for the 

specified differential pressure (d/P). The AE 

then requested that the manufacturer re-perform 

the calculations using the methodology by which 

the valves were originally sized. The revised 

Crane-Aloyco calculations, based on the original 

valve sizing methodology, indicated positive 

margins in both the opening and closing 

directions, but using a valve factor of zero. Upon 

review of the Crane-Aloyco calculations, TMI 

convened a review group which verified that the 

two containment isolation valves which had 

negative margins in the manufacturers' first 

calculations were, in fact, operable and would be 

able to perform their designed safety function 

because the required operating conditions were 

less severe than the original design basis 

conditions. This conclusion was based on TMI's 

calculations using a 0.75 friction factor and d/P 

of 1600 psi (that required for containment 

isolation). 

A design modification was proposed to 

increase the closing margin for one of the valves 

because normal operating differential pressure 

(2375 psi) was greater than the differential pres

sure encountered when performing its safety 

function (1600 psi). The proposed modification

NUREG/CR-6654 26
included the installation of an accumulator and 

piping to provide air assist to the spring. Subse

quently, TMI performed an analysis of the valve 

using the EPRI PPM methodology (2375 psi and 

0.67 friction factor) and confirmed the operabil

ity of the valve. Design problems with a pressure 

booster for the AOV were also noted. Originally, 

these conditions were not covered by an LER or 

other correspondence with the NRC. 

During or shortly after the plant visit for this 

AOV study, the TMI engineers agreed to inves

tigate to determine if there were 10 CFR Part 21 

concerns to be addressed, either by TMI or the 

valve manufacturer (Crane-Aloyco). They 

determined that the conditions were not report

able under 10 CFR Part 21 for the valves 

because 10 CFR Part 21 had not been involved 

in the original purchase documents. However, as 

described in the next paragraph, there may be 

generic or common-cause issues with these 

types of AOVs, related to the design or the OEM 

calculations, that need to be addressed. 

It was learned that Crystal River 3 (a similar 

B&W plant with the same AE, Gilbert Associ

ates, as TMI-1) had a similar margin problem, 

with a Crane-Aloyco AOV, as reported in 

LER 30297015. Crystal River previously 

encountered design margin problems with Crane 

motor-operated valves due to degradation caused 

by normal wear in the valves, as reported by 

LER 30292004. For the configuration at Crystal 

River 3, the licensee replaced the AOV's 

actuator with a larger one. 

8.6 AOV Operating Experi

ence at Indian Point 3, 

Trip No. 6 

8.6.1 Failure of Diaphragms in AOVs 

Recent events that occurred at Dresden and 

Quad Cities, involving potential failure of AOV 

diaphragms on Copes-Vulcan D100 valve 

operators, were discussed with the maintenance 

engineers at IP3 during the visit incidental to 

this study. See Dresden LER 23798003, Morn

ing Report H-98-0045 dated March 6, 1998, and 

10 CFR 50.72 Report 33620 of January 28,

AOV Operating Experience at the Plants Visited 

1998. The problem was that the elastomer cov

ering the fibers in the diaphragms was too thin 

and the diaphragms wore out prematurely as the 

valves were operated. 1P3 engineers had not 

been aware of this problem and, since they have 

a large number of this type of valve, they started 

to investigate immediately. 

Copes-Vulcan modified the diaphragm 

design in 1996 to increase the amount of rubber 

on the side of the diaphragm in contact with the 

diaphragm plate. Licensees have to determine if 

diaphragms in the AOVs at their plants have a 

deficient design such as that described. 

A search of the nuclear industry database 

indicated that there are over 1900 AOVs of this 

type in U.S. power plants. IP3 has about 90 such 

AOVs. About 18 failures that involve the dia

phragms of these actuators were recognized to 

be premature failures and reported as such in the 

industry database. It is not known how many 

diaphragms having this deficiency were replaced 

for normal wear and not recognized to have 

experienced premature failures. 

Refer to Sections 9.2 and 9.4 in this study for 

descriptions of other design problems related to 

Copes-Vulcan AOVs. 

8.6.2 Two Containment Isolation, In

Series AOVs were Inoperable 

Due to Inadequate Design 

On February 15, 1996, two air-operated dia

phragm type vapor containment isolation valves 

in series were found to be inoperable, violating 

plant Technical Specifications (LER 28696004). 

The original isolation valve design would close 

at system pressure when there was a differential 

pressure or accident design pressure, but not 

against system pressure with no pressure differ

ential. The cause of this inadequacy in the origi

nal design was not determined and the condition 

was considered by the licensee to have not 

required any corrective action (although the 

valve operators and SOVs were subsequently 

replaced). This was a common-cause event. 

These two AOVs (RC-AOV-519 and 

RC AOV-552) are vapor containment isolation

27 NUREG/CR-6654
AOV Operating Experience at the Plants Visited 

valves on the 3-inch primary water system line 

to supply spray water to the pressurizer relief 

tank and make-up water to the reactor coolant 

pump stand pipes. Both RC-AOV-519 and 

RC-AOV-552 are 3-inch air operated diaphragm 

valves manufactured by the ITI Grinnel Com

pany. The valves are normally closed during 

operation but can be opened when required since 

they are supposed to automatically close on a 

containment isolation signal. The engineers 

determined that both valves behaved radically 

different, depending on whether or not the 

system was pressurized or depressurized. 

The valve vendor explained that when 

designing a diaphragm valve/actuator, one must 

consider whether it closes with a 100% differen

tial pressure across the valve (i.e., valve closed 

with little or no downstream pressure), a 0% 

differential pressure (i.e., valve closed with a 

constant line pressure upstream and down

stream) or for both conditions. RC-AOV-519 

and RC-AOV-552 are designed to positively 

seal with a differential pressure and would close 

against a differential pressure of 150 psi (the 

primary water system design pressure) but 

would not close with a line pressure greater than 

about 120 psig when there is no differential 

pressure. The original specification for these 

valves was for a maximum differential pressure 

of 200 psi, with no reference to a minimum 

pressure differential or constant line pressure 

requirement. The valve design had not been 

modified since the plant was built. 

The condition could have existed when 

closing isolation valve RC-AOV-560, down

stream and in series with RC-AOV-519 and 552, 

in a sequence that maintains a line pressure 

greater than 12 psi (probably a typographical 

mistake in LER 28696004 and should be 

120 psi). There would then be no differential 

pressure during stroke testing. Stroke testing 

under these conditions could have created 

another problem if the limit switches were 

adjusted after the valves were to have been shut 

but did not fully close. The valves could have 

been assumed to be fully closed during the 

adjustment so subsequent failures to close would 

be masked and could have allowed the existence 

of a spurious control room indication that the

valves were fully closed when they were not. 

There is no direct external indication on the 

valves to show if they are fully closed. 

Tests were conducted that verified the condi

tion described above. Other valves were sus

pected of having the same problem but subse

quent investigation indicated that these two 

valves were the only ones at IP3 subject to the 

differential pressure problem described. 

The operators and air-supply solenoids for 

these valves were replaced with re-sized 

components so that the AOVs meet design 

requirements. 

8.7 AOV Operating Experi

ence at Turkey Point 3 

and 4, Trip No. 7 

8.7.1 Failure of Auxiliary Feedwater 

System AOVs from Air System 

Contamination 

Turkey Point Units 3 and 4 experienced 

recurring common-cause failures in AOVs, pres

sure transmitters, and pressure controllers in the 

Auxiliary Feedwater (AFW) systems for both 

units during surveillance testing in July 1985. 

The failures were traced directly to IA system 

moisture and corrosion product contamination 

and involved simultaneous failure of the AFW 

flow control valves and steam generator bypass 

valves. 

The event involved the loss of all three trains 

of AFW (one train for each unit and a swing 

train). This event was categorized as a precursor 

event in the NRC's Accident Sequence Precur

sor (ASP) program. The conditional core dam

age probability for this event was estimated in 

the ASP analysis to be approximately 9E-04, 

making it the fourth highest of the 40 precursor 

events identified that year. 

Other safety-related systems were also sup

plied with contaminated air during this period 

(prior to correction) and could have also been 

adversely affected. Functions, other than AFW, 

which could have been affected (simultaneously) 

were steam dump to atmosphere, salt water sysNUREG/CR-6654 28
tem flow from the essential heat exchanger, the 

charging system, and the residual heat removal 

system. 

These conditions, the events leading to them, 

and the potential consequences were described 

in detail in Section 5.1.2 of NUREG-1275, Vol

ume 2. LER 25085021, as well as several NRC 

Inspection Reports (referenced in NUREG-1275, 

Volume 2), also described these events and 

conditions. 

The licensee installed effective drying 

equipment and filters, along with continuous 

dew point monitoring equipment to correct these 

problems. See Section 10.1 in this report. 

Although the conditions have since been cor

rected by dramatic improvement of instrument 

air quality, they are noted here because of their 

severity and (then) potential consequences. 

8.7.2 Pilot Lockup Valves (POLVs) for 

Emergency Containment Cooler 

(ECC) Valves 

One failure of a POLV (CV-4-2908) to oper

ate was confirmed in 1996. As-found exaniina

tion of other POLVs subsequently removed from 

service indicated that 3 of 12 were stuck. Several 

contributors to failure including 0-ring distor

tion and grease caking were identified as causing 

excessive drag. Increased exercise and spring 

modifications were implemented as fixes. Tur

key Point Condition Report 96-0535 dated 

April 29, 1996, Supplement 1, dated May 16, 

1996, Supplement 2, dated September 16, 1996, 

Supplement 3, dated February 7, 1997, and a 

summary report, "Investigation of Increased 

Drag Forces in Pilot Operated Lock-Up Valves," 

prepared by Altran Corp., dated September 

1996, described these conditions; however, an 

LER could not be found. 

The function of the POLVs is to open the 

emergency core cooling (ECC) outlet valves 

when air pressure drops to about 45 psig, Unit 4, 

or 60 psig, Unit 3. The ECC outlet valves are 

intended to move to the open position on loss of 

IA, because the actuators require air pressure to

AOV Operating Experience at the Plants Visited 

function and the IA air system is not safety 

related. Failure of the POLV to shift on loss of 

IA pressure could result in insufficient compo

nent cooling water flow to support the contain

ment temperature/pressure control design basis 

safety function. Therefore, the POLVs are con

sidered safety-related. According to the sum

mary report of September 1996, the POLVs (at 

least those tested) are ITT Hammel Dahl Cono

flow Model GVB-12. 

The tendency of some of the POLVs to stick 

was attributed to several factors, including: 

"* POLVs were originally equipped with 

Buna-N 0-rings. Several POLVs were 

refitted with Viton 0-rings in an attempt 

to increase their service life; however, it 

was subsequently discovered that the 

Viton material is subject to compression 

set. The detailed discussions in the Con

dition Report and it supplements included 

information and conclusions about the 

various complex formulations, tempera

ture considerations, and interactions of the 

elastomers with various lubricants. In the 

end, it was decided to install stiffer 

springs to overcome any tendency of the 

valves to stick and reinstall Buna-N 

0-rings. 

"* The silicone grease used to lubricate the 

valves was found to separate and lose 

lubrication capacity in static applications. 

This was considered to contribute to the 

sticking problem. 

" The valve's tendency to stick was 

observed to increase over time. A sur

veillance interval of 31 days was estab

lished, based on the results of the 

investigation. 

8.7.3 Intake Cooling Water Header 

Inlet Isolation Valve Failed to 

Fully Close 

The Intake Cooling Water (ICW) Header 

Inlet Isolation Valve (POV-3-4882) failed in 

approximately 1/4 closed position. Corrosion

29 NUREG/CR-6654
AOV Operating Experience at the Plants Visited 

was found on the lower portion of the yoke and 

housing below the 0-ring seal. The corrosion, 

caused by lack of maintenance, was identified as 

the root cause of failure. Minor wear was also 

found in the housing and cover guides. 

POV-3-4882 was described by the licensee 

as a risk significant, key component and, there

fore, Maintenance Rule "a(1)" goals were estab

lished based on this event. Failure of this valve 

to fully close was a functional failure because it 

prevented the valve from automatically closing 

on receipt of a safety injection signal and 

diverting ICW flow from the turbine plant cool

ing water heat exchangers to the safety-related 

component cooling water heat exchangers. 

However, the licensee did not consider the 

problems to be reportable under 10 CFR 50.72 

or 50.73 because no LCO was exceeded. The 

plant engineers prepared a Condition Report 

(CR 96-0304) which was reviewed during and 

shortly after the visit to the site for this AOV 

study. 

A grease fitting was installed to enable relu

brication of the yoke-to-housing interface. 

POV-3-4883, POV-4-4882, and POV-4-4883 

were also overhauled. 

8.7.4 RCS Letdown Isolation Valve 

Failure 

CV-3-204, a safety-related letdown isolation 

AOV, failed to close when remotely operated 

from the control room on September 25, 1996. 

The active safety-related function of the SOV 

(SV-3-204) for this AOV is to vent when de

energized to enable CV-3-204 to isolate. The 

SOV was determined to be defective and was 

replaced, but no definitive root cause was identi

fied for the SOV failure. A small amount of 

Teflon tape was found in the body of the SOV. 

Condition Report 96-1202 refers to this event 

but no corresponding LER could be found. 

Condition Report 96-1202 contains the results 

of a failure history review for ASCO 8316 SOVs 

at Turkey Point. A population of 168 SOVs was

listed. The licensee concluded that the failure 

history indicated that these SOVs operated 

reliably. However, it was noted that 27% of these 

SOVs had been replaced, for all reasons including 

both planned maintenance and failure, in the 

previous 10 years. 

8.7.5 Revised Diaphragm Areas Fur

nished by AOV Vendor were 

Lower than Designed 

Based on the report from LaSalle (see Sec

tion 8.4.1) that a valve vendor had furnished 

AOVs with incorrect effective diaphragm areas, 

Turkey Point engineers investigated their own 

(similar) AOVs, as described in their Condition 

Report 96-1598 of December 17, 1996. The 

vendor of the actuators at Turkey Point, BS&B, 

was contacted and it was discovered that 

28 AOVs, divided among the two units, had 

effective diaphragm areas lower than originally 

designed by as much as 10%. 

The engineers at Turkey Point use the effec

tive diaphragm area to establish bench set, lift

off, and supply regulator settings for AOVs. The 

lower actual diaphragm areas would (poten

tially) affect the leak tightness and ability to 

fully stroke the valves. BS&B actuators are used 

on the emergency diesel generator inlet control 

valves, the instrument air control valves for the 

containment header, the bypass control valves 

for component cooling water from the emer

gency containment cooler, the main steam to 

main steam relief control valves, the atmos

pheric dump valves, and the containment sump 

pump discharge isolation valves. 

The licensee's subsequent operability 

assessment indicated that there was no safety 

concern and that all of the AOVs would perform 

their functions within the existing operational 

parameters. Although, in this case, the AOVs 

had sufficient margin to perform their intended 

function, the actual design margins had been 

unknown for some time. 

These examples of mistakes in vendor cal

culations are a common-cause failure concern.

NUREG/CR-6654 30
9. OTHER PERTINENT AOV EVENTS 

AND OPERATING EXPERIENCE

The NRC described many deficiencies in 

AOVs in their Bulletins, Circulars, Generic 

Letters and Notices. Several of these publica

tions, applicable to the discussion of the capa

bility of AOVs to perform their intended func

tions, are discussed in this section. Operating 

experience involving air-operated or solenoid

operated dampers, as well as AOVs or SOVs 

that serve diesel generators, are also discussed in 

Section 11. The focus of the following discus

sions was primarily on the design and qualifica

tion problems of AOVs, as opposed to mainte

nance or air system problems. 

NRC Inspection Reports and Immediate

Notification Reports (i.e., reports covered by 

10 CFR 50.72) were consulted. In addition, 

searches were conducted in the NRC's Licensee 

Event Report (LER) database for pertinent 

events since 1985 involving AOVs. LERs and 

events described in other reports that were con

sidered of interest in this study of AOVs are 

listed in Tables 2, 3, and 4. Table 7 includes 

some recent (approximately last 5 years) events 

and conditions involving AOVs and/or air

operated components where the design basis was 

not met and/or not known. Tables 6 and 7 were 

included in response to comments on the draft of 

this report dated April 26, 1999, concerning the 

number of AOVs that are risk significant and the 

timeliness and pertinence of events presented in 

the draft report, respectively. LERs that were 

cited in Tables 2, 3 and 7 are also listed in 

numerical order in Table 5 to provide a conven

ient cross-reference for the reader. Pertinent 

NRC generic communications regarding AOVs, 

SOVs, dampers and air systems are listed in 

Appendix A. 

There were very large numbers of pertinent 

LERs to be reviewed regarding AOVs, dampers, 

SOVs and air systems. Word searches in the 

NRC's NUDOCS database, the NRC's Public 

Document Room (BRS) database, and the 

Sequence Code Search System (SCSS) Database 

were conducted. Many key-word searches in 

NUDOCS and the BRS databases commonly

returned hit lists in the high hundreds but the 

events in these lists were somewhat difficult to 

retrieve because of limitations of the on-line 

systems. It is expected that many interesting and 

pertinent events and failures were not included. 

Listings of failures and maintenance data 

were collected from the nuclear industry's data

base (Nuclear Plant Reliability Data System 

[NPRDS] and Equipment Performance and 

Information Exchange [EPIX]), and a prelimi

nary attempt was made to incorporate the data in 

a spreadsheet format. However, the effort was 

abandoned and this data was not used in this 

study because insufficient time was available to 

devise a workable sorting scheme, and then sort 

and evaluate an enormous volume of informa

tion in order to determine if and how the data 

was pertinent. 

9.1 Undersized AOV 

Acuators 

In NRC Information Notice 88-94, dated 

December 2, 1988, the NRC described actuators 

for Fisher AOVs that were undersized and 

potentially incapable of providing sufficient 

thrust to seat, unseat, or properly operate for 

some design conditions. The problem was traced 

to a design change where graphite stem packing 

was substituted for Teflon stem packing. The 

design calculations did not account for the 

increased packing drag of the graphite. 

Fisher indicated that until the mid-1970s it 

had supplied sliding stem valves with Fisher's 

standard single arrangement Teflon packing, and 

its actuator sizing technology was principally 

based on Teflon packing. Fisher did not account 

for the Teflon packing friction forces for sizing 

actuators with Teflon-packed valves because, 

based on their experience, the friction forces 

were calculated to be small compared to the 

actuator air pressure forces, actuator spring 

forces, valve differential pressure forces, and 

valve seating forces. About 1975, Fisher began 

supplying a number of valves with graphite

31 NUREG/CR-6654
Other Pertinent AOV Events 

and Operating Experience

laminate packing. This change was based on 

nuclear power industry requests to eliminate 

Teflon material from radiation environments and 

because of general personal health and safety 

concerns resulting in the replacement of asbestos 

packing material. In April 1976, Fisher began to 

explicitly account for packing friction forces in 

sizing all valve actuators after they discovered 

that graphite packing friction forces and other 

non-Teflon packing materials contributed sig

nificantly to the overall friction forces. The 

change was fully implemented for all Fisher 

orders shipped after January 1, 1977. 

Fisher also stated that some nuclear power 

plant licensees may have installed packing dif

ferent from that provided in the original valve. 

Such a change would be of concern if the new 

packing creates higher friction forces than the 

original packing. For example, data provided by 

Fisher indicated that increased friction forces 

can result if Teflon packing is replaced with 

graphite packing or graphite laminate packing is 

replaced with graphite ribbon packing. If the 

size of the installed actuator cannot overcome 

the increased friction forces associated with the 

packing change, then the valve may be incapable 

of performing some of its intended functions. 

In addition, Fisher informed the NRC of the 

potential for undersized actuators on Fisher 9200 

series butterfly valves. These particular valves 

employ rotary shafts rather than sliding stems 

and are not subject to the same packing friction 

concerns previously discussed; however, some 

of these butterfly valves may be equipped with 

undersized actuators for a different reason. 

Fisher sized butterfly valve actuators ordered 

before March 1, 1982, used a method that under 

some circumstances underestimated the torque 

required to seat or unseat the butterfly disc in the 

9200 series valves. Butterfly valves ordered after 

March 1, 1982, were not subject to this concern 

because they were equipped with actuators sized 

by the current Fisher method that more accu

rately predicts valve seat/disc frictional forces.

The following AOVs were subject to the 

design deficiency described above: 

Sliding stem valves supplied by any 

manufacturer that were repacked using 

materials or procedures that increased the 

packing friction forces beyond those 

accounted for in sizing the actuators. 

* Fisher Controls sliding stem valves 

shipped before January 1, 1977, supplied 

with graphite and other non-Teflon pack

ing. The actuators for these valves were 

sized by Fisher without accounting for 

packing friction forces. Actuators for 

sliding stem valves supplied by other 

manufacturers may also be undersized, 

depending on the actuator sizing methods 

used by these manufacturers. 

Fisher Controls 9200 series butterfly 

valves ordered before March 1, 1982. The 

method used to size the actuators for these 

valves may have underestimated the 

torque needed to seat or unseat the butter

fly disc. 

9.2 Weight and Center-of

Gravity Discrepancies in 

AOVs 

NRC Information Notices 89-28 of March 

14, 1989, and 90-17 of March 8, 1990, described 

discrepancies in the information on weights and 

center-of-gravity (CG) provided by Copes

Vulcan to a large number of nuclear plants. 

Numerous types, sizes, and classes of AOVs (as 

well as other valve types) in various services 

were involved. 

Westinghouse Electric Corporation and 

Copes-Vulcan prepared an extensive set of 

revised valve drawings showing the corrected 

weights and centers of gravity. Westinghouse 

compiled these drawings into valve lists that

NUREG/CR-6654 32
Other Pertinent AOV Events 

and Operating Experience

represented a complete set of correct weights 

and centers of gravity as was then currently 

known to them. However, these valves repre

sented only a portion of the total population of 

valves that Westinghouse had supplied to vari

ous licensees. Tables included in the Information 

Notices provided condensed information to 

licensees of the extent of the problem including 

a brief description of the type and class of the 

valves most significantly affected by these 

corrections. 

In the information provided by Westinghouse 

and Copes-Vulcan, approximately 16 models of 

Copes-Vulcan valves were noted as having 

weight increases greater than 25%. The largest 

individual increase was 1710 pounds on a 

16-inch, Class 900, air-operated feedwater 

valve. This represented an increase of 83% in 

the weight originally given for the valve. Other 

notable weight increases included a 515-pound 

weight increase for a 10-inch, Class 600, air

operated modulating valve, and a 107-pound 

weight increase for a 2-inch, Class 1500, air

operated modulating valve. 

The center of gravity discrepancies were far 

more extensive than the weight discrepancies. 

Approximately 32 models of Copes-Vulcan 

valves were noted as having potentially signifi

cant discrepancies of this type. In some cases, 

the center of gravity information was not origi

nally supplied along with the valve. In those 

instances, piping design analyses were typically 

performed using conservative estimates. How

ever, in a large number of cases the original 

information provided with the valves was non

conservative by more than 20%. In the worst 

case, the center of gravity on a one-inch, 

Class 600, air-operated isolation valve changed 

by 15 inches. The staff also concluded that these 

weight and center of gravity discrepancies were 

not unique to Westinghouse plants. 

Several licensees reanalyzed the associated 

piping systems using corrected small bore valve 

weight and center of gravity information. The 

results of these initial reanalyses disclosed 

instances where the calculated stress would 

exceed 100,000 psi during a design-basis

earthquake. In all cases the acceptability of the 

installation was confirmed, but in several cases 

use of more sophisticated time-history type 

analyses and NRC-approved increases in allow

able stresses were necessary. In addition, some 

modifications to the seismic support configura

tions were required in order to bring the instal

lations within the allowable FSAR stress criteria. 

This information is included in this current 

study of AOVs to make the point that the quality 

of design information provided by vendors has 

been questionable concerning several important 

aspects in the design of AOVs. 

9.3 Failure of Air Actuators 

Upon Gradual Loss of Air 

Pressure 

The NRC published Information 

Notice 82-25 on July 20, 1982, to alert licensees 

of the potential for certain Hiller actuators to fail 

in "non-safe" positions if the air supply is 

gradually decreased. 

Mississippi Power and Light Company 

reported that a large number of isolation valves 

in the instrument air system at Grand Gulf 

Nuclear Station (GGNS) failed to pass test 

requirements. The valves were supplied by the 

William Powell Company and equipped with 

actuators supplied by the Ralph A. Hiller Com

pany. Various plant systems were affected, pri

marily those associated with containment 

isolation. 

During pre-operational testing designed to 

simulate a slow loss of air in accordance with 

position C.9 of Regulatory Guide 1.80, a large 

number of pneumatically operated valves failed 

to go to their fail-safe condition when the 

instrument air header was slowly depressurized. 

Additional testing to simulate an air-line break 

in accordance with position C.8 was accom

plished by depressurizing the instrument header 

supplying the containment, drywell, and auxil

iary building from operating pressure (110 psig) 

to atmospheric pressure in one minute. Forty

eight valves failed this test.

33 NUREG/CR-6654
Other Pertinent AOV Events 

and Operating Experience

The specifications for the actuators required 

the valve to fail to a specified position upon a 

loss of instrument air but did not specify the rate 

of depressurization. The Hiller actuators would 

operate in the specified manner only if the 

actuator itself was promptly depressurized. The 

pneumatic actuators include an accumulator with 

stored air. The stored air is transferred to the 

actuator cylinder to stroke the valve by means of 

pneumatic-operated selector valves controlled 

by (non-safety-related) instrument air. Upon a 

gradual loss of instrument air pressure, the 

selector valves will bleed the accumulator air to 

the atmosphere rather than to the actuator cylin

der. This occurs near 20 psig when the selector 

valve plunger is in an intermediate position. 

Mississippi Power and Light Company added 

safety-related pressure switches to sense air 

supply pressure to the valves. When the supply 

air pressure drops to a point slightly above that 

at which the actuator selector valve would begin 

to move and bleed off the accumulator, the pres

sure switch de-energizes the solenoid pilot and 

immediately causes the valve to go to the fail

safe position. 

The instrument air system at Grand Gulf is 

not seismic Category 1; therefore, a line break 

causing a rapid loss of instrument air is a realis

tic concern. Had the condition simulated in the 

tests (i.e., instrument air-line break) occurred 

coincident with a postulated loss of coolant 

accident, then the failure of the pneumatic 

valves to go to their fail-safe position could have 

resulted in a loss of containment integrity. 

On September 10, 1992, the NRC issued 

Information Notice 92-67 to alert licensees to 

problems that might occur as a result of their 

licensee modifications, made to address poten

tial failure of Hiller actuators upon a gradual 

loss of air pressure. 

Valve assemblies for three Shearon Harris 

Main Feedwater Preheater Bypass Isolation 

Valves were specified, procured, and installed 

for Q Class application. The Anchor Darling 

Valve Company supplied the valves and the 

associated Hiller actuators. On January 7, 1992, 

Carolina Power and Light Company discovered

several components associated with the air sup

ply to the actuators of the three main feedwater 

preheater bypass isolation valves were not quali

fied for a Q Class application. Specifically, the 

failure of the air pump in the non-Q Class, non

seismic instrument air supply to the valve 

actuator accumulator could prevent pressure 

switches upstream of the air pump from detect

ing slow leakage in the Q Class, seismic portion 

of the actuator air lines. The pressure switches 

were installed to ensure valve closure by sending 

an automatic close signal if the instrument air 

system pressure (upstream of the actuator air 

pump) dropped to 66 psig as discussed in Infor

mation Notice 82-25. 

The Main Feedwater Preheater Bypass Isola

tion Valves at Shearon Harris function as con

tainment isolation valves upon receipt of a 

feedwater isolation signal. The function of the 

air pump is to raise the normal instrument air 

supply pressure from between 70 and 100 psig 

to approximately 150 psig. If accumulator pres

sure drops from 150 psig to 122 psig, the main 

feedwater preheater bypass isolation valve may 

not close within 10 seconds. If pressure drops to 

a value as low as 20 psig, the air pressure may 

not be sufficient to close the Main 

Feedwater Preheater Bypass Isolation Valves 

and keep them closed against the maximum dif

ferential pressure across the valve seats. 

Upon discovery of this condition, Shearon 

Harris established a surveillance interval for 

verifying that the actuator components were 

functioning properly and that the accumulators 

were fully pressurized. Subsequently, non-Q 

components were replaced with suitable compo

nents and testing was completed satisfactorily. 

A postulated air leak in the non-safety-related 

IA piping could reduce the air inlet pressure to 

just low enough to affect proper operation of the 

actuator's 3-way and 4-way pilot valves for the 

Main Feedwater Preheater Bypass Isolation 

Valves, and not be detected by the pressure 

switches in the main header of the Instrument 

Air System. Refer to Shearon Harris 

LER 40098001 in Table 2. If this occurred, the 

pilot valves would shuttle, causing the

NUREG/CR-6654 34
Other Pertinent AOV Events 

and Operating Experience

accumulator pressure to bleed off, which would 

prevent the valves from closing as required. 

Operations personnel would have no indication 

of accumulator low pressure other than local 

observations made by an auxiliary operator and 

possibly dual valve indication in the main 

control room due to the valves cycling slightly. 

This potential scenario was reported to the NRC 

via the emergency notification system on 

January 9, 1998. Refer to LER 40098001. 

9.4 Problems with Copes

Vulcan Pressurizer 

Power-Operated Relief 

Valves (PORVs) 

The NRC issued Information Notice 94-55 to 

describe design-related problems involving 

cracking of plug material and severe wear of 

plugs and cages in Copes-Vulcan PORVs. 

On April 7, 1994, at the Salem Generating 

Station, Unit 1, a reactor trip was followed by 

two automatic actuations of the safety injection 

(SI) system. The continued injection of water 

from the safety injection system filled the pres

surizer steam space with subcooled water and, 

without the normal pressurizer steam space to 

dampen pressure excursions, resulted in repeated 

actuation of the plant PORVs to limit reactor 

coolant system pressure. Salem Unit I has two 

pressurizer PORVs, each of which is actuated 

through separate automatic controls. During the 

event, one PORV (1PR-2) cycled at least 

200 times, and the other PORV (lPR-1) cycled 

at least 100 times. 

The PORVs are 2-inch, air-operated valves 

manufactured by Copes-Vulcan that have a plug 

and cage-type internal trim design. The Copes

Vulcan PORV design used at Salem Unit 1 has a 

plug which is guided by a cage. Close clearances 

exist between the outside diameter of the plug 

and the inside diameter of the cage. When the 

valve opens, fluid in the system flows from 

under the plug through several equally spaced 

ports in the cage, and then to the valve outlet. 

When the valve closes, the plug seats against a 

machined surface of the cage. The cage is posi

tioned on a gasket in a close-clearance

counterbore in the valve body. The stem 

transmits the motive force from the air actuator 

to the plug and is threaded into the plug. A steep 

taper section on the stem just above the threaded 

section produces a wedging action at the 

relatively thin pinning boss. The valve stems 

were made of Type 316 austenitic stainless steel, 

and the plug and cage are both made of 

Type 420 hardened martensitic stainless steel. 

The licensee inspected the internal compo

nents of the PORVs and discovered three defi

ciencies: (1) scoring in the plug and cage area 

(both valves); (2) axial cracking on the pinning 

boss through which the anti-rotation roll pin 

passes (both valves); and (3) galling on the stem 

where it passed through the bonnet (1PR-2 

only). The licensee believed the scoring found 

on the plug and cage of IPR-2 was the result of 

out-of-tolerance machining of the inside diame

ter of the cage and the increased thickness 

caused by deposition of material as the scoring 

occurred. 

The licensee determined that the cracks in the 

pinning boss were caused by intergranular stress 

corrosion cracking (IGSCC). The extent of 

cracking was similar for both 1PR-1 and 1PR-2. 

There was no evidence that fatigue contributed 

to the failure. The licensee found similar crack

ing, though less prominent than the cracking in 

1PR-1 and 1PR-2, on internal components of 

valves maintained in the warehouse as new spare 

valves. The licensee performed stress and frac

ture mechanics analyses to evaluate the stress 

condition in the valve plug and to assess the 

potential for additional crack growth. These 

analyses indicated that differential thermal 

expansion of the stem and plug materials causes 

significant stresses in the pinning boss. In addi

tion, because of the steep taper wedging action 

in the stem-to-plug assembly, high stress con

centrations are found in the vicinity of the anti

rotation pin hole of the plug boss. The licensee 

determined that continued crack growth further 

into the much heavier plug itself is possible, if 

the plug is left in service. 

To prevent further problems with the internal 

components made of Type 420 stainless steel, 

the licensee installed plugs made of Type 316

35 35NUREG/CR-6654
Other Pertinent AOV Events 

and Operating Experience

stainless steel overlaid with Stellite and cages 

made of 17-4 PH stainless steel in the Unit 1 

PORVs. The design of the replacement plug 

eliminates the pinning boss used for the Type 

420 stainless steel plug, thus eliminating the 

tendency for crack formation in the thin boss 

section. The stem is now pinned to a thick 

section of the plug rather than through the rela

tively thin boss, and the plug height has been 

increased (to account for the elimination of the 

boss) to provide the same stroke length as 

before. The licensee determined that the Unit 2 

PORVs did not require modification because 

their plugs and cages are made of 17-4 PH 

stainless steel. 

The manufacturer determined that Salem was 

the only nuclear power facility that installed the 

Type 420 stainless steel internal components in 

Copes-Vulcan pressurizer PORVs. However, 

according to the manufacturer, there may be 

other Copes-Vulcan valves in nuclear service 

utilizing Type 420 material for the internal com

ponents, and these components could experience 

similar problems. 

In summary, NRC Information Notice 94-55 

raised questions with respect to: (1) the use of 

Type 420 stainless steel in this or similar valve 

applications; (2) valve component misalignment 

during field assembly; (3) out-of-tolerance 

machining on Copes-Vulcan PORVs; and 

(4) high stress concentrations because of the 

design of the internal components of Copes

Vulcan PORVs. 

Additional problems related to air actuators 

and air supply regulators in Copes-Vulcan 

PORVs were reported in NRC Information 

Notice 95-34, dated August 25, 1995. During an 

inservice testing surveillance at the Haddam 

Neck nuclear power plant on February 19, 1994, 

both pressurizer PORVs failed to open fully on 

demand. The cause was leaks in the air actuator 

assemblies of both PORVs. Reduced pressure 

output of the control air regulators compounded 

the problem. The Haddam Neck PORVs are 

2-inch nominal size, air-operated plug valves 

manufactured by Copes-Vulcan (Model 

D-100-160).

The licensee determined that the primary 

cause of the unacceptable valve stroke perform

ance was air leakage from the PORV air actua

tors caused by improper installation of the dia

phragms. Both PORV diaphragms had been 

replaced by a new style during a 1993 refueling 

outage. The principal difference in the 

replacement diaphragms was a change in the 

material composition. The replacement dia

phragm was made of EPDM while the old style 

diaphragm was made of Buna-N. The licensee 

switched to the EPDM diaphragms because of a 

vendor recommendation that EPDM would pro

vide enhanced performance under the tempera

ture and radiation conditions experienced by the 

PORV. In addition, the EPDM diaphragms have 

a 24-bolt hole configuration while the old style 

Buna-N diaphragms only have a 12-bolt hole 

configuration. The diaphragms also have slightly 

different shapes. 

The licensee apparently had some difficulty 

installing the EPDM diaphragms because of the 

bolt hole pattern and shape differences between 

the EPDM and Buna-N style diaphragms. The 

licensee believes that the sealing surfaces of the 

diaphragms were damaged as a consequence of 

the installation difficulties. Extrusion of the dia

phragm from between the base and cover and 

away from the bolt holes led to small tears at 

several diaphragm bolt holes locations which 

ultimately resulted in the air leakage. The licen

see chose to use the Buna-N diaphragms to 

avoid the installation difficulties encountered 

with the EPDM diaphragms. Copes-Vulcan 

indicated that they have no reports from other 

users on installation difficulties of either type of 

diaphragm. 

According to IN 95-34, numerous air

actuator related problems were reported within 

the nuclear industry regarding this model of 

Copes-Vulcan valve. The problems can be 

grouped into three categories: (1) actuator air 

leaks resulting from in-place diaphragm failures 

(e.g., holes, rips, and tears); (2) actuator air leaks 

resulting from loose actuator cover bolts; and 

(3) valve stroke malfunctions resulting from 

improper supply of air pressure from the air 

regulating valves.

NUREG/CR-6654 36
Other Pertinent AOV Events 

and Operating Experience

The air pressure regulators used at Haddam 

Neck are ITT Conoflow and contributed to the 

PORV problems on three separate occasions. In 

1993, one of the air pressure regulators failed 

high, subjecting one of the PORV diaphragms to 

the full air supply line pressure of 120 psig 

which is greater than the diaphragm design 

pressure of 100 psig. Although the PORV 

actuators are equipped with relief valves to pro

tect the diaphragms from over pressurization, it 

is believed that the high pressure contributed to 

or caused premature failure of the diaphragm. In 

the other two instances, the air supply regulator 

setpoints had drifted low, resulting in inadequate 

stroke performance of the valves. The air pres

sure regulator set points for the Haddam Neck 

configuration are 85 psig. The PORVs need 

65 psig to start opening and 85 psig to open 

fully. An engineering evaluation by the licensee 

showed that the valves will come to the full open 

position with control air pressure reduced to 70 

psig and reactor coolant system pressure as low 

as 840 psig. 

Several postulated causes of air pressure 

regulator set point drift were described in 

IN 95-34. One was that moisture intrusion from 

the control air system could cause corrosion of 

the regulating mechanism; another was that the 

drift might be configuration related. At Haddam 

Neck, the air regulating valve is upstream of the 

(normally closed) solenoid operated valve, 

meaning that the air regulating valve is con

stantly subjected to system pressure. The air 

regulating valve vendor indicated that this con

figuration may cause set point drift. 

9.5 Air-Operated Dampers 

Dampers are used in such vital services as 

control room ventilation and to direct sources of 

combustion air to emergency diesel generators. 

Air-operated dampers may be piston-operated 

with the pistons powered from instrument air or 

accumulators and SOVs may serve as pilot 

valves to direct air to the pistons. Alternatively, 

SOVs may be used to power dampers directly. 

Some or all air-operated dampers may not be 

included in the licensees' AOV programs. For

example, Palisades specifically excluded HVAC 

dampers from the scope of their AOV program. 

Other licensees may not explicitly state specific 

exclusions in their program scope; however, 

their lists of AOVs may not include dampers. 

Motor-operated dampers are not included in the 

scope of the Generic Letter 89-10 program. 

In the visits to the plants, it was noted that 

dampers and their operators are usually main

tained by the technicians in charge of the par

ticular system in which the dampers are located. 

This appears to be an approach that could result 

in nonuniform maintenance or incomplete fail

ure analyses because of the diversity of person

nel and concentration. 

Several events were noted where air-operated 

dampers with safety-related and non-safety

related functions may be served by non-safety

related systems such as instrument air without 

adequate accumulator backups. Interruption of 

the non-safety-related service would cause inter

ference with the safety functions. A variation of 

this type of problem is that some safety-related 

functions of air-operated dampers have been 

performed by non-safety-related air-operators or 

SOVs. Some licensees attributed these condi

tions to design error, inadequate maintenance, or 

inadequate surveillance testing. 

NUREG-1275, Volume 2, includes 

(page 2-1) descriptions of events at Brunswick, 

H. B. Robinson, and Cooper, involving potential 

loss of emergency diesel generators in the event 

of a loss of offsite power. The loss of offsite 

power could have resulted in a loss of instru

ment air which would cause the dampers that 

serve the emergency diesel generator rooms to 

close, and subsequently would cause overheat

ing of the diesel generator controls. 

By Generic Letter 88-14, dated August 8, 

1988, the NRC directed licensees to perform 

design verifications. Some deficiencies were 

found as a result but similar conditions were 

recently discovered. Selected LERs that describe 

safety-related events or deficient conditions 

involving air-operated dampers are listed in 

Table 3.

37 NUREG/CR-6654
Other Pertinent AOV Events 

and Operating Experience

9.6 AOVs and SOVs Used in 

Services Related to 

Emergency Diesel 

Generators 

AOVs and SOVs are used extensively in 

emergency diesel generator service in nuclear 

power plants. For example, cooling water for the 

emergency diesel generators is controlled by 

AOVs at several plants. SOVs are used as pilots 

for a number of AOVs mounted directly on the 

diesel engines. 

In general, AOVs and SOVs that are 

mounted on the diesel generators at the plants 

visited were maintained by the technicians 

assigned to the diesel generators. It was 

observed during the visits that the operability of 

the valves might be improved if responsibility 

for maintenance of these valves were assigned to 

personnel specifically concerned with AOV and 

SOV maintenance. 

NUREG-1275, Volume 2, subsection 5.1.7 

and its Appendix B, include detailed descrip

tions of a number of problems and events related 

to air service to the AOVs and SOVs that supply 

diesel generators. 

Water in the instrument air caused sticking of 

an air-operated radiator exhaust damper for 

Diesel Generator No. 2 at Ft. Calhoun 

(LER 28587025). Shortly thereafter, it was dis

covered that Diesel Generator No. 1 had the 

same degraded condition. If a loss of off site 

power event had occurred while the DG radiator 

damper AOVs were inoperable, both DGs would 

have run to destruction because the high coolant 

temperature shutoff would have been bypassed 

during a real demand. This event was analyzed 

under NRC's ASP program, categorized as a 

precursor event, and estimated to have a condi

tional core-damage probability of 6.2E-04, 

which ranked it as the second highest event of 

33 precursor events identified for that year. 

Several diesel generator events involving 

SOVs were described in NUREG-1275, 

Volume 6, including a simultaneous common

cause emergency diesel generator failure

(Section 5.2.1.2 therein) that was traced to 

deterioration of the elastomer materials used in 

the SOVs. 

A recent example of an SOV design defi

ciency that was responsible for common-cause 

inoperability of several diesel generators at sev

eral nuclear plants involves SOV springs which 

were redesigned to solve one problem (air 

leakage in the SOV) and, as a result, did not 

meet the minimum requirements for combined 

voltage and air pressure to actuate the solenoids. 

Refer to LERs for Point Beach (LER 26698008) 

and Clinton (LER 46198009) that reference 

Engine Systems Inc. 10 CFR Part 21 Notifica

tion No. 1998120, dated January 26, 1998 (NRC 

Accession Number 9802020122). The condi

tions as described in the Part 21 notification 

were: 

"MKW Power Systems, Inc. (former 

name of Engine Systems, Inc.) issued 

Report No. 1OCFR21-0055 on June 13, 

1990 which provided a 1OCFR21 report

able defect with the air start solenoid 

valve commonly used on EMD diesel 

generators (EMD #9513134/ Graham

White #712-051). The valves exhibited air 

leakage when maximum allowable inlet 

air (200 psig) was applied. This would not 

prevent the engine from starting; 

however, the leakage may cause the air 

compressor(s) to constantly cycle. After 

consulting with the valve manufacturer 

(Graham-White), MKW determined that a 

275 psig valve was also available. The 

275 psig valve was identical to the 

200 psig valve except that it had a differ

ent spring. MKW made the recommenda

tion to replace the valve's spring with a 

different spring which was rated for 275 

psig. MKW also discontinued sale of the 

200 psig valve for nuclear service and 

recommended the 275 psig valve as a 

replacement. The 275 psig valve is Gra

ham-White #712-065 (without mounting 

bracket) or #712-015 (with mounting 

bracket)." 

* "Engine Systems, Inc. (ESI) has recently 

learned that the 275 psig valve does not

NUREG/CR-6654 38
Other Pertinent AOV Events 

and Operating Experience

meet the minimum DC voltage require

ment of most nuclear applications when 

inlet pressures below 200 psig are applied 

to the valve. The solenoid valve relies on 

system air pressure to assist the coil in 

overcoming the force of the spring and 

thus operating the valve plunger. The 

reduced system pressure combined with 

the reduced coil voltage results in the 

inability of the valve to operate satisfacto

rily. This problem is not applicable to the 

original 200 psig valve because it has a 

weaker spring and therefore less air pres

sure is needed to overcome the lower 

spring force." 

The SOVs control air flow from the storage 

tanks to the diesel starting air system. There 

were 335 SOVs, distributed among about 

17 U.S. nuclear plant owners, listed in the 

Part 21 notice; however, only the two LERs 

noted above, which refer to these conditions, 

were found. 

9.7 Establishment of and 

Conformance with the 

Design Basis for AOVs at 

Millstone 3 

Mistakes in establishing the original plant 

design basis and/or mistakes in designing 

equipment in accordance with the plant design 

basis were reported in several LERs involving 

AOVs and the air systems that support AOVs at 

Millstone 3 in 1996. LERs 42396013, 

42396028, 42396031, 42396036, and 42396040, 

as outlined in Table 2, described these events. A 

NRC Combined Inspection Report, "NRC Com

bined Inspection Report 50-245/98-206; 

50-336/98-206; 50-423/98-206, and Notice of 

Violation," dated May 26, 1998 (NRC Acces

sion No. 9806030375), summarized the resolu

tion of the issues and the plant modifications 

resulting from the events and conditions. The 

events described below are considered to be 

common-cause failure conditions. 

LER 42396013 described an original plant 

design deficiency where the Residual Heat 

Removal System (RHS) was outside the design

basis of the plant. A loss of control air supplied 

from the non-safety-related instrument air sys

tem could cause the RHS control valves to fail 

open. If this condition occurred during the initial 

phase of a plant cool down, the Reactor Plant 

Component Cooling Water System (CCP) tem

peratures could exceed 125'F. This is the design 

temperature used in the system stress analysis. If 

RHS heat exchanger operation was initiated at 

350'F RCS temperature, as assumed, then the 

RHS heat exchanger CCP outlet temperature 

could be as high as 250'F if the valves failed 

open. Under the resultant conditions the CCP 

piping would not meet the ASME Section III, 

Appendix F stress criteria. The original plant 

design did not consider that if the RHS flow 

control valves failed open on a loss of control 

air, it could create unacceptably high RHS heat 

exchanger CCP discharge temperatures. 

LER 42396028 described a failure scenario 

in which a loss of instrument air to temperature 

control valves in the Charging Pump Cooling 

(CCE) system serving the charging pump lube 

oil coolers, coincident with 330F Service Water 

(SWP) temperature could result in overcooling 

of both trains of the charging pump lube oil 

system and challenge charging pump operabil

ity. Failure of the air-operated CCE valves to the 

full open position due to a loss of the non-safety 

related IAS system would adversely affect both 

trains of the charging pumps by allowing exces

sive cooling of the CCE system, which cools the 

lube oil system. This condition alone could have 

prevented the fulfillment of the safety function 

of the system. The cause of the charging pump 

inoperability was inadequate original design. 

This condition would result from overcooling of 

the lube oil system from a failure of the non

safety related instrument air system coincident 

with a worst case minimum SWP temperature 

and maximum flow and heat exchanger 

cleanliness. Under these conditions, the air

operated CCE valves would fail open and 

excessive cooling of the lube oil system would 

occur. This particular combination of conditions 

was not considered in the initial design. 

LER 42396031 described a condition where 

specific safety related control valves could fail 

due to exceeding the manufacturer's (ASCO)

39 NUREG/CR-6654
Other Pertinent AOV Events 

and Operating Experience

maximum operating pressure differential pres

sure (MOPD) rating of the SOVs installed on the 

control valves in a number of safety-related 

systems. SOVs could fail to perform safety 

related functions because of excessive operating 

pressure differentials. This can result from fail

ures of non-qualified air regulators installed in 

the instrument air system upstream of the SOVs. 

The failure of an air regulator would, in turn, 

result in full IA system pressure being applied to 

the SOV. The SOV can potentially fail to oper

ate properly since they are not rated for full IA 

system pressure. A total of 48 SOVs which per

form safety-related functions had been originally 

identified in various systems which would be 

susceptible to such a failure. The cause of this 

condition was the failure to consider the poten

tial for pressure regulator failure in the original 

design and selection of SOVs. 

LER 42396036 described a scenario where 

the High Pressure Safety Injection (S1I) and 

Low Pressure Safety Injection (SIL) systems 

would have been subject to degraded perform

ance due to possible mispositioning of normally 

closed safety related air operated valves. Mispo

sitioning of these 21 valves can be postulated to 

occur under post-accident harsh environmental 

conditions due to failure of non-qualified power 

and control circuits. As a result, the potential 

diversion of SIH and/or SIL flow under accident 

conditions could have been more than the mar

gin allowed within the Loss of Coolant Accident 

analysis. Seventeen additional safety-related air 

operated and solenoid operated valves were sub

sequently identified where failures of non

qualified control circuits could degrade 

performance of a safety system function. These 

additional valves are located in the following 

systems: Reactor Plant Component Cooling 

Water (CCP), Containment Vacuum, Reactor 

Plant Sampling (SSR); Post Accident Sampling 

(SSP); and Main Steam to the auxiliary feedwa

ter steam turbine. The cause of the reported con

ditions was a design mistake. The initial plant 

design did not adequately consider the potential 

mispositioning of these valves under harsh envi

ronmental conditions or active failure. 

LER 42396040 described a scenario where a 

loss of the non-category 1 Instrument Air

System (IAS) would allow CCP valves to 

reposition to a maximum cooling configuration. 

Coupled with a low heat load and minimum 

Service Water (SWP) inlet temperature, the CCP 

system could reach temperatures lower than 

values for which they are analyzed, thereby 

rendering the CCP system, and systems that it 

serves, potentially inoperable. The CCP system 

temperature is controlled by three-way valves, 

each comprised of a single pneumatic operator 

and a mechanically linked pair of butterfly 

valves located at the outlet end of each of the 

three CCP heat exchangers. Each pair of valves, 

a heat exchanger (through-flow) outlet control 

valve and a heat exchanger bypass valve, are 

mechanically linked such that when one valve is 

fully open, the other is fully closed. This design 

ensures adequate heat removal under the high 

heat load conditions assumed in the design basis 

analysis. However, if the non-safety-related IAS 

system that operates the three-way CCP tem

perature control valves were lost, the valves 

would fail in a configuration that would result in 

maximum CCP system cooling (i.e., the heat 

exchanger bypass would fully close and the heat 

exchanger outlet control valve would fully 

open). This maximum cooling, if coupled with 

low CCP heat loads (experienced during an 

extended plant shutdown and very low SWP 

temperatures (during winter seasonal condi

tions), could result in CCP system overcooling. 

Such an event could cause temperatures to 

decrease below analyzed limits, thereby render

ing the CCP system, and systems that it serves, 

potentially inoperable. The root causes of this 

condition were: 

"* Improper initial design of the CCP sys

tem. A review of the CCP design basis 

and system-related correspondence 

revealed that the described failure mode 

described was a design oversight. The 

design basis analysis for the CCP system 

focused on high CCP heat load condi

tions. The designers did not analyze for 

extremely low CCP heat loads concurrent 

with very low SWP temperatures. 

"* Inadequate review of industry and plant 

operating experience evaluations associ

ated with the CCP system and loss of the

NUREG/CR-6654 40
Other Pertinent AOV Events 

and Operating Experience

IAS system. Since original design imple

mentation, there have been both industry

wide and plant events dealing with these 

systems and their interfaces. Many issues

dealt specifically with system temperature 

problems and these should have been 

investigated.

41 NUREG/CR-6654
10. AIR SYSTEMS OPERATING EXPERIENCE

10.1 Air Systems 

Air systems in nuclear plants can be a source 

of common-cause failures in AOVs if the air 

provided is not clean, dry, oil-free, and furnished 

at specified pressure. Also, if air supplied to 

AOVs is once contaminated, the potential for 

AOV problems is not necessarily eliminated 

once the quality of the air has been restored. 

Residues of moisture in the systems can generate 

AOV problems both directly and over time, 

from corrosion products that may form and be 

released into the system. 

NUREG-1275, Volume 2, "Operating Expe

rience Feedback Report - Air Systems Prob

lems," was published in December 1987. In the 

NUREG report, the AEOD staff described air 

system problems caused by moisture contami

nation as well as hydrocarbon and desiccant 

contamination, among other problems. These 

findings led directly to the NRC issuing 

Generic Letter (GL) 88-14, "Instrument Air 

Supply system Problems Affecting Safety

Related Equipment," in August 1988. In 

GL 88-14, licensees were asked to verify the 

quality of instrument air and discuss their 

programs for maintaining instrument air quality. 

This verification was to include: 

1. Verification by test that actual instrument 

air quality is consistent with the manu

facturers' recommendations for individual 

components served. 

1. Verification that maintenance practices, 

emergency procedures, and training are 

adequate to ensure that safety-related 

equipment will function as intended on 

loss of instrument air. 

2. Verification that the design of the entire 

instrument air system including air or 

other pneumatic accumulators is in accor

dance with its intended function, includ

ing verification by test that air-operated 

safety-related components will perform as 

expected in accordance with all

design-basis events, including a loss of 

the normal instrument air system. This 

design verification should include an 

analysis of current air operated 

component failure positions to verify that 

they are correct for assuring required 

safety functions. 

In addition to the above, each licensee/ 

applicant should provide a discussion of their 

program for maintaining proper instrument air 

quality." Note that recommendations for 

periodic verification of air system operability 

and air quality were not specifically included in 

GL 88-14. 

As noted previously, air systems (including 

backup nitrogen systems) are supposed to pro

vide clean, dry, oil-free air, at specified pressure, 

to AOVs. All seven of the licensees visited have, 

in the past, encountered numerous AOV prob

lems caused by moisture in the air supplied to 

their AOVs. Six of the seven licensees took 

aggressive actions to improve the moisture 

reducing performance of their air systems. 

Among the actions taken were the installation of 

additional or more reliable air capacity, installa

tion of redundant dryers, installation of addi

tional air filters for the system and for individual 

AOVs, installation of automatic drains, and 

installation of continuous dew point monitoring 

and alarms. 

The history and current condition, as of the 

time of our visit, of the instrument air system at 

one of the sites visited demonstrates the impor

tance of supplying dry and clean air to valves 

and instruments. Two of the events noted in 

Section 5.1.1 of NUREG-1275, Volume 2, con

cerned the failure, in 1978 and 1981, of a shut

down cooling system heat-exchanger outlet 

AOV, CV-3025, to open while the reactor was 

shut down. LERs 25578003 and 25581030 refer 

to these events. CV-3025 has a non-safety

related decay heat removal function and is 

required to be opened after the reactor is depres

surized. CV-3025 also has a safety-related func

tional requirement to open after a small-break 

loss-of-coolant accident (LOCA) in order to

NUREG/CR-6654 43
Air Systems Operating Experience

provide for shutdown cooling. This AOV is also 

intended to be throttled to adjust cooling capac

ity after a small-break LOCA, but the licensee 

considers this function to be beyond the design 

basis for the plant. Note that CV-3025 is pow

ered from the non-safety-related IA system. A 

handwheel was added after the 1981 valve fail

ure to allow operators to open the valve 

manually. 

Water was present in the IA system at this 

plant because of improper dryer operation in 

each of these events. In each event, the water 

filled a positioner and caused the valve to fail 

closed. Finally, in each event, the primary cool

ant temperature approached boiling before 

operators took alternative actions to remove 

decay heat. This recurring single failure, in a 

non-safety-related mode for the valve operation, 

could have resulted in uncovering the reactor 

core had the conditions not been mitigated by 

other operator actions within a few hours. 

As of the time of our visit, these events were 

still significant because the quality of instrument 

air is compromised by the periodic introduction 

of moisture into the system. There was, at the 

time of the visit, only one dryer available for the 

IA system. As a result, no air drying capability is 

available when the dryer must be taken off line. 

This, almost certainly, is causing contamination 

of the air system from moisture, and subse

quently, as a direct result of the presence of 

moisture in the piping, from corrosion products 

which can be dislodged from the pipe or valve 

internals. According to the licensee, air quality 

directly downstream of the dryer is monitored 

and moisture content is estimated by direct 

observation on a daily basis of changes in the 

color of desiccant crystals. However, such 

readings are meaningless if the dryer is 

bypassed. The readings do not alert operators to 

downstream moisture that would be introduced 

when the dryer is bypassed. 

In addition to the instrument air system, the 

plant has a high-pressure air system that serves 

"vital" valves. The air for this system is supplied 

by three compressors, each with a refrigerant

type dryer that cannot dry the air to a 

satisfactorily low dew point. The dryers do not

operate reliably, i.e., the air compressors for the 

high-pressure air system must be cycled on and 

off in order to prevent these refrigerant dryers 

from freezing up. In April 1997, nine air 

regulators in the high-pressure air system were 

found to be contaminated, and it is concluded 

that this contamination resulted from moisture 

intrusion because dry air was not provided by 

this system. This condition was not reported in 

an LER. 

It was also noted that several equipment air 

filters were, as of the time of the visit, installed 

downstream of equipment rather than upstream 

of the components they were designed to protect 

from particulates. Further, during a more recent 

NRC Regional inspection, several gallons of 

water were found at a low point in the air line. 

The licensee had not installed a low-point drain 

to remove water from the high-pressure IA sys

tem. The licensee committed to do this in 

response to GL 88-14, but the modification had 

not been implemented as of the time of the visit 

for this study. The plant staff prepared plans to 

improve the air systems but these had not been 

implemented. They did not make modifications 

to the air dryers because there were no licensing 

requirements applicable to them. This is an older 

plant which was operating before the existence 

of the ISA 7.3 (1975) instrument air standard. 

A somewhat more satisfactory situation 

regarding air quality was found at Turkey Point 

during the recent visit for this study. In 1985, 

Turkey Point 3 and 4 experienced recurring fail

ures in AOVs in the auxiliary feedwater system 

(LER 25085021) caused by moisture or moisture 

contamination of the IA system. The events are 

described in Section 5.1.2 of NUREG-1275, 

Volume 2. See Section 8.7.1 in this report, as 

well. Water had been allowed to condense and 

remain in the instrument air system, subse

quently producing corrosion particles which then 

migrated, along with the moisture, to operating 

valves and valve control equipment. In addition 

to the AFW systems for both units, other safety 

related systems which could have been affected 

included steam dump to atmosphere, salt water 

flow to the essential heat exchanger, charging 

system, and the residual heat removal system.

NUREG/CR-6654 44
Air Systems Operating Experience

The Turkey Point units are now equipped 

with a full-capacity IA system, continuous 

digital-readout dew point monitors, dew point 

alarms (although not in the control room), drain 

traps, and enhanced filtering arrangements. Each 

unit is equipped with two "double-capacity" 

compressors (one motor-driven and one diesel

driven), each of which is hard-piped to serve 

both plants. In addition, the IA dryers are con

nected to the emergency bus, thereby ensuring a 

reliable flow of clean, dry air during design basis 

accidents or events, including loss of offsite 

power. Recent AOV performance indicates that 

these measures have been effective in preventing 

AOV failures from contaminated air. 

All of the plants visited took actions to 

improve the performance of their air systems in 

response to in-plant events and the recommen

dations in GL 88-14, although the individual 

responses and results varied considerably from 

plant to plant, as indicated above. Modifications 

included installation of additional air production 

capacity, additional parallel dryer trains, instal

lation of additional filters, and more attention to 

preventive or corrective maintenance of the air 

system. 

Palo Verde, Fermi 2, and Palisades do not 

have continuous dew point monitors in their air 

systems. They rely heavily on observations of 

blown-down air to evaluate moisture contami

nation. TMI-1, Indian Point 3, and Turkey Point 

have each installed continuous dew point moni

tors. LaSalle was originally designed with a dew 

point monitoring system, including alarms. 

LaSalle, TMI-1, and Indian Point 3 have 

installed dew point alarms in the control room. 

Indian Point 3 and Turkey Point 3 and 4 each 

have installed automatic drain traps on their air 

receivers. The automatic drain traps have proved 

to be effective in removing moisture from the air 

prior to processing by the dryers. 

Three of the seven licensees visited do not 

definitively know, on a day-to-day basis, if the 

quality of air (or nitrogen) in their plants is 

acceptable or has degraded with regard to mois

ture contamination. This could adversely affect 

the operation of important and safety-related 

AOVs. LaSalle, TMI, Indian Point 3, and

Turkey Point 3 and 4, are equipped with 

continuous dew point monitoring and alarming 

of the instrument air systems. The alarms at 

Turkey Point are local, whereas the alarms at the 

other three stations are located in the control 

rooms. Note that these plants also have other air 

systems containing AOVs of interest that are not 

equipped with devices to measure dew point. 

The licensees without continuous monitoring 

devices check moisture content in their air sys

tems by periodic observation ranging from tak

ing readings at each shift (on rounds) to testing 

during outages. The extent of particulate con

tamination at all of the plants visited is estimated 

from examination of filters and ranges from 

actual measurement of particle number and size 

to casual examination of the filters. 

Recent operating experience indicates that 

moisture contamination of air systems in nuclear 

power plants occurred since GL 88-14 was 

issued. Some examples are: 

At Beaver Valley 1 in 1990 

(LER 33490007), moisture in the IA 

system caused the failure of feedwater 

regulating valves and a subsequent reactor 

trip. 

At Dresden 2 in 1994 (LER 23794005), 

an air line ruptured because of corrosion 

in an IA line and this led to a reactor trip. 

During a test of the pressurizer PORVs at 

Haddam Neck in 1993 (LER 21393007), 

a failure in the PORV air supply regulat

ing valve was traced to moisture intrusion 

in the containment control air system due 

to a faulty air dryer. 

* Similarly at Haddam Neck, a letdown 

AOV failed in 1993 (LER 21393005), 

because of moisture in the containment 

control air system. 

* In 1992 (LER 32792018) at Sequoyah 1, 

about 1000 gallons of water was found to 

have been entrained in the non-essential 

air system that serves the essential air 

system. This caused malfunction of the 

feedwater regulating valves and resulted 

in a trip.

45 NUREG/CR-6654
Air Systems Operating Experience

In 1997 (LER 32797012) at Sequoyah 1, 

loss of control air header pressure was 

attributed to corrosion products in the 

lines (residue from the moisture intrusion 

event which had occurred about 5 years 

earlier). 

In several of these events, air was provided to 

the equipment. However, the air was contami

nated with moisture and/or corrosion-products 

which caused the failures. The interest in air 

systems in this study is directed toward both the 

quantity and quality of air supplied to AOVs. If 

the air supply is not clean and dry, its source 

cannot be considered reliable regardless of how 

much air can be provided. 1A systems that are 

not kept clean and dry are sources of common

cause failures of AOVs, both immediate and 

delayed (See NUREG-1275, Volume 2). As dis

cussed elsewhere in this study, air must be sup

plied at constant pressure (within specified lim

its) in order to ensure reliable operation of 

AOVs and SOVs. 

The effects of moisture contamination can be 

immediate, i.e., the moisture itself causes AOVs 

or instrumentation to fail. The effects of mois

ture contamination can also be delayed, i.e., cor

rosion or contamination products form or build 

up and cause a number of failures or events. 

While the long-term hazards of moisture in air 

lines are not readily predictable, they are cer

tainly detrimental to equipment operability and 

the hazards may increase or continue to exist 

long after any one or several particular problems 

are discovered and corrected. Similar conclu

sions were noted in NSAC-128, "Pneumatic 

Systems and Nuclear Plant Safety," in 1988. 

NUREG/CR-5472, "A Risk Based Review of 

Instrument Air Systems at Nuclear Power 

Plants," included an analysis, from a risk per

spective, of the data presented in NUREG-1275, 

Volume 2, and LERs available between 1980 

and 1989. Among the observations in that study, 

Section 2.4.2 indicated that "(I)nstrument air 

contamination predominates as a root cause and 

air-operated valve failure predominates as a 

direct cause in instrument air common cause 

events." Other conclusions and data in

NUREG/CR-5472 are discussed in Sections 15.1 

and 15.4. 

A previous NRC report, NUREG/CR-2796, 

"Compressed Air and Backup Nitrogen Systems 

in Nuclear Power Plants," focused primarily on 

loss of instrument air as a source of common

cause failure of AOVs, and concluded that the 

air systems were reliable in terms of the ability 

to provide the necessary quantities air. This cor

responds to the observations made from the 

plant visits. Air capacity appears to be plentiful 

and redundant. As far as the issue of reliably 

providing air at the necessary pressure is con

cerned, there have been a number of instances 

where the failure of a pipe, header, or piece of 

equipment caused an air leak that lowered pres

sure locally, or even generally for a short time, 

but this was not considered of particular concern 

for the non-safety-related air system, especially 

when a safety-related event did not occur at the 

same time. 

It must be noted that until NUREG-1275, 

Volume 2, was published and GL 88-14 was 

subsequently forwarded to all licensees, the 

assumptions made by licensees and used in air 

system studies (including risk studies) concern

ing adequacy of backups to instrument air 

(accumulators), as well as intended AOV failure 

positions, were unverified. Licensees were asked 

to verify that their plants were capable of 

responding safely to a loss of (instrument) air, 

that valves would respond as intended, and that 

accumulator capabilities were acceptable. Ques

tions still remain (in some of the plants we vis

ited) about the effectiveness of ongoing activi

ties to protect important AOVs against partial or 

gradual air pressure losses and the effects of 

contaminated air. 

10.2 Accumulators for AOVs 

Accumulators are used to provide an air sup

ply to AOVs which are required to function if 

the normal air supply is lost. This strategy is 

commonly applied to safety-related AOVs so 

that credit is not needed for a non-safety-related 

instrument air system. Essentially, if an accident 

analysis indicates that air must be supplied to 

operate a valve, an accumulator is provided.

NUREG/CR-6654 46
Air Systems Operating Experience

A number of failures involving accumulators, 

check valves, and piping are recorded in Table 2 

and include the following: 

"• LER 44595005 at Comanche Peak; 

"* LER 34688007 at Davis Besse; 

"* LER 24993005 at Dresden 3; 

"* LER 33191005 at Duane Arnold; 

"* LER 28588028 at Ft. Calhoun; 

"* LER 33689011 at Millstone 2; and 

"* LER 27093002 at Oconee 2. 

In addition to hardware failures, several 

instances were identified where the capacity of 

the accumulators was found to be insufficient to 

meet design basis demands. These include the 

following: 

* LER 25085020 at Turkey Point; and 

* LER 29389002 at Pilgrim. 

In these last two examples, the design basis 

demands for the AOVs and/or the margins to 

meet those design basis demands were not cor

rectly established. 

Malfunction (sticking open or closed) of the 

accumulator check valves associated with par

ticular valve operators can be a significant 

problem. Note that accumulators, along with the 

accumulator check valves, are sometimes 

assumed to be piece parts of a safety-related 

AOV. First, the loss of air may not be readily 

apparent and the valve would then fail when

operation is attempted. Second, a partial loss of 

air pressure could cause the valve to fail in an 

intermediate position or not fail safe if the 

springs are not properly designed and installed. 

Another potential problem involving accu

mulators was noted during the visits for this 

study. Most of those encountered have no drains 

in the bottom and no means of inspection with

out disassembly of the piping. This condition is 

typical at many plants. Further, there was no 

regular inspection program in the plants we vis

ited to determine conditions within the accumu

lators. Some accumulators have been in service 

for many years. 

Accumulators, by the nature of their design 

and location in the air systems, are natural drain 

traps for moisture and/or particulate contamina

tion. The possibility exists for contamination of 

the valve(s) served by an accumulator if there is 

moisture or particles (corrosion products for 

example) in the accumulator. Plants that have 

had problems with moisture in the air system at 

one time or another (most plants) could have 

either moisture in the accumulators, or corrosion 

products resulting from previous moisture con

tamination, or both, absent objective evidence 

(inspection results) to the contrary. The safety

related functions of the AOVs served by the 

accumulators could be compromised. 

Another problem associated with the lack of 

drains in accumulators is the possibility that the 

accumulators may trap an appreciable volume of 

water, thus decreasing the volume of air avail

able to the valve served. If the volume of the 

accumulators is less than called for in the design 

bases, the AOVs may not be able to stroke the 

number of times required (as assumed in the 

plant safety analyses).

47 NUREG/CR-6654
11. SOLENOID OPERATED VALVE (SOV) OPERATING EXPERIENCE

As noted in Section 5, characteristics of 

SOVs pertinent to this study include: 

"* Exceeding the maximum operating pres

sure differential (MOPD), which is the 

pressure difference between the inlet port 

and the outlet port, may cause the SOV to 

spuriously open or not open, depending 

on the SOV's design. 

"* Small orifices make the SOVs subject to 

interference from small contamination 

particles or moisture intrusion. 

"* The design logic of SOV operation and 

control can be complex and this is a 

source of design or installation mistakes. 

"* Many piloted SOVs require a minimum 

operating pressure differential (Min OPD) 

to operate properly. 

"* Many SOVs will not function properly 

when subject to reverse pressurization or 

flow in the valve. 

Materials used to construct the valve 

bodies, internal parts, seals, O-rings, etc., 

may be subject to binding, creep, corro

sion, erosion, and/or environmental dete

rioration. 

"* Rather small forces, in the range of 

10 pounds or less, are produced by a 

solenoid to operate an SOV; therefore, 

opposing forces of similar small 

magnitude can interfere with successful 

operation. 

"* SOVs can be damaged if subjected to 

higher-than-designed pressure in the 

operating fluid. 

"* SOVs are subject to damage from the use 

of thread locking compounds in adjacent 

pipe or tube connections that tend to 

migrate into the working parts of the 

SOVs.

Elastomers and lubricants used in SOVs 

have been, and continue to be, a source of 

many SOV operational problems. 

One of the pertinent observations reported in 

NUREG/CR-4819, Volume 1 (page 25 therein), 

was that "SOVs used in nuclear power plants 

were originally designed for industrial applica

tions where an energize-to-operate philosophy is 

prevalent. In the industrial applications, the SOV 

is normally maintained in the de-energized state, 

resulting in the coil and elastomers being main

tained at room-ambient or process fluid tem

perature. The nuclear safety philosophy of 

returning to a fail safe condition on loss of air or 

electrical power can result in a valve being 

maintained in an energized state throughout 

most of its installed life. The consequence of the 

continuously energized SOV is more rapid deg

radation of the elastomers and solenoid coils 

because of the long periods at elevated tem

peratures. Although SOV manufacturers have 

considered the (continuously energized state) in 

design changes and material changes (to) their 

products, the continuously energized SOV will 

generally undergo more rapid degradation than 

an SOV used mostly in a de-energized state." 

Pertinent reports that described various 

aspects of the design, operational, or testing 

problems involved in the use of SOVs include: 

NUREG-1275, Volume 6, "Operating 

Experience Feedback Report - Solenoid

Operated Valve Problems;" 

"* NUREG/CR-3424, "Equipment Qualifi

cation Research Test Program and Failure 

Analysis of Class 1E Solenoid Valves;" 

"* NUREG/CR-4819, Volume 1, "Aging and 

Service Wear of Solenoid-Operated 

Valves Used in Safety Systems of Nuclear 

Power Plants, Operating Experience and 

Failure Identification;" 

"* NUREG/CR-4819, Volume 2, "Aging and 

Service Wear of Solenoid-Operated 

Valves Used in Safety Systems of Nuclear

49 NUREG/CR-6654
Solenoid Operated Valve (SOV) Operating Experience

Power Plants, Evaluation of Monitoring 

Methods;" 

"* NUREG/CR-5008, "Development of a 

Testing and Analysis Methodology to 

Determine the Functional Condition of 

Solenoid Operated Valves;" 

"* NUREG/CR-5141, "Aging and Qualifi

cation Research on Solenoid Operated 

Valves;" 

"* NUREG/CR-5292, "Closeout of IE Bul

letin 80-23, Failures of Solenoid Valves 

Manufactured by Valcor Engineering 

Corporation;" and 

"* NUREG/CR-6246, "Effects of Aging and 

Service Wear on Main Steam Isolation 

Valves and Valve Operators." 

SOV failure causes and mechanisms dis

cussed in NUREG-1275, Volume 6 include: 

High ambient temperatures; 

Heatup of the SOV from being continu

ously energized; 

* Subjecting SOV to pressure or flow in a 

direction not designed for; 

Misapplication of materials (metallic or 

organic) in design or replacement; 

* Over pressurization; 

* Incorrect wiring; 

* Incorrect electric current or voltage; 

* Inadequate preventive maintenance; 

* Mistakes made in rebuilding SOVs; 

* Misapplication of lubricants; 

* Misapplication of thread sealants;

* Contamination 

particulate);

(moisture and/or

"* Inadequate surveillance testing; 

"* Not recognizing and treating SOVs as 

(internal) piece parts of AOVs or other 

equipment; and 

* Use of unqualified SOVs. 

SOVs are sources of common-cause failures 

in AOVs. SOV-related root causes of such 

common-cause AOV failures were described in 

NUREG-1275, Volume 6 and include: 

" Incorrect specification of operating 

parameters such as maximum operating 

pressure differential (MOPD) or Min 

OPD; 

"* Incorrect specification of valve orienta

tion with respect to fluid flow; 

"* Incorrect material selection for the oper

ating environment; 

"* Incorrect specification of environmental 

(temperature, radiation, moisture) condi

tions; 

Incorrect estimate of expected service life 

when coil is continuously energized; 

Failure to replace worn parts or parts with 

limited service life in time; 

Rebuilding or re-assembling SOVs incor

rectly; 

Failure to maintain the supply of clean, 

dry, oil-free, instrument air at required 

pressure; 

* Excessive lubrication of SOV internals; 

Installing SOVs in an incorrect orienta

tion; 

Failure to provide correct electric current 

or voltage; 

Inadequate or incorrect electrical connec

tions; and

NUREG/CR-6654 50
Solenoid Operated Valve (SOV) Operating Experience

0 Manufacturing defects such as lubrication 

mistakes, defective materials, and assem

bly mistakes. 

A search of the LER database returned 

739 LERs since 1985 that involved SOVs. A 

widely used industry SOV application guide 

indicates that there were 2350 SOV failures 

cataloged in the nuclear industry database 

between 1985 and 1990. Several pertinent LERs 

involving failures of SOVs, which were 

reviewed as part of this study, are listed in 

Table 2. Refer to Section 9.6 for a description of 

a recent common-cause failure condition in 

diesel generator air-start SOVs. 

One recent Event Report (#34262, dated 

5/19/98, see Table 4) from Susquehanna 

described a 10 CFR Part 21 Notice by 

Valcor Corporation regarding their Model 

V70900-65-11 AC powered solenoid-operated 

air pilot valves. Also see Susquehanna Condition 

Report No. 98-2296. According to the reports, 

there were 40 of these SOVs in service, all at 

Susquehanna. The design of the SOV was such 

that residual magnetism in the plunger caused 

the solenoid to stick against the stop if the air 

gap was too small. Compression set in the 

0-ring seat caused the air gap to decrease to 

where the solenoid would stick after about 4 to 

18 months in service. This was a common-cause 

failure that resulted in multiple failed and 

degraded primary containment isolation valves 

and at least 5 degraded primary containment 

penetrations. The failures included one stuck 

open PCIV for over 4 hours, 2 delayed closing 

PCIVs, 7 imminent solenoid failures to the

non-fail-safe position, and a 59% expectation of 

near term failure to the non-fail-safe position for 

all 40 SOVs. 

Observations made during the visits to the 

seven nuclear plants for this study of AOVs, as 

well as an examination of pertinent LERs and 

other reports, indicated that SOV failures similar 

to those described above are currently common 

in the nuclear industry. However, the Mainte

nance Rule is driving efforts by licensees to 

implement many of the independent recommen

dations made in 1991 in Section 9 of 

NUREG-1275, Volume 6. Those recommenda

tions were: 

"* Take corrective actions to address the 

root-causes of SOV failures based on risk 

significance and determined from a plant

specific prioritization scheme; 

"* Review SOV design specifications, cal

culations, and operating conditions 

including temperature and pressure limi

tations; 

Conduct reviews to identify SOVs that 

may have been overlooked and to verify 

the orientation of SOVs; 

Replace or refurbish SOVs on a timely 

basis; and 

Ensure that the air systems that serve 

SOVs are a reliable source of clean, dry, 

oil-free air at proper pressure.

51 NUREG/CR-6654
12. AOV PROGRAM PLANS AT THE SITES VISITED

Six of the seven plants visited all had AOV 

action plans in varying stages of completion to 

address design basis determination or verifica

tion, calculation of margins, diagnostic testing to 

confirm margins, and maintenance. One plant 

was focusing on maintenance practices and 

assumed that the design basis operability, as 

defined in the original design is acceptable. 

The AOV action plans were comprehensive 

and indicate an awareness of the problems asso

ciated with AOVs, diagnostic tools available in 

the industry, and the work of various organiza

tions and utilities regarding AOVs. 

The licensees visited were aware of the work 

of the AOV Users Group, EPRI, NMAC, and 

ASME, among others. The scope and schedule 

for implementing program plans on AOVs, var

ied widely among the plants. 

Recent licensee efforts are being stimulated 

by their implementation of the Maintenance 

Rule, 10 CFR 50.65, recent industry-wide corre

spondence, a number of events involving AOVs, 

results from plant IPE and PRA investigations, 

and/or experience gained from the motor

operated valve programs established to meet the 

recommendations of Generic Letter 89-10. At 

the time of the plant visits, some licensees had 

not started these efforts while others were well 

on their way. A similar situation existed with 

regard to licensees' determination of the margins 

available for their AOVs to meet the design 

basis demands. It appears that improvements 

might include further aging assessments and 

consideration of additional failure mechanisms 

that the plants may not be considering now, in 

their assessments of the available margins in 

AOVs. 

12.1 Palo Verde Nuclear Gen

erating Station (PVNGS) 

AOV Program 

In March 1989, PVNGS Unit 3 suffered a 

loss of offsite power which resulted in the loss

of the air system. A comprehensive review of 

the air system design was subsequently per

formed and they concluded that a number of 

improvements with respect to the air system and 

associated equipment, including several types of 

AOVs, were needed to ensure the as-designed 

operability of the air system and the equipment 

served by it. A comprehensive "Compressed Gas 

System Analysis Report" was prepared 

(13-MS-A20, Revision 2, dated June 15, 1989, 

Revision 3 undated), and still serves as an active 

corrective action guidance document for the 

plant engineering staff concerned with the 

PVNGS pneumatic systems and equipment. 

The PVNGS action plan regarding AOVs 

was established based on their discovered need 

to improve AOV performance and included the 

following: 

Establish and maintain lists of Category I 

(safety-related with "active" safety func

tions) and II (safety-related without 

"active" safety functions, or ASME Sec

tion XI tested, or 10 CFR 50, Appendix J 

local leak-rate tested) AOVs. 

Evaluate and improve actuator margin for 

Category I and II AOVs, including 

dynamic testing of DCFWIVs and steam

generator blowdown isolation valves in 

open direction. Note that the tests of 

DCFWIVs were completed and valve 

factors of 0.64 and 0.57 were established 

in these tests. Refer to Section 8.1.3 in 

this study. 

Evaluate the need to develop procedures 

such as for diagnostic testing, trending, 

and revision/expansion of maintenance 

procedures. 

* Determine Category I and H AOVs. 

Develop a method to determine which 

Category II AOVs should be subjected to 

a sizing review, set point verification, and 

PM review.

53 NUREG/CR-6654
AOV Program Plans at the Sites Visited

"* Develop a method to determine which 

non-safety-related (NQR) AOVs should 

be subjected to a sizing review, set point 

verification, and PM review, and coordi

nate with the Maintenance Rule system 

AOVs. 

"* Perform design sizing evaluations of 

Category I AOVs. 

"* Revise the AOV Controlled Set Point 

Database. 

"* Evaluate uncertainties regarding actuator 

sizing for Category I and II AOVs. 

"* Evaluate possible permanent installation 

of test connections and transducer 

mountings on Category I AOVs. 

"* Determine a replacement for the (then) 

current AOV diagnostic system. 

"* Evaluate Category I AOVs to determine if 

current diagnostic equipment can be used. 

"* Determine scope of baseline and periodic 

test program for Category I AOVs. (No 

testing was proposed for Category II 

AOVs.) 

"* Review the maintenance program for 

Category I AOVs. 

"* Determine the post-maintenance require

ments for AOVs. 

"* Develop or revise database for AOV 

trending. 

"* Evaluate the quality class of LA compo

nents used on Category I AOVs. 

"* Review the AOV training program. 

"* Review the interface between the AOV 

program and the valve packing program. 

"* Verify the Safety Issue Management 

Systems (SIMS) database for AOVs.

"* Evaluate live-loading of the packing for 

throttling valves. 

"* Perform pressure-locking and thermal

binding evaluations for air-operated gate 

valves. 

"* Determine the necessary tasks for inclu

sion of hydraulic-operated valves in the 

AOV program. 

This program was being implemented and 

work on individual items was either completed 

or ongoing as of the time of the visit for this 

study. 

12.2 Fermi 2 AOV Program 

Fermi 2 is the lead BWR plant in an EPRI 

program to improve the operability and per

formance of AOVs in nuclear power plant appli

cations. The following quoted pilot program 

description was extracted from a Detroit Edison 

(owner of Fermi 2) Statement of Work 

(WO-5436-01): 

"Detroit Edison (DECO) is involved in an 

effort sponsored by the Electric Power Research 

Institute (EPRI) to develop an overall Air Oper

ated Valve Program document and, following 

the methods defined in the document, to conduct 

design basis system level (differential pressure, 

flow, temperature) and component level 

(required thrust/torque and actuator output capa

bility/margin) evaluations for their category I 

AOVs. Where applicable and with DECO con

currence, it is desired to apply the EPRI MOV 

Performance Prediction Methodology (PPM) to 

evaluate required thrust/torque and to utilize 

methods defined in EPRI Report TR-107321, 

'Application Guide for Evaluation of Actuator 

Output Capability for Air Operated Valves in 

Nuclear Power Plants' to evaluate actuator out

put capability.''a 

a. The Electric Power Research Institute (EPRI) MOV 

Performance Prediction Methodology (PPM), EPRI 

TR-103237, Revision 2, "MOV Performance Prediction 

Program Topical Report" was referenced in EPRI report 

TR-107321 for application to AOVs.

NUREG/CR-6654 54
The Detroit Edison Statement of Work 

described the objectives of the program as 

follows: 

"* "Develop an AOV Program Document 

which defines the overall approach and 

specific methods to be used in conducting 

system and component level design basis 

reviews of air operated valves." 

" "Determine the design-basis system 

parameters for input to AOV thrust/torque 

evaluations. These include stroke direc

tion, differential pressure, line pressures, 

flow, fluid media and temperature." 

"* "Determine thrust/torque requirements for 

41 AOVs selected by the Fermi 2 AOV 

Program staff under design basis flow and 

differential pressure conditions." 

"* "Determine the air actuator thrust/torque 

capabilities and margin available for 

AOVs selected ... above. Provide recom

mendations for corrective action cases 

where margin improvement is desired." 

DECO, Fermi 2, and their contractors were in 

the process of implementing the program plan at 

the time of the visit for this study. The categori

zation process was completed (see Table 1) and 

calculation efforts were underway. The original 

schedule called for the evaluations to be com

pleted by the end of 1997. 

The Fermi 2 AOV categorization process is 

closely tied to their efforts to satisfy the 

requirements of the Maintenance Rule, 

10 CFR 50.65. They are using risk-based 

techniques and analysis tools developed to 

implement the Maintenance Rule in their efforts 

to categorize AOVs in terms of risk and safety 

significance, as well as economic importance. 

Several AOVs were found to be much more 

important after such techniques and analyses 

were applied. As with all of the plants visited, 

the category definitions vary (see Table 1) 

between licensees. 

DECO is a member of a group of utilities 

known as the Utilities Service Alliance (USA

AOV Program Plans at the Sites Visited 

Group) that pool their resources to devise inte

grated solutions to problems and concerns. The 

plants in the USA Group were, at the time of the 

visit, Fermi 2, WNP 2, Palisades, Fort Calhoun, 

Clinton, Cooper, and Wolf Creek. AOVs are 

being studied at Fermi 2 as one of the USA 

Group efforts. 

12.3 Palisades AOV Program 

Palisades was the lead PWR in the EPRI 

program to improve the operability and per

formance of AOVs in nuclear power plant appli

cations and also was a member of the USA 

Group of utilities at the time of the visit. The 

Palisades program was described in their Proce

dure No. EM-28-03, dated September 26, 1997, 

generally as follows: 

"* "The purpose of this procedure is to pro

vide a systematic approach for addressing 

and resolving issues associated with all 

aspects of air operated valve (AOV) per

formance and to provide an auditable pro

gram plan for plant compliance and utili

zation of industry recommendations 

affecting AOVs.... This procedure also 

incorporates the reciprocal agreement in 

program methodology being shared with 

other USA member plants." 

"* The Background section of the Palisades 

AOV program plan referred to ASME 

OM-19 "Preservice and Periodic Per

formance Testing of Pneumatically and 

Hydraulically Operated Valve Assemblies 

in LWR Power Plants." Palisades 

intended to use the information in this 

(soon to be published as final) draft guide 

in developing their testing methods. 

"* The general description of the Palisades 

AOV program included the following: 

- "This program will be similar in 

content to the existing Palisades 

Motor Operated Valve Program 

which was implemented as part of 

NRC Generic Letter 89-10. The 

program is designed to capture the 

design basis of the system and

55 NUREG/CR-6654
AOV Program Plans at the Sites Visited 

establish a design database of ven

dor references, design calculations, 

and field settings for valves. The 

objectives of this plan are to 

improve the design basis knowl

edge of the AOVs, to support 

Operations and Maintenance in 

their various activities, and to 

improve the performance of the 

valves through maintenance or 

modifications based on design basis 

evaluations. Where appropriate, 

lessons learned from the MOV 

Program will be applied to the 

AOV Program. Information and 

data from the MOV program, 

where applicable, will be utilized as 

necessary to facilitate AOV 

evaluations." 

"GL 89-10 for MOVs identified the 

following basic problems in design 

and maintenance practices that are 

generally applicable to AOVs. 

Disc Factor for Gate Valves-It is 

now generally accepted that past 

sizing practices for actuators on 

gate valves may have resulted in 

underestimating thrust by a factor 

of 1.5 to 2.0. Recent industry gate 

valve testing supports this conclu

sion. 

Design Basis-The design bases 

for MOVs (were) found to contain 

incomplete information and/or non

conservative assumptions in a 

number of cases. Since AOVs and 

MOVs appear in the same systems, 

it is likely that this is a problem for 

AOV design bases as well. 

Control of Field Adjustments

MOV switch settings were not pro

cedurally controlled prior to the GL 

89-10 program. As a result, a num

ber of switch settings were found to 

be outside acceptable min/max 

ranges. AOV regulator and bench 

settings may also need to be procedurally controlled and packing 

forces verified to ensure they are 

within design assumptions. The 

program will define the necessary 

testing requirements to ensure the 

valves can perform their design 

function." 

"* "The goal of this program is to economi

cally verify that AOVs providing active 

safety-related functions and balance-of

plant AOVs which are critical in terms of 

cost, radiation exposure, and reliable 

power generation, perform as required. 

All valves in this program will be catego

rized based on safety significance, PSA, 

an importance to plant availability and 

maintenance history. Design basis review, 

testing and corrective action (are) then 

performed on valves prioritized above a 

plant defined action threshold." 

"* The scope section indicated that HVAC 

dampers were excluded from the AOV 

program. AOVs were then categorized as 

1, 2 or 3 (see Table 1). 

"* The Palisades AOV program scope that 

was obtained during the site visit divided 

AOVs into 3 categories as follows: 

- "Category 1: Valves in this cate

gory are safety-related with active 

safety functions, are important to 

safety based on their PSA risk sig

nificance, or are included based on 

experience from Expert Panel 

meetings. Category 1 AOVs 

require a documented design basis 

review and setpoint verification 

periodically confirmed by diagnos

tic testing." 

- "Category 2: Valves in this cate

gory may be safety-related with 

low risk significance for PSA or 

nonsafety-related in critical appli

cations that could affect plant 

availability, capacity factor, heat 

rate, or have high maintenance 

costs. Category. 2 AOVs may

NUREG/CR-6654 56
receive design reviews and/or diag

nostic testing on an as-need basis at 

the discretion of the AOV Program 

Engineer. Any AOV with a Main

tenance Preventable Functional 

Failure on record in the Mainte

nance Rule Program shall be con

sidered for inclusion in the AOV 

Program as a Category 2 AOV by 

the AOV Program Engineer." 

"Category 3: Valves in this cate

gory are the remaining air operated 

valves not in Categories 1 or 2. 

Design basis reviews or diagnostic 

testing are not performed for Cate

gory 3 AOVs." 

"* The design basis review was described 

and consists of a system level review and 

a component level review. 

"* The system level review identifies the 

design basis conditions under which the 

AOV must operate and includes: 

- Line pressure (both upstream and 

downstream) 

- Fluid media 

- Fluid flow (only if required) 

- Fluid temperature. 

" A component level review section 

described requirements for AOV actuators 

in order to meet worst-case system 

demands and includes a list of elements to 

be considered, as follows: 

- Minimum air pressure required 

- Maximum air pressure allowed 

- Diaphragm air 

- Required bench set 

- Spring rate

AOV Program Plans at the Sites Visited 

- Seat load 

- Valve factor (for gate valves if 

PPM is not used) 

- Minimum required thrust/torque 

- Stroke length 

- Seat diameter 

- Valve trim (includes valve dimen

sions and balanced/unbalanced 

areas) 

- SOV characteristics/ratings 

- Packing load assumptions 

- Flow direction (for globe valves 

either under or over seat) 

- Shaft orientation (upstream or 

downstream for butterfly valves) 

- Maximum spring allowable force 

- Butterfly valve bearing material 

and coefficient of friction 

assumptions 

- Piston breakaway force. 

"* The EPRI Motor Operated Valve Per

formance Prediction Methodology (PPM) 

was referenced for predicting torque/ 

thrust requirements without the need for 

dynamic testing. The PPM methods are 

considered to be directly applicable to 

gate and butterfly AOVs; however, the 

PPM methods may be of limited 

applicability for many globe/plug valves 

because of the wide variety in use in the 

nuclear industry compared to the types 

and small population tested to establish 

the PPM methodology. 

"* Diagnostic testing methods are to be used 

to establish baselines for periodic moni

toring and post-maintenance testing.

57 NUREG/CR-6654
AOV Program Plans at the Sites Visited 

"* Corrective and preventive maintenance is 

to be coordinated with the requirements of 

the Maintenance Rule. 

"* Criteria for AOV modifications are dis

cussed, as well as pressure locking and 

thermal binding, SOV evaluations, and air 

regulator setting guidelines. 

12.4 LaSalle AOV Program 

The AOV program at LaSalle was described 

in a draft administrative procedure dated 

December 1997 just prior to our visit on 

December 17-18, 1997. The stated purpose is to 

define the AOV program, identify the interfaces 

and provide a comprehensive approach to verify 

that the valves covered in the scope of the pro

cedure are capable of performing their intended 

function under normal, abnormal, or emergency 

operating design basis conditions. 

The procedure included the following provi

sions: 

"* Systematically review AOV requirements 

"* Categorize AOVs based on their safety 

significance and functional requirements 

"* Prioritize AOVs based on past perform

ance and industry data 

"* Establish AOV design bases by reviewing 

design data 

"* Determine as-built configurations by per

forming walkdowns 

"* Track and trend data. 

AOVs are to be categorized as 1, 2, 3 or 4 

(see Table 1) and the categorization process is 

coordinated with the requirements of the Main

tenance Rule. 

We were informed that LaSalle intended (as 

of December 1997) to rely on diagnostic testing 

to ensure the operational readiness of AOVs. 

They had done many static diagnostic tests but 

had not tested any AOVs under dynamic

conditions. The method(s) by which they 

intended to establish the available margins with 

respect to design basis loadings of AOVs was 

unclear as of the time of the site visit. 

12.5 Three Mile Island I AOV 

Program 

The scope of the draft AOV program at 

TMI-1 included the following: 

"Identify and categorize the population of 

AOVs into specific groups. The valve 

category will define the requirements for 

each valve in the group based on risk 

ranking. The techniques used to imple

ment the requirements of the Maintenance 

Rule are being used in this process. AOVs 

are categorized as 1, 2 or 3, depending on 

their safety and economic significance. 

See Table 1. 

"* Determine the limiting fluid system oper

ating conditions under which each valve 

in Categories 1 and 2 must operate, 

including design basis accidents, safe 

plant shutdown, and normal/abnormal 

operating conditions (system level design 

basis review). 

"* Develop and implement a methodology to 

determine valve thrust/torque require

ments and air actuator thrust capabilities 

(component level design basis review). It 

appeared from the program text that this 

methodology will be applied to 

Category 1 and 2 AOVs. 

"* Evaluate the design capability against the 

valve requirements and initiate any neces

sary valve or actuator modifications or 

adjustments. 

"* Adopt a validated test methodology with 

known accuracy characteristics that is 

compatible with the different valve design 

characteristics. 

"* Develop procedures and provide training 

for using the adopted test methodology.

NUREG/CR-6654 58
"* Upgrade and improve the periodic main

tenance and test program to ensure long

term AOV operability. This includes the 

provision for a qualified review of the test 

data for valve operability. 

"* Continue to trend and evaluate AOV fail

ures, preventive maintenance results and 

diagnostic test data. When appropriate, 

preventive maintenance task frequencies 

will be revised to address the findings. 

The AOV program is to be coordinated with 

the requirements of the Inservice Testing Pro

gram mandated in 10 CFR 50.55a. The draft 

AOV program description (GPUN Topical 

Report 118) included an appendix that described 

the system level design basis reviews and an 

attachment that provided samples of the design 

basis reviews for individual AOVs. 

12.6 Indian Point 3 (IP3) AOV 

Program 

The IP3 AOV program was initiated primar

ily because the plant AOV failure rates were 

averaging about twice the industry AOV failure 

rate over the previous few years. The purpose of 

the program was described as "to improve plant 

safety, reliability, and efficiency." 

An action plan was developed, the key ele

ments of which included: 

Identify the key elements of a successful 

AOV program 

* Categorize AOVs 

"* Perform design basis reviews (both sys

tem level and component level) 

"* Validate actuator sizing 

"* Make program recommendations based 

on these reviews. 

The program was divided into two phases, as 

follows:

AOV Program Plans at the Sites Visited 

"* Phase 1, Concentrate on AOVs to be 

maintained during Refueling Outage 

No. 9 

"* Phase 2, Address remaining AOVs. 

Phase I included the following: 

"* Twenty AOVs were selected based on 

safety-related or plant efficiency/thermal 

performance considerations. 

"* The scope of work included overhauls, 

engineering reviews, and diagnostic test

ing. 

"* A review of the Phase 1 results was con

ducted to verify the program and provide 

insights for necessary modifications. 

About half of the 20 AOVs tested were found 

to have problems, some of them long-standing, 

that needed correction. Corrections made, based 

on diagnostic testing methods, were considered 

successful. The majority of problems involved 

packing, seating/unseating, and positioner cali

bration. 

Phase 2 implementation included: 

Conduct regular meetings with the AOV 

peer group 

Complete and verify a list of all safety

related AOVs to ensure that none are 

excluded from the evaluation 

Categorize all AOVs for the program 

based on PRA significance, active safety 

function, effect on plant availability, 

effect on plant thermal efficiency, 

maintenance history, and operational 

problems 

Perform system and component level 

reviews on Phase 2 AOVs, validate 

actuator sizing requirements, make rec

ommendations regarding scope and fre

quency of testing/maintenance, and 

evaluate training procedures.

59 NUREG/CR-6654
AOV Program Plans at the Sites Visited 

The stated goal was to complete the program 

by January 1999 and implement Phase 2 during 

refueling outage 10 on the full scope of AOVs. 

12.7 Turkey Point 3 and 4 AOV 

Program 

The AOV program at Turkey Point was initi

ated in 1991 as a maintenance initiative and 

focused on procedures, diagnostics, and preven

tive maintenance. Since that time, and based to 

some extent on compliance with the Mainte

nance Rule, AOVs at Turkey Point have been 

categorized in accordance with their risk and

safety significance (see Table 1). Turkey Point 

now employs diagnostic testing, troubleshoot

ing, and AOV history as tools to guide preven

tive and corrective maintenance. An extensive 

training effort is used to ensure that technicians 

can diagnose and repair AOV problems. 

At the time of the visit, Turkey Point had no 

plans to further investigate the design bases vs. 

available margins for AOVs. They planned to 

rely on the information originally furnished by 

the vendors and considered that their primary 

focus should be on maintenance of their 

equipment.

NUREG/CR-6654 60
13. INDUSTRY INITIATIVES REGARDING AOVS

The nuclear power industry is aware of the 

significance of AOV operability as a logical 

extension of their work to ensure the design 

basis operability of motor-operated valves. They 

are also aware of the significance of AOVs as a 

result of the work done to prioritize the impor

tance of equipment in nuclear power plants, in 

order to comply with the requirements of the 

Maintenance Rule. A number of industry 

organizations and individuals are working on 

these issues. 

Key points of interest in the nuclear power 

industry regarding assessing margins for AOVs 

include: 

The Electric Power Research Institute 

(EPRI) MOV Performance Prediction 

Methodology (PPM), EPRI TR-103237, 

Revision 2, "MOV Performance Predic

tion Program Topical Report" was refer

enced in EPRI Report TR-107321, 

"Application Guide for Evaluation of 

Actuator Output Capability for Air Oper

ated Valves in Nuclear Power Plants." 

The PPM methodology (originally devel

oped as part of industry efforts regarding 

MOVs) was endorsed by EPRI as pro

viding guidance for predicting the forces 

needed to operate globe, gate, and butter

fly valves for at least some AOVs. The 

PPM methods are considered to be 

directly applicable to gate and butterfly 

AOVs; however, they may not be appro

priate for many globe/plug valves because 

of the wide variety in use in the nuclear 

industry compared to the types and small 

population tested to establish the PPM 

methodology. 

NRC Information Notice 96-48 included a 

summary of important contributions and 

findings resulting from the EPRI MOV 

PPM. Important findings (or confirmatory 

information) from the EPRI MOV Pro

gram included the following: 

a. The traditional methods for pre

dicting gate valve performance

might not be conservative for many 

applications because of incomplete 

equations, design features, manu

facturing controls, and wide-rang

ing friction coefficients. 

b. The edge radii on disk seats and 

guide slots are critical to gate valve 

performance and predictability. 

c. Stellite friction coefficients 

increase with differential-pressure 

valve strokes in cold water to a 

plateau level, stabilize quickly in 

hot water, and decrease as differ

ential pressure increases. 

d. Gate valves with carbon steel 

guides and disk guide slots with 

tight clearances might fail to close 

under blowdown conditions. 

e. Many existing gate valve manu

facturing and design processes and 

controls, and plant maintenance 

practices, might contribute to poor 

valve performance. 

f. Traditional methods for predicting 

globe valve performance for 

incompressible flow conditions are 

not conservative for globe valves in 

which differential pressure acts 

across the plug guide. 

g. Globe valve thrust requirements for 

some designs can be excessive 

under compressible flow and blow

down conditions because of the 

potential for plug-side loading. 

h. Rate-of-loading effects (load-sen

sitive behavior) can reduce the 

static thrust output by up to 30 per

cent under dynamic conditions. 

i. Hydrodynamic torque coefficients 

used by some butterfly valve manu

facturers might not be conservative

61 NUREG/CR-6654
Industry Initiatives Regarding AOVs 

for certain applications, with valves 

located near piping elbows espe

cially vulnerable. 

j. Butterfly valve seats should be 

periodically replaced to avoid hard

ening or degradation. 

In addition to these reported important find

ings, EPRI confirmed that thrust requirements to 

unwedge a gate valve can be higher under 

dynamic conditions than under static condi

tions." 

"* With the exception of item h above, 

which refers to a characteristic that may 

be unique to motor operators, the findings 

quoted from IN 96-48 are applicable to 

AOVs using the same types of valves. 

However, note that MOVs include large 

populations of gate valves while AOVs 

include large populations of globe valves. 

"* Thrust requirements for some AOVs may 

be greater than originally assumed in ven

dor calculations and some plants have yet 

to address this issue. 

"* Valve packing friction force estimates 

may not be conservative for graphite 

packing or for packing modifications 

made to reduce leakage. 

"* Calculated effective diaphragm areas used 

by some vendors were found to be non

conservative. 

EPRI contracted with Detroit Edison 

(Fermi 2) and Consumers Energy (Palisades) as 

part of the development of an overall AOV 

program document. Based on the (draft) EPRI 

AOV Program document, a pilot program was 

being prepared which included design basis 

system level tests (differential pressure, flow, 

and temperature) and component level tests 

(required thrust or torque and actuator output 

capability or margin) for Category 1 (safety

related and/or highly risk significant) AOVs. 

The PPM, originally prepared in response to 

MOV issues, and TR-107321 are to be used in 

the evaluations of AOVs.

The Air-Operated Valve Users Group (AUG) is 

an organization dedicated to improved perform

ance of AOVs. They have held 17 semi-annual 

meetings so far. The 14th and 16th semi-annual 

meeting, in December 1997 and December 1998 

respectively, were held jointly with the Motor

Operated Valve Users Group. The focus areas of 

the AUG were: 

"* Provide a forum for exchange of technical 

information for the industry on AOV 

issues. 

"* Support the member's regulatory interest 

by providing reliable guidance on regu

latory issues, as well as striving to achieve 

an equitable balance between safety and 

efficient plant operation. 

"* Provide technical expertise for problem 

identification and resolution, implement

ing advanced technologies, and technical 

direction. 

"* Develop products and services that allow 

members to operate in the most safe, reli

able, and cost-effective manner. 

A Joint Owners Group comprised of the 

B&W Owners Group, CE Owners Group, BWR 

Owners Group, Westinghouse Owners Group, 

and 104 operating U.S. nuclear power plants was 

formed to address issues involving AOVs (and 

is referred to herein as the JOG AOV). In 

November of 1997 the JOG AOV agreed to: 

"* Develop an AOV program document 

which encompasses all aspects necessary 

to ensure the required function of AOVs; 

"* Utilize existing, ongoing, and planned 

industry efforts to the extent possible in 

the development of the program; and 

"* Produce a document which would be suit

able as guidance for all plants to utilize in 

developing and maintaining an acceptable 

AOV program. 

The JOG AOV described their efforts to the 

NRC in a presentation on June 3, 1999, at NRC

NUREG/CR-6654 62
headquarters. NEI forwarded a copy of the JOG 

AOV Program (Revision 0, dated March 9, 

1999) to the NRC in a letter from D. Modeen to 

E. Imbro dated July 19, 1999. In the letter NEI 

stated that "(w)ithin the industry there is broad 

recognition that AOV design configuration, 

operation, testing, and maintenance are impor

tant factors in safe, reliable, and efficient plant 

operation." NEI concluded: 

"Industry experience and various pub

lished reports do not indicate safety-sig

nificant AOV concerns that warrant 

industry action. The industry actions 

described above provide ample evidence 

that industry is addressing any AOV per

formance issues. Given this perspective, 

these industry activities are not a topic 

that the industry desires credit for in the 

context of SECY 99-063, The Use by 

Industry of Voluntary Industry Initiatives 

in the Regulatory Process. Consequently, 

we are not requesting NRC review or 

endorsement of the enclosed document." 

Observations and comments regarding the 

JOG AOV Program (Revision 0, dated March 9, 

1999) are as follows: 

"* "CURRENT DOCUMENT STATUS," 

page iib 

No formal implementation action is being 

recommended with issuance of the JOG 

AOV Program Document, Revision 0. 

The text indicated that "...individual util

ity executive implementation directions 

are provided at some future time." No 

schedule recommendations to licensees 

were included to implement the JOG 

AOV Program. 

"* Section 1.3, "AOV Program Elements," 

page 2 

b. Section numbers and page numbers refer to those found 

in the JOG AOV Program plan.

Industry Initiatives Regarding AOVs 

"Nine key elements for an AOV program 

are identified (in the JOG AOV Program) 

as follows:

Scoping and Categorization 

Setpoint Control 

Design Basis Reviews 

Testing 

Preventive Maintenance 

Training 

Feedback 

Documentation/Data Management 

Tracking and Trending of 

Performance"

AOV

The elements appear to be comprehensive 

and include the necessary scope. 

"* Section 1.5, "Instrument Air Systems," 

page 2 

It is critically important that licensees 

provide reliable supplies of clean, dry, oil

free air at the proper pressure, to ensure 

the successful operation of AOVs. 

"* Section 2, "DEFINITIONS," page 3 

The stated definition of an "active valve: 

a valve that must perform a mechanical 

motion during the course of accomplish

ing a system safety-significant function" 

appears to preclude consideration of 

mispositioning as a consideration in the 

JOG AOV program. Refer to the remarks 

about Section 4.1.4 of the JOG AOV pro

gram below. 

The stated definition of "high safety sig

nificance: designation referring to the 

importance to plant safety by a blending 

process of risk ranking and expert panel 

evaluations," includes risk ranking and 

expert panel opinion as the only criteria, 

without reference to the contents of the 

existing plant design basis. 

The term "system safety significant func

tion" is not defined but is used in several 

of the other definitions.

63 NUREG/CR-6654
Industry Initiatives Regarding AOVs 

" Section 3, "PROGRAM REQUIRE

MENTS," page 4 and Table 3-1, page 5 

Design basis reviews, baseline testing, 

periodic testing and post-maintenance 

testing are not required in the JOG AOV 

Program for Category 2 AOVs even 

though these valves are either safety

related or have high safety-significance. 

This appears to be a relaxed standard for 

such important equipment. 

"* Section 4.1.2, "Scope," page 6 

Dampers are specifically excluded from 

the program, based on similar treatment 

by the industry for motor-operated damp

ers. Air-operated dampers provide several 

critical functions such as control room 

ventilation, containment protection, diesel 

generator protection, and diesel generator 

air supply. 

"* Section 4.1.3, "Categorization Process," 

page 6 

AOVs within the scope of the JOG AOV 

Program are classified into two categories 

as follows: 

Category 1: AOVs that are safety-related, 

active, and have high safety significance. 

Category 2: AOVs that are safety-related, 

active, and do not have high safety 

significance, or AOVs that are non-safety

related, have high safety significance, and 

are active. 

The JOG AOV Program indicates that 

"AOVs not in Categories 1 or 2 are con

sidered outside the scope of this (the JOG 

AOV) program, as they are deemed not to 

be critical to plant safety." Any valve 

could be excluded from Category 1 based 

on expert panel opinion or findings of low 

risk ranking (see next comment). Further, 

only active valves, as defined in accor

dance with the terminology of the pro

gram document, are included in either

Category 1 or 2 and thus subject to any 

requirements under the JOG AOV 

program. 

"* Section 4.1.3, "Categorization Process," 

page 6 

Common-cause failures are not taken into 

consideration in the JOG AOV Program 

when ranking the risk-based importance 

of AOVs. Individual AOVs may be found 

to be of low risk significance, but the 

common-cause failure (CCF) of two or 

more AOVs performing the same function 

may have considerable risk significance. 

(Refer to the last paragraph in Section 

15.4 of this study for specific examples.) 

The JOG AOV Program does not refer to 

this issue 

"* Section 4.1.3.2, "Determination of Safety 

Significance," page 7 and 8 

Five specific references are included in 

the paragraph, "any one" providing an 

acceptable method for ranking safety sig

nificant AOVs. A conservative approach 

would be for each licensee to consult at 

least two or three of the five, rather than 

relying on one as the paragraph suggests, 

especially considering the sentence at the 

top of page 8. 

"* Section 4.1.4, "Mispositioning," page 8 

Mispositioning or inadvertent operation of 

an AOV is specifically excluded from the 

JOG AOV Program. It is yet to be dem

onstrated, on the basis of safety signifi

cance, that mispositioning of AOVs is a 

consideration that can be ignored. 

"* Section 4.3, "Design Basis Reviews," 

pages 10-15 

The JOG AOV Program document 

includes detailed instructions for con

ducting design basis reviews. Several spe

cific references to methods developed in 

response to Generic Letter 89-10 pro

grams for MOVs are included. In general,

NUREG/CR-6654 64
Industry Initiatives Regarding AOVs

design basis reviews are only required for 

Category 1 AOVs.

Section 4.3.3.2, 

Capability, page 14

"Actuator Output

This section refers to EPRI TR-107321, 

"Application Guide for Evaluation of 

Actuator Output Capability for Air Oper

ated Valves in Nuclear Power Plants," as 

providing acceptable first principle equa

tions for evaluation of actuator output 

capability for AOVs. EPRI TR-107321 

references the EPRI MOV Performance 

Prediction Methodology (PPM), EPRI TR

103237, for application to AOVs. 

Extrapolation of the methodology, originally 

developed in response to issues involving 

MOVs, may be acceptable; however, the 

range and conditions of applicability remain 

to be examined and verified. 

" Section 4.4, "Testing," pages 15-19 

Baseline, periodic, and post-maintenance 

testing methods are described in some 

detail but testing, including post

maintenance testings, is restricted to 

Category 1 AOVS. 

" Section 4.9, "Tracking and Trending," page 

21 and Section 5, "FULL PROGRAM 

IMLEMENTATION," page 22 

Tracking and trending did not include 

establishing and maintaining a history of 

the program, including changes and 

updates. 

Previous activities of a JOG for motor

operated valves involved production of an 

industry topical report for periodic verification 

of design-basis motor-operated valves. This 

report was assembled to meet the recommended 

actions in NRC Generic Letter 96-05, "Periodic 

Verification of Design-Basis Capability of 

Safety-Related Motor-Operated Valves."

The ASME Committee on Operations and 

Maintenance Activities developed a guideline 

(to have been published as a standard) for 

AOVs, ASME OM Part 19, "Preservice and 

Periodic Performance Testing of Pneumatically 

and Hydraulically Operated Valve Assemblies in 

LWR Power Plants." OM-19 is a set of 

voluntary applied guidelines which apply to 

active safety-related AOVs and can be applied to 

other AOVs at the user's option. In the draft 

that was reviewed for this study (October 18, 

1996), preservice testing is recommended 

including consideration of dynamic loads. 

Several testing options are offered including 

dynamic testing at service conditions. Periodic 

stroke/timing testing and other tests include 

measurement of parameters such as bench set, 

maximum available pneumatic pressure, seat 

load, spring rate, actual travel, pneumatic 

pressure required to accomplish safety function, 

and friction forces. Provisions for analysis and 

evaluation of data as well as corrective action 

criteria are included. 

13.1 The Maintenance Rule 

The requirements for "Monitoring the 

Effectiveness of Maintenance at Nuclear Power 

Plants," were invoked in 10 CFR 50.65. The 

implementation of these requirements is being 

accomplished in accordance with NRC 

Regulatory Guide 1.160, which endorsed an 

industry document, NUMARC 93-01. The 

Maintenance Rule does not deal with specific 

maintenance requirements, but rather with plant 

and system performance which demonstrates 

that nuclear power plant maintenance activities 

are ensuring acceptable performance of 

structures, systems, and components. 

Licensees are addressing how effective their 

maintenance practices are with respect to the 

operability of safety-related and important non

safety-related systems and equipment in their 

plants using a variety of risk related tools and 

techniques. The licensees are ranking systems 

and equipment in terms of risk significance in 

order to get improved effectiveness out of their 

maintenance activities as well as improved 

safety. The system of establishing goals for

65 NUREG/CR-6654
Industry Initiatives Regarding AOVs 

systems and equipment found to have opera

tional problems appears to be effective (at least 

as far as AOVs are concerned) in the plants 

visited. Its effect (and perceived effectiveness) 

was noted in all of the plants visited. 

Licensees at the plants visited are using tech

niques developed in response to the Maintenance Rule to assess the performance of equip

ment outside, and in addition to, the scope of 

NRC's interest in order to increase the reliability 

and efficiency of their plants. As a result of their 

Maintenance Rule activities, licensees became 

increasingly aware of the importance of AOVs, 

in terms of plant safety, availability, and heat 

rate.

NUREG/CR-6654 66
14. DIAGNOSTIC SYSTEMS, MAINTENANCE, 

AND TESTING OF AOVS

The rapid evolution of computerized diag

nostic systems for evaluating the condition and 

performance of valves in nuclear plants provided 

an essential tool to the nuclear power industry. 

All nuclear plants have used and are using such 

diagnostic equipment to comply with the rec

ommendations of NRC Generic Letter 89-10 and 

its supplements. Diagnostic equipment for 

evaluating the condition and performance of 

AOVs has been available since the mid-1980s 

and evolved in parallel with computer software, 

computer hardware, testing equipment, and ven

dor experience. Recently, diagnostic system 

vendors who were primarily focused on provid

ing diagnostic systems for motor-operated 

valves turned their attention to AOVs and 

released state-of-the-art products specifically 

designed for AOVs. At least one diagnostic 

system on the market was specifically designed 

for SOVs. 

The AOV diagnostic systems are used to 

monitor, record, and provide numerical and 

graphical output for a number of measured AOV 

parameters. These parameters include total 

travel, signal times incidental to travel, friction, 

bench set (the high and low pressure values 

applied to a single-acting actuator to produce the 

nominal valve travel with no external forces on 

the actuator stem), spring rate, seat load, signal 

pressures at various points of travel, positioner 

parameters, I/P (current-to-pressure) or E/P 

(voltage-to-pressure) transducer signal parame

ters, and system air pressure variations. 

AOV diagnostic systems are extremely use

ful and powerful tools for preventive or predic

tive maintenance. In all of the plants visited, the 

engineers were aware of what is available and 

several were actively seeking to upgrade their 

existing diagnostic system capabilities. The 

diagnostic system vendors and users have 

become adept at diagnosing a wide variety of 

valve and operator defects and deteriorated 

conditions.

The problem of predicting, with confidence, 

design basis behavior of a valve from informa

tion gathered from no-load testing (no pressure 

and/or fluid flow in the line) remains. The EPRI 

valve performance prediction methodology 

(PPM) and the included data are offered by 

EPRI for predicting the forces needed to operate 

globe, gate, and butterfly AOVs at design-basis 

conditions. Refer to EPRI TR-103237 as 

endorsed in EPRI report TR-107321. Extrapola

tion of the methodology, originally developed in 

response to issues involving MOVs, may be 

acceptable; however, the range and conditions of 

applicability remain to be examined and veri

fied. Data from recent or planned flow or pres

sure testing (EPRI or other sources and individ

ual plant tests) could be useful in verifying the 

assumptions used in the PPM. 

As was the case originally with motor

operated valves, the claims of the AOV and 

SOV diagnostic system vendors regarding accu

racy and abilities of their systems remain to be 

verified. Calculated estimates of packing loads, 

side loads, and other parameters should be veri

fied to be within realistic ranges. The relation

ship between diaphragm pressure and stem 

thrust delivered to the valve disc or plug should 

be confirmed to be as assumed. 

The currently required test for most safety

related AOVs in nuclear power plants consists of 

a periodic stroke-timing test in accordance with 

ASME Code Section XI. The test consists of 

operating the valve with no required pressure or 

flow in the lines and measuring the time 

required to stroke in the safety-related direc

tion(s). Records are kept for trending purposes 

and to satisfy NRC documentation requirements. 

The stroke-timing test by itself does not provide 

much in the way of assurance of future oper

ability. AOVs can and have been successfully 

stroke-time tested and then failed on the next 

test or the next service demand. The use of diag

nostic systems to identify AOV part failures or

67 NUREG/CR-6654
Diagnostic Systems, Maintenance, and Testing of AOVs

misadjustments can provide for enhanced reli

ability. However, the problem of predicting 

design basis operability of safety-related AOVs 

from a no-load test remains, if not supplemented 

by knowledge of the design basis capability and 

demand. 

The plants visited have a number of AOVs 

(including dampers) associated with safety

related, important non-safety-related, or risk sig

nificant systems, such as the diesels or HVAC

systems, that are not maintained under the con

trol of the AOV engineer(s). These AOVs are 

serviced by the technicians and engineers that 

service the particular system. These AOVs may 

not receive the same level of attention relative to 

engineering, maintenance, and testing, or the 

benefit of safety-related performance informa

tion available to personnel more closely 

involved with the design, maintenance, and per

formance evaluation of AOVs within each 

plant's AOV program.

NUREG/CR-6654 68
15. RISK ANALYSIS INVOLVING AOVS

15.1 Previous Risk Analyses 

Results 

Several risk analyses have been conducted 

concerning air systems and AOVs. 

NUREG-1275, Volume 2, Section 6.4 refers to 

three risk studies regarding instrument air 

systems' relation to nuclear, plant safety. 

NUREG-1275, Volume 6, Section 8.1 includes a 

discussion of the risks associated with common

cause failures of SOVs. 

A study produced for the industry by 

Pickard, Lowe and Garrick, Inc., NSAC-128, 

"Pneumatic Systems and Nuclear Plant Safety," 

October 1988 offered the following findings: 

"The safety significance of support sys

tems (especially pneumatic systems) is 

not always obvious. Most plant analysis 

and documentation (FSAR, technical 

specifications, emergency procedures) is 

produced by reactor vendors and focuses 

on the front line systems they design. In 

PRA analysis and maintenance records, 

support system failures are often catego

rized as failures in the affected front line 

system." 

* "Contamination has far more impact on 

pneumatic system performance than does 

loss of pressure. System designers provide 

highly reliable, redundant air supply sys

tems, often backed up by nitrogen bottles, 

and then add another layer of protection 

by installing valves that fail-safe on loss 

of air. On the other hand, they generally 

provide neither redundancy for preventing 

contamination nor warning against its 

occurrence or effects. The effects of con

tamination are unpredictable: sometimes 

failures occur immediately; in other cases, 

contamination causes multiple failures in 

air operated equipment at random times 

long after the contamination has been 

released. Contamination has the potential 

to defeat all redundancy in supply."

"* "There is no way to predict exactly what 

sequence of events will occur during 

gradual degradation of air pressure and no 

assurance that, if it happens twice at the 

same plant, the sequences will be the 

same." 

"* "Air system capabilities and require

ments, as well as the extent and signifi

cance of air system problems, are not 

always well understood by plant staffs. 

System capacity, dryer lifetime, and the 

effects of degraded desiccant are some

times not known. Symptoms, rather than 

root causes, are often corrected." 

"* "The extent of reliance on pneumatic 

systems for safety and operational func

tion is plant specific." 

"* "Plant availability is more likely to be 

affected by air system problems than is 

plant safety." 

"More extensive lists of findings and rec

ommendations found elsewhere (refer

ences in quoted text are omitted) tend to 

obscure the significance of contamination 

stressed (in NSAC-128)." 

In a 1989 study, NUREG/CR-5472, "A Risk

Based Review of Instrument Air Systems at 

Nuclear Power Plants," the researchers studied 

the IA systems in nuclear power plants in terms 

of their contribution to risk. The overall Conclu

sions and Recommendations section (5) of that 

report included the following paragraphs: 

"A systematic review of IA-related events, 

system designs, and risk impacts was performed. 

Although many events related to the IA system 

have been reported, there is neither such a plu

rality of events nor do the events place a typical 

plant in such danger of core damage or signifi

cant release of radioactivity that treatment of the 

IA system should be significantly revised. This 

study yielded three general conclusions:

69 NUREG/CR-6654
Risk Analysis Involving AOVs

1. The IA system contribution to total core 

melt frequency is generally much lower 

than that of frontline systems, and is sig

nificantly lower at BWRs than at PWRs. 

2. The risk contribution of the IA system 

cannot be significantly reduced by modi

fications or reliability improvements 

within the IA system. 

3. Most plants which had notable IA-related 

risk sequences needed modifications out

side the IA system (e.g., the condensate 

system at Oconee, and the HPI system at 

Haddam Neck)." 

"However, risk and reliability analyses that 

have systematically considered the IA system, 

its interactions with frontline systems, and the 

effect of loss of IA on the plant have uncovered 

plant-specific operating and design weaknesses 

that impact risk. The following conditions 

increase the risk impacts of the IA system: 

"* The possibility of common cause failures 

of air-operated equipment (e.g., Calvert 

Cliffs); 

"* Unique designs in fail-safe valve posi

tions (e.g., Oconee); 

"* Contamination of the air system such that 

the common-cause failure probabilities of 

air-operated components are significantly 

increased (e.g., Turkey Point) 

"* Accumulator and associated check valve 

reliabilities especially those not ade

quately or frequently tested; (and) 

"* EDG dependencies on IA during an actual 

(loss of offsite Power)." 

"The following actions can ensure that IA 

system contributions to plant risk remain low: 

1. Ensuring that appropriate standards of 

design quality (moisture, particulate size), 

design intent (compressor capacity, backup

sources of air), and operational performance 

(minimization of maintenance-related and 

other human errors) are maintained. 

2. Including the IA system in risk-based 

review of plant systems (e.g., PRAs) and, 

when risk sequences are quantified using 

an estimate of the frequency of loss of IA 

that reflects the generic frequency and 

nature of problems in the system. 

3. Locating and correcting any EDG/IA 

interactions in which non-safety grade 

portions of IA can cause EDGs to fail 

during a LOOP (loss of offsite power). 

Such a review would include identifica

tion and elimination of diesel room 

dependence for cooling on systems that 

are off-line during a LOOP. 

4. Ensuring that the design and functionality 

of accumulators is consistent with safety 

analyses. 

5. Prior to making changes to solenoid 

valves and/or air pressure regulators(,) 

perform (an) analysis that takes into con

sideration potential overpressurization 

problems. Follow changes by a test to 

(ensure) that the fail-safe conditions of 

AOVs are not being compromised." 

The situation at LaSalle, a BWR, where loss 

of instrument air is the largest initiating event, 

contributing almost one-third to core-damage 

probability (refer to the trip report in Appendix 

C), is at odds, for at least one plant, with the in 

NUREG/CR-5472 that IA system contribution 

to core melt is generally much lower than for 

frontline systems. Observations from the site 

visits, particularly Palisades and Three Mile 

Island, are not in agreement with the conclusion 

in NUREG/CR-5472 that the risk contributions 

of the IA system cannot be reduced by 

improvements to the system. The other conclu

sions and recommendations are generally in 

agreement with this current study. 

Many of the details identified as sources of 

concern in this current study of AOVs were

NUREG/CR-6654 70
Risk Analysis Involving AOVs

described in NUREG/CR-5472 (sections therein 

noted in parenthesis), such as: 

"* A significant reduction in risk due to 

common-cause events could be achieved 

by reducing instrument air contamination. 

(2.4.4) 

"* An important safety concern is the ability 

of AOVs to achieve the fail-safe position. 

Moisture contamination of the IA system 

is an important common-cause event from 

a risk perspective. (3.2) 

"* The IA system affects PRA models in 

somewhat the same way as does the elec

tric power system. The IA system sup

ports frontline components, both in nor

mally operating and standby systems, and 

a loss of IA causes a reactor trip. IA uses 

accumulators as a back-up source of 

motive force to important components. 

These accumulators function in a manner 

analogous to that in which emergency AC 

power is applied to important loads upon 

loss of normal AC power. (4.) 

"* In all PRAs reviewed, IA was discussed 

qualitatively both as a potential initiating, 

and as a support system. However, most 

PRAs did not treat loss of IA as a unique 

initiating event because it is not regarded 

as being significantly different from the 

other events that cause a loss of the PCS 

and a reactor trip. (4.) 

"* The significance of relying on an air

pressure regulator that may be non-safety

related to prevent the over-pressurization 

and failure of a safety-related (solenoid

operated valve) appears not to have 

received the emphasis it deserves in the 

original design basis. Even though safety

related components that depend on the air 

system are designed to assume a fail-safe 

condition on loss of air, the converse 

condition, of air over-pressurization, may

not have been consistently considered. 

(4.4.1) 

15.2 Accident Sequence Pre

cursor Analysis 

As part of this study, an analysis was made 

of previous Accident Sequence Precursor (ASP) 

analyses. The results of that review are included 

in Appendix B. Several relatively significant 

events involving failures of AOVs and SOVs 

were identified. These events involved failures 

of the valves caused by design, installation, and 

maintenance activities, as well as failures in the 

air systems. Conclusions from the review of the 

ASP analyses include: 

" AOV-related precursors have been 

important contributors in terms of condi

tional core-damage probability. There 

were 26 AOV-related precursors, of the 

288 total precursors (i.e., events with 

conditional core damage probability 

(CCDP) contribution greater than or equal 

to lE-6) from 1984 through 1995. The 

1985 Turkey Point AFW event ranked 

highest among AOV precursors, with a 

CCDP contribution of about 9E-4. 

" Twelve AOV-related precursors had 

CCDP contributions greater than or equal 

to 1E-4 from 1984 to 1995. 

" Nineteen (12 operational, 5 testing, 

2 maintenance) of the 26 AOV-related 

precursors occurred while the plants were 

at power. 

"* Eleven of the 26 AOV-related precursors 

involved common-cause considerations. 

Four of those eleven events had CCDP 

contributions greater than 1E-4. 

"* A variety of safety-related and non

safety-related systems were impacted by 

the precursors.

71 NUREG/CR-6654
Risk Analysis Involving AOVs

15.3 Recent AOV risk Studies 

15.3.1 Risk Studies in Several of the 

Plants Visited 

Information on risk significance of air sys

tems, particularly contributions of failure of the 

air system as a percentage of core damage fre

quency (CDF), was gathered during the plant 

visits. Relative risk significance of particular 

AOVs and SOVs were also studied and infor

mation was offered at most of the plants visited 

because the licensees were engaged in PRA

related activities. These activities were prompted 

by their efforts to comply with the Maintenance 

Rule (10 CFR 50.65), and their ongoing efforts 

regarding probabilistic and risk-informed stud

ies, including Individual Plant Examinations 

(IPEs) in the form of Probabilistic Safety 

Assessments (PSAs) produced in response to 

NRC Generic Letter 88-20 and its supplements. 

The methods endorsed in NRC Regulatory 

Guide 1.160 and described in NUMARC 93-01 

rely heavily on PRA tools such as risk-reduction 

worth (RRW), CDF, and risk-achievement worth 

(RAW) to rank the importance of systems, and 

thereby, the valves of interest here. 

The engineers at Fermi 2 used PRA tech

niques and results from their risk analyses to 

classify AOVs in terms of risk significance. 

Lists of risk significant and "critical" AOVs 

were generated. Among the AOVs that ranked 

high in terms of Fussell-Vesely (FV) importance 

were the hardened vent isolation valves and 

those associated with the RHR heat exchangers. 

Thirty-three AOVs (22 safety-related and 

11 non-safety related that perform a risk 

significant function) and 370 control-rod-drive 

scram inlet and outlet valves were considered 

safety-significant. 

Similar techniques were used at Palisades to 

identify "important" AOVs. A list of 11 risk 

significant AOVs (out of 84 modeled as active 

in the PSA, 9 of which are not categorized as 

active in the plants' IST program) was provided 

during the site visit. 

According to the LaSalle Summary PRA, 

"(t)ransients with loss of instrument air, (TI 1),

are the largest initiating event category, contrib

uting 32% of the CDF. These transients are sig

nificant because venting containment cannot be 

performed without instrument air. Failure to vent 

results in the loss of the ADS function (and sub

sequent loss of the low pressure injection sys

tems) and eventual containment failure, causing 

potential loss of injection systems in the reactor 

building due to severe environments." 

The LaSalle Summary PRA indicated that 

"Loss of offsite power, LOSP, events are the 

second highest contributor to CDF. Single unit 

LOSP events contribute 6.5% of CDF and dual 

unit LOSP events contribute 22.9%. The core 

damage contribution of the SBO sequences 

(subset of LOSP) is 17.2%." A loss of offsite 

power to Unit 1 or to both units at LaSalle leads 

directly to a loss of instrument air, and the CDF 

resulting from such an event would be expected 

to be at least as significant as a loss of instru

ment air alone. The "Accident Sequence Event 

Descriptions" listed in the tables in the LaSalle 

Summary PRA include the event described as 

"LOSS OF INSTRUMENT AIR lE OR LOSP 

AT UNIT 1;" however, the Summary PRA does 

not provide a discussion of the relationship or 

dependencies between the two events. 

Risk studies were also conducted at TMI-1 to 

rank valves of all types in terms of relative risk 

significance. Several air-operated containment 

isolation valves ranked high in importance. It 

was also noted that loss-of-instrument-air was 

ranked sixth among the top 10 PRA core dam

age sequences, accounting for 5.3 percent of the 

total (small LOCAs contributed 18.8 percent). 

Tables of AOVs indicating risk significance 

and tables of events showing importance rank

ings were furnished by the engineers at Indian 

Point 3 during the site visit. By hand count, 

21 of 176 AOVs (with safety-function listed as 

"active" or "unknown") were considered by the 

engineers to have a "High PRA." Over 60 AOVs 

in the plant were classified as risk significant, 

based on a combination of evaluation criteria. 

The individual failures of MSIVs and EDG Flow 

Control Valves were found to be not particularly 

risk significant but the CCF of two or more of 

each type of valve was found to have

NUREG/CR-6654 72
Risk Analysis Involving AOVs

considerable risk significance. (RAW was com

puted by the licensee to be 1.614 for failure of 

an MSIV to close on demand. RAW was com

puted to be 50.73 for CCF of two or more 

MSIVs. RAW was not computed or not signifi

cant for failure of an individual EDG flow con

trol valve. RAW was computed to be 46.97 for 

CCF of both EDG flow control valves.) 

The Reliability and Risk Assessment Group 

(RRAG) from Florida Power and Light (FPL) 

undertook a preliminary AOV risk ranking 

analysis during and shortly after the visit for this 

AOV study. Their analysis was based on the 

44 AOVs that are modeled in the Level 1 PSA 

for Turkey Point 3 as well as some shared 

systems between Units 3 and 4 (e.g., AFW). Of 

the 44 AOVs, 17 met the Maintenance Rule risk 

significance criteria for further evaluation by an 

expert panel. Three AOVs of the 17 that met the 

Maintenance Rule risk significance criteria were 

reported to have a risk achievement worth 

(RAW) of about 2.4; however, they were con

sidered to be not risk significant by the expert 

panel because of mitigating factors such as iso

lating leaks or breaks using in-line manual 

(hand-operated) valves that were not credited in 

the PSA. According to FPL, one of the 17 AOVs 

had a RAW of about 1.3, and one had a RAW of 

1.2; the other 12 AOVs had a RAW of 1. 

The RRAG/FPL, incidental to discussions 

during the visit for this study, increased the 

assumed failure rate for AOVs by a factor of 10. 

The CDF increased from 5E-5 to 1.1E-4. They 

also described a study that they did for the 

common-cause failure sensitivity of two AOVs, 

856A and B, which had been replaced by MOVs 

in 1987 as part of Bulletin 86-03 compliance. 

See NUREG/CR-5295, "Closeout of IE Compli

ance Bulletin 86-03: Potential Failure of Multi

ple ECCS Pumps Due to Single Failure of Air

Operated Valve in Minimum Flow Recirculation 

Line." The sensitivity study, prior to modifying 

the plant design, indicated a risk achievement 

worth (RAW) of 21.3 for the common-cause 

failure of the two (originally air-operated) valves 

due to loss of IA.

Table 6 lists air-operated valves considered 

by the licensees in the plants visited to be risk 

significant. 

15.3.2 Generic Issue 158 Draft Study 

and Subsequent Sensitivity 

Analysis 

In 1996, a sensitivity analysis draft report, 

INEL-95/0550, entitled "GI-158: Performance 

of Safety Related Power Operated Valves Under 

Operating Conditions," was prepared for the 

NRC as part of their process for resolving 

Generic Issue (GI) 158, "Performance of Safety

Related Power-Operated Valves Under Design

Basis Conditions." That draft study included a 

sensitivity analysis and has since been desig

nated as NUREG/CR-6644 and published in 

September 1999. 

As was noted in the Executive Summary of 

NUREG/CR-6644, "(t)his study reveals that 

changes in the failure probabilities of some 

POVs will result in proportional changes in the 

system unreliability and indicates that POVs 

have an important role in system reliability." 

Another sensitivity analysis was prepared 

(see reference 57), as part of this current study 

of air-operated valves (AOVs) in nuclear power 

plants, to assess the impact on core damage fre

quency (CDF) resulting from the operating per

formance of AOVs. Additional sensitivity analy

ses, focused on AOVs and SOVs, were 

performed in an effort to assess core damage 

frequency (CDF) for failures of power-operated 

valves. Some of the insights and results of these 

analyses were used in the preparation of 

NUREG/CR-6644. 

Failure probabilities used in PRAs of 

44 plants were examined in NUREG/CR-6644. 

Bounding failure probabilities for AOVs and 

SOVs of three per hundred demands were identi

fied. Bounding failure probabilities for HOVs of 

five per hundred demands were also identified. 

These values were used in the CDF sensitivity 

analyses conducted as part of the study. It was

73 NUREG/CR-6654
Risk Analysis Involving AOVs

noted in NUREG/CR-6644 (Section 2.1 after 

Section 2.1.3) that "(s)ome plants with several 

years of operating experience are using very low 

generic probabilities in PRA calculations." 

Common-cause beta factors of 0.022, 0.062, and 

0.115 were calculated for AOVs, HOVs, and 

SOVs respectively. The SOV beta factor was 

calculated as 0.015 if "...the anomalous per

formance of ... one site is excluded from the cal

culation." 

Sensitivity analyses for a General Electric 

BWR, two Westinghouse 4-loop PWRs, a 

Westinghouse 3-loop PWR, a B&W PWR, and 

two Combustion Engineering PWRs (7 plants) 

were conducted as part of the GI-158 study. 

The analysts who conducted the GI-158 

study used the models and data that were avail

able to them and employed standard techniques 

for performing the risk studies. However, the 

failure scenarios and failure data may not have 

reflected higher risk significance of AOV and 

SOV failures and degraded conditions because: 

"* The risk models for the plants considered 

were not set up to account for common

cause AOV or SOV failure contributions 

resulting from air contamination (over 

long periods), gradual air pressure degra

dation, or over-pressurization. 

"* Failures of AOVs or SOVs in the BWR 

reactor protection system were not 

included in the BWR risk model. Scram 

pilot SOVs contribute about one third to 

the total unavailability of scram systems 

(NUREG/CR-5500, Vol. 3). 

"* A number of pertinent AOV and SOV

related events and conditions, including 

common-cause failure events and condi

tions were not included in LERs and other 

reports and thus, were not available for 

consideration in these (or other) risk 

studies. 

Core damage was the only "figure-of

merit" included in the studies. Large early 

release is a pertinent figure-of-merit for 

AOVs, but was not considered.

15.4 Common-Cause Failures 

NUREG/CR-4780, "Procedures for Treating 

Common Cause Failures in Safety and Reliabil

ity Studies" was reviewed during the preparation 

of this study of AOVs. That report is devoted to 

an exploration of common-cause failures and 

events in their relation to risk analyses, as well 

as the difficulties involved in arriving at consis

tent characterizations of common-cause failures 

and events. Section 1.4 of NUREG/CR-4780 

includes the following definition, which is con

sidered useful for the discussion here. "In a PRA 

model, a common cause event is defined as the 

failure or unavailable state of more than one 

component at the same time and due to the same 

shared cause. Common cause events require the 

existence of some cause-effect relationship that 

links the failures of a set of components to a sin

gle shared root cause." 

Recently, NUREG/CR-6268, "Common

Cause Failure Database and Analysis System:..." 

(Four Volumes) offered, in Volume 3, 

Section 1.3, the following criteria to define a 

common-cause failure event: 

1. "Two or more components fail or are 

degraded at the same plant. Failures are 

discovered during equipment challenges 

to operate, surveillance testing, or design 

deficiencies that are detected prior to 

operating the equipment. In the case of a 

failure resulting from a design deficiency, 

a potential failure is considered to have 

the same severity as a failure that results 

from a challenge to the equipment, pro

vided the design deficiency would have 

caused a component to fail on demand. 

For example, a wiring discrepancy that 

would prevent a pump start is considered 

to be a complete failure, even if no start 

was attempted. 

2. Component failures occur within a 

selected period of time. 

3. The component failures result from a sin

gle shared cause and are linked by a cou

pling mechanism such that other

NUREG/CR-6654 74
Risk Analysis Involving AOVs

components in the group are susceptible 

to the same cause and failure mode. 

4. The equipment failures are not caused by 

the failure of equipment outside the 

established component boundary." 

Section 2.4 of NUREG/CR-5472, "A Risk

Based Review of Instrument Air Systems at 

Nuclear Power Plants," included a review of 

instrument air events based on the events and 

findings in NUREG-1275, Volume 2. A "Sum

mary of Instrument Air Common Cause Events" 

prior to 1986 was included in Table 1-5 of 

NUREG/CR-5472. Some of the events described 

in NUREG-1275, Volume 2, as common-cause 

failures were characterized in NUREG/CR-5472 

as not really fitting the common-cause concept 

because the time interval between events or the 

discovery of failures was considered to be too 

long and corrective or defensive measures could 

have been taken to mitigate the effects of the 

failures or events. However, many of these 

events meet the definition of common-cause 

failure described in NUREG/CR-6268. 

The NRC/AEOD published a report, 

AEOD/E92-02, "Insights From Common-Mode 

Failure Events," dated June 1992 and Supple

ment 1, dated February 1993. These reports were 

referred to in NRC Information Notice 93-35, 

"Insights From Common-Cause Failure Events," 

dated May 12, 1993. The significance of com

mon-cause failures in the plants was reviewed in 

the reports and the reviews indicated that many 

common-cause failures are associated with 

errors in design and equipment qualification. 

The reports emphasized that common-cause 

failures defeat the single-failure criteria which is 

basic to the satisfactory safety performance of 

nuclear plants. The reports also referred to sev

eral events that occurred over extended periods 

of time. 

Common-cause failures of AOVs are consid

ered to include failures that occurred or could 

occur from the same event or root cause, such as 

loss of instrument air, contamination of instru

ment air, SOVs that failed or were damaged 

from the same source of particulate contamina

tion, SOVs improperly installed in a harsh

environment, SOVs designed with a common 

defect, or a group of the same AOVs installed 

with inadequate thrust capability to meet design 

basis conditions. The point is that the common

cause failure mechanism did or could cause the 

same failure of two or more AOVs at the same 

time or over time, and if not detected and cor

rected, compromised or could compromise 

safety because these multiple failures occurred 

or could occur at some, not necessarily simulta

neous, minimum demand conditions. 

Functional failures in one AOV that remain 

undetected until another of the same type AOV 

fails in the same or a different system from the 

same cause, and then the failure in the first AOV 

is subsequently discovered, are of particular 

concern because the vulnerabilities may exist 

over long periods. 

Common-cause failures should be taken into 

consideration when ranking the risk-based 

importance of AOVs. Individual AOVs may be 

found to be of low risk significance but the 

common-cause failure (CCF) of two or more 

AOVs performing the same function may have 

considerable risk significance. For example, at 

Indian Point 3, the individual failures of MSIVs 

and EDG Flow Control Valves were found to be 

not particularly risk significant but the CCF of 

two or more of each type of valve was found to 

have considerable risk significance. (RAW was 

computed by the licensee to be 1.614 for failure 

of an MSIV to close on demand. RAW was 

computed to be 50.73 for CCF of two or more 

MSIVs. RAW was not computed or not signifi

cant for failure of an individual EDG flow con

trol valve. RAW was computed to be 46.97 for 

CCF of both EDG flow control valves.) 

15.5 AOV Failure Rates 

NUREG/CR-1363, "Data Summaries of 

LERs of Valves at U.S. Commercial Nuclear 

Power Plants, January 1, 1975 - December 31, 

1980," reported an estimate of two failures per 

thousand demands for AOVs. This number 

appears to be a number commonly used in risk 

analyses.

75 NUREG/CR-6654
Risk Analysis Involving AOVs

NUREG/CR-1363 also reported an estimate 

of six failures per thousand demands for motor

operated valves, a commonly used figure at that 

time. NUREG/CR-5140, "Value Impact Analy

sis For Extension of NRC Bulletin 85-03 To 

Cover All Safety-Related MOVs," reported an 

estimated failure rate of over eight per hundred 

demands, based on specific licensee reports of 

MOV failures for loaded demands, prior to the 

publication of Generic Letter 89-10. The dis

crepancy between the estimates in the two 

reports may have been in the "demand" denomi

nator. No-load demands for stroke-timing tests 

and tests after repairs were included as demands 

in NUREG/CR-1363 data, along with the loaded 

demands in which the motor-operated valves 

were subjected to pressure or flow. 

The failure rate for AOVs is expected to be 

higher than two failures per thousand loaded 

demands, although it may not be as high as it 

was found to be for MOVs because of the expe

rience gained by the licensees over the past 10 

years. However, much of the experience gained 

by licensees from motor-operated valves has yet 

to be incorporated into the design and setup of 

AOVs. Also, as was the case with MOVs, many 

important AOVs have not been tested at design 

basis (including accident) conditions. Perform

ance predictions for AOVs that have been 

extrapolated from no-load and low-load tests or 

service conditions might not be valid at design 

basis conditions. 

15.6 Observations Pertinent to 

AOVs from Examination 

of Risk-Related Analyses 

and Probabilistic Safety 

Assessments (PSAs) for 

Motor-Operated Valves 

(MOVs) 

The NRC and the nuclear power industry 

have used risk analysis tools in several studies 

involving the implementation of Generic Letters 

89-10 and 96-05 to improve the reliability and 

performance of safety-related MOVs. Many of 

the methods and observations in these studies of 

MOVs provided insights that are considered

pertinent to the present study of AOVs and 

licensees' prospective programs for the 

improvement of AOV performance. The studies 

of MOVs that were reviewed included: 

"* NUREG/CR-5140 (BNL-NUREG-52145), 

"Value-Impact Analysis For Extension Of 

NRC Bulletin 85-03 To Cover All Safety

Related MOVs," July 1988. 

"* BNL TECHNICAL REPORT A-3869

T52-10-91, "Safety Significance Of Inad

vertent Operation Of Motor Operated 

Valves In Safety-Related Piping Systems 

In Boiling Water Reactors," October 

1991. 

"* BNL TECHNICAL LETTER REPORT 

E-2071-Ti -12-93, "Safety Significance 

Of Inadvertent Operation Of Motor Oper

ated Valves In Safety-Related Piping 

Systems In Pressurized Water Reactors," 

Revision 1, March 1995. 

" BWR Owners Group Report 

NEDC-32264A, "Application Of 

Probabilistic Safety Assessment To 

Generic Letter 89-10 Implementation," 

October 1, 1996. 

"* Westinghouse Engineering Report V-EC

1658-A, "Risk Ranking Approach For 

Motor-Operated Valves In Response To 

Generic Letter 96-05," Revision 2, July 

1998. 

BWR Owners Group Report NEDC-32264A 

and Westinghouse Owners Group Report V-EC

1658-A were referenced in the Joint Owners 

Group AOV Program (references 6.10 and 6.11 

respectively), Revision 0, dated March 9, 1999. 

These reports were endorsed to provide accept

able methods, among others (see Section 4.1.3.2 

of the JOG AOV Program) for ranking safety 

significance and conducting an expert panel 

review of AOVs. 

The value/impact analysis in NUREG/CR-5140 

included a risk-based analysis of the effects of 

the higher MOV failure estimates reported by 

licensees in response to NRC Bulletin 85-03 in

NUREG/CR-6654 76
Risk Analysis Involving AOVs

relation to the previously estimated failure rates. 

We have not been able to accurately quantify 

AOV failure rates for our current study of AOVs 

in a manner comparable to those documented for 

MOVs in NUREG/CR-5140. However, based on 

our previous observations, we believe that AOV 

failure rates for demands at pressure or flow 

conditions are substantially higher that the esti

mate of approximately two per thousand 

demands commonly used in risk studies. If a 

sufficiently reliable failure rate for AOVs can be 

documented, the methods used in NUREG/CR

5140 would be appropriate for a value/impact 

study of AOVs. 

BNL reports A-3869-T52-10-91 and E-2071

T1-12-93 were prepared in response to BWR 

Owners Group and Westinghouse Owners 

Group questions regarding the issue of the risk 

significance of MOV mispositioning. The 

NRC's concern was related to the ability of 

"position changeable" MOVs in the event of 

their inadvertent operation from the control 

room. In the Davis-Besse event in 1985 (that 

subsequently led to the publication of NRC 

Bulletin 85-03 and Generic Letter 89-10), sev

eral MOVs were inadvertently closed from the 

control room and could not be reopened because 

their torque switches were improperly set. The 

reports indicated that plant risk increased for 

BWRs and PWRs by about one order of magni

tude for mispositioning of MOVs and their sub

sequent failure. 

Failure of AOVs in previously unanticipated 

positions has been documented a number of 

times. Drift of AOVs is also possible as a result 

of partial and/or gradual loss of pneumatic pres

sure. Such losses have occurred by failures in 

SOVs or other pneumatically controlled devices. 

Although the hardware and failure mechanisms 

may be somewhat different for the MOV and 

AOV operators, the concern is the same, i.e., the 

ability of the valve to perform the designed 

function. In addition, the failure mechanisms for 

AOVs include common-cause failure mecha

nisms based on failure of the air supply or con

tamination therefrom, and other failure mecha

nisms such as material-related failures in control 

solenoids. The methods used in these two BNL 

technical reports provide useful guidance for

work to estimate risks associated with misposi

tioning of AOVs. 

15.6.1 BWR Owners Group Report 

NEDC-32264-A 

BWR Owners Group Report NEDC-32264-A 

described a method for categorizing MOVs for 

initial testing and subsequent periodic verifica

tion to meet the recommendations in GL 89-10. 

The methods discussed in NEDC-32264-A 

described a risk-based categorization and rank

ing process for MOVs that may be of interest in 

the study of AOVs. This BWROG topical report 

described a seven step process for ranking 

MOVs in order of risk significance, as follows: 

REVIEW PLANT PSA TO DETER

MINE HOW WELL THE PSA REPRE

SENTS THE MOVS PERTINENT TO 

THE GL 89-10 PROGRAM. 

The PSA for a particular plant is to be 

reviewed to develop an understanding of 

MOV performance. Level 1 and 2 PSAs 

are to be reviewed in order to consider 

MOVs that affect core damage and con

tainment integrity. Failure of MOVs to 

change position is to be emphasized. The 

PSA may "mask" the importance of some 

MOVs depending on how it is con

structed. High-energy line break scenarios 

are to be considered. Finally, the role of 

MOVs in initiating events is to be 

investigated.

REVIEW GL 89-10 MOVS 

INCLUDED IN THE PSA.

NOT

The reasons why particular MOVs are not 

modeled in the PSA are to be considered 

and documented. 

REVIEW IMPORTANCE MEASURES 

(FIGURES-OF-MERIT) USED IN THE 

PSA. 

Completion of this task results in a listing 

of the MOVs and their associated numeri

cal importance measures. Various ranking 

processes and their attributes are 

described in the report.

77 NUREG/CR-6654
Risk Analysis Involving AOVs

"* QUANTIFY THE IMPORTANCE OF 

MOV IMPLICITLY MODELED IN THE 

PSA. 

Because PSAs generally treat some 

potential failures implicitly, an expanded 

review of MOV importance is required. 

Implicit modeling refers to actions where 

the successful function of an MOV is 

required but may not be explicitly stated, 

for example "operator aligns containment 

venting." The role of MOVs in such 

events is to be considered. 

"* PERFORM SENSITIVITY ANALYSES 

TO ENSURE THAT MOVS ARE 

EVALUATED PROPERLY. 

Three issues are to be addressed: 

- Truncation (base events dropped) 

in the generation of cutsets from 

fault trees due to low failure 

probabilities; 

- Truncation due to calculation cut

off values (set to minimize program 

run-times); and 

- Common-cause failures of MOVs 

due to potential widespread per

formance deficiencies. These issues 

are discussed in the report and are 

to be considered in the analyses. 

Testing and verification of per

formance are suggested, along with 

expert-panel opinions, as are meth

ods to minimize the impact of these 

issues, particularly common-cause 

failures. 

MOVs determined to be important due to 

intersystem common-cause failure poten

tial would, at least initially, be subjected 

to increased verification testing. Methods 

for considering sensitivity analyses 

involving intersystem common-cause 

failures, which are not generally modeled 

in PSAs, are discussed.

" CATEGORIZE MOVS AND APPLY 

TESTING CRITERIA. 

A tabular form is offered for MOV risk 

prioritization results. Various risk ranking 

methods are discussed and review of the 

results by an expert panel is recom

mended. 

"* MOVS ARE CATEGORIZED AS HIGH, 

MEDIUM, OR LOW, IN ACCOR

DANCE WITH THEIR IMPORTANCE 

TO CORE DAMAGE OR LARGE 

RELEASE OF CONTAMINATION. 

Table 2 of NEDC-32264-A provides the pro

posed ranking criteria for MOVs as follows:

TABLE 2 (OF NEDC-32264-A)RANKING CRITERIA 

FOR MOVSFOR GENERIC LETTER 89-10 

RANK CRITERIA(a)(b)(c) NOTES 

High >1% CDF Additional MOVs can be

GL 89-10 MOVs 

Medium >1%CDF>0.1%GL 

89-10 MOVs 

Low Remaining GL 89-10 

MOVs 

<0.1% CDF

added based on judge

ment, sensitivity 

analyses. 

Adequate justification for 

valves in this category 

should exist.

(2) These importance criteria establish the baseline for valve inclu

sion. However, as noted in Task 4, qualitative assessments further 

evaluate the inclusion of other MOVs. 

(b) Similar criteria for Level 2/RRF should be utilized. 

c See ADDENDUM 1 for correlation of % CDF and F-V. 

Failure rates for the MOVs were modeled in 

NEDC-32264-A primarily as 3 per 1000 demands 

and 8.7 per 100 demands for the BWRs studied. 

A failure rate of around 3 per 1000 demands 

had been used for MOVs in previous risk studies 

such as WASH-1400 (as reported in 

NUREG/CR-1363). NUREG/CR-5140 docu

mented a failure rate of about 8.7 per 

100 demands based on the licensee responses to 

NRC Bulletin 85.-03. 

Section 2.3.1 and 2.3.2 of NEDC-32264-A 

included discussions of "multi-component"

NUREG/CR-6654

78
Risk Analysis Involving AOVs

issues where an MOV might be more important 

because of the simultaneous failure of another 

MOV, along with related discussions of com

mon-cause failure modeling. It appears that con

ditions related to common-cause failure of mul

tiple MOVs caused by common adjustment to 

design defects (e.g., torque switch setting defi

ciencies or insufficient operator margin) are 

minimized in the discussions. The common

cause failure mechanisms for AOVs and MOVs 

may not be directly comparable with respect to 

the deficiencies in the powering mechanisms 

and valve operators, and thus the assumptions 

used to estimate risks due to common-cause 

MOV failures might not be appropriate if 

extrapolated to discussions of common-cause 

failures of AOVs. 

A discussion of the results of application of 

the above outlined methods to five BWR plants 

follows in the NEDC-32264-A report. The MOV 

prioritization results are summarized in Table 4 

of NEDC-32264-A. Details of the studies for 

each of the five BWRs are included in Appendi

ces to the NEDC-32264-A report. The ranking 

process and use of various risk importance 

measures were studied further in a section 

labeled Addendum 1 which was included at the 

end of the report. 

Definitions of importance measures were 

included in Section 3.1 of the Addendum in the 

NEDC report. These are considered to be perti

nent to discussions of risk and are therefore 

reproduced below for information. 

SYMBOLS: 

T = Base core damage frequency for all 

basic events 

U = Failure probability (or unavailability) 

of individual basic event 

T(O) = CDF with basic event assumed to 

never occur (i.e., probability set 

equal to 0) 

T(1) = CDF with basic event assumed to 

occur (i.e., probability set equal to 1)

From the above basic PSA inputs the following 

importance measures can be calculated for each 

individual basic event: 

a. Risk increase where basic event is 

assumed to occur (i.e., basic event prob

ability set equal to 1). 

Risk Increase = T(l) - T 

b. Risk reduction where basic event is 

assumed never to occur (i.e., basic event 

probability set equal to 0). 

Risk Decrease =b T - T(O) 

c. Fussell-Vesely (FV) importance is the 

fraction of the CDF which involves the 

basic event divided by the base CDF. In 

some PSAs this represents the sum of the 

CDF for the minimum cutsets containing 

the basic event divided by base CDF. A 

minimum cutset is defined as the smallest 

combination of failures (or basic events) 

which, if they all occur, will cause the top 

event (core damage) to occur. 

FV = {bT - T(0) }/T

d. Criticality (CRMr) importance is 

follows:

as

CRIT = {I[T(lb)-T(O)]xU}/T=FV 

e. Risk Reduction Worth (RRW) is the base 

CDF divided by CDF with U = 0. 

RRW = T/T(0) 

f. Birnbaum (BIRN) importance is as 

follows: 

BIRN = T(1)-T(0)=CRITx(TIU) 

g. Risk Achievement Worth (RAW) is the 

CDF with U = 1 divided by base CDF. 

RAW = T(1)/T 

h. Cumulative % Risk Contribution is cal

culated by first ranking (sorting) the basic

79 NUREG/CR-6654
Risk Analysis Involving AOVs

events by decreasing "Risk Decrease" or 

decreasing "F-V." The % risk reduction is 

the risk decrease divided by the sum of 

the risk decreases of all basic events. The 

cumulative % risk reduction is then the 

sum of the individual % risk reduction in 

order of their size. The number of SSCs 

included depends upon the total cumula

tive % of interest. 

% Risk Reduction = [T-Tj(0)] 

Y_[T -Tj (0)] xl00 

j=1 

n 

Cumulative % Risk Contribution = X i=1 

(% Risk Reduction) 

where "n" is the number of SSCs required to 

obtain the cumulative % risk of interest. 

A NRC Preliminary Safety Evaluation 

Report (SER) on the first revision of 

NEDC-32264-A, comments from the BWR 

Owners Group, and a final SER were included at 

the end of the report. The NRC staff determined 

that the methodology described in the report was 

acceptable, subject to a number of comments in 

the SER. The use of risk-related rankings to 

establish test intervals was accepted and it was 

noted that the methodology was considered to be 

a pilot in the NRC's PRA Implementation Plan. 

The nuclear power industry proposed to apply 

risk analysis techniques in the inclusion of 

MOVs for the licensees' Generic Letter 89-10 

programs. They planned to exclude certain "low 

risk" safety-related MOVs and perhaps include 

certain "high risk" but non-safety-related 

MOVs. It appeared that their proposed commit

ments toward surveillance of high risk but non

safety-related MOVs were ambiguous. How

ever, the NRC's SER indicated that licensee 

commitments to ASME Code testing and to the 

recommendations in Generic Letter 89-10 to 

include all safety-related MOVs were to remain 

in place.

15.6.2 Westinghouse Engineering 

Report V-EC-1658-A 

The purpose of the Westinghouse Report is 

to provide a guideline on how Westinghouse 

NSSS Owners can rank MOVs according to 

their risk importance. The report describes a 

suggested process for MOV risk ranking and is 

based on previous risk ranking programs 

including NEDC-32264-A, described above. 

This is part of a Joint Owners Group (JOG) 

effort to ensure MOV operability. The Westing

house guideline is oriented toward implementa

tion of NRC Generic Letter 96-05 regarding 

periodic verification of MOV operability at 

design basis conditions. 

The Westinghouse report includes a separate 

(from NEDC-32264-A) risk ranking guideline 

for MOVs based on a six step process, as 

follows: 

"* IDENTIFY MOVS TO BE CONSID

ERED. 

A separate JOG report (MPR-1807) is 

used to identify MOVs within the scope 

of the program. 

"* CALCULATE MOV AT-POWER RISK 

IMPORTANCES. 

Plant PSAs are used to determine at

power risk importances for MOVs within 

the scope of the program. The failure 

modes considered were failure to change 

position on demand. Importance measures 

similar to those described above in 

NEDC-32264-A are used in the risk 

ranking process. In addition, the Westing

house Report notes (Section 3.2.1) the 

relationship between FV and RRW as 

RRW = 1/(1 - FV) and quotes the impor

tance measure used by the BWROG for 

risk ranking MOVs for implementation of 

GL 89-10 as "Importance = (Sum of all 

accident sequence frequencies related to a 

specific MOV) / (Total CDF)."

NUREG/CR-6654

80
Risk Analysis Involving AOVs

A three level approach to risk ranking MOVs 

is recommended in the Westinghouse Report 

based on the following criteria: 

Risk Category Ranking Criteria 

High FV > 0.01 or RAW >10 

Medium 0.01 > FV > 0.001 and RAW < 10 or 

10 > RAW > 2 and FV < 0.01 

Low FV < 0.001 and RAW < 2 

0 ASSESS PSA COMPLETENESS ISSUES. 

Several issues are discussed in Section 3.3 

of the Westinghouse report that 

are directly pertinent to the study of 

AOVs as well as MOVs. The issues of the 

completeness of failure data and the 

accuracy of failure rates are discussed in 

the Westinghouse report. The report 

indicates that failure rates ranging from 4 

in 10000 demands to 1 in 100 demands to 

open or close were used. It was noted that 

the NRC took issue with these rates in 

their SER (enclosed as part of the 

report) and noted the failure rate of 

8.7 per 100demands documented in 

NUREG/CR-5140. 

Accident sequence and cutset truncation 

limits that account for at least 90% of the 

PSA CDF are recommended. Also, com

mon-cause evaluations for at-power risk 

analyses are to be considered in the nor

mal manner for PSA analyses. 

The Westinghouse Report endorses PSA 

Level II analyses for evaluating the 

importance of MOV failures on release of 

contamination due to containment failures 

and indicates that RAW and FV impor

tances should be evaluated. 

The Westinghouse Report notes that 

CDFs for some plant initiating events are 

based on plant-specific design and com

ponent reliability. The initiators listed are 

loss of normal service water, loss of com

ponent cooling, loss of chemical and 

volume control system, and total or partial

loss of main feedwater. These issues, 

along with loss of instrument air or air 

system or component overpressurization, 

as noted previously, are pertinent to 

evaluation of the risk impact of AOVs. 

Similarly, it is noted in the Westinghouse 

Report that not all MOV failures are 

modeled as basic events in the PSA model 

and that they may be subsumed or com

bined with other events. Engineers should 

be aware of these factors and consider 

them in the risk analyses. This is also the 

case for AOVs. 

EVALUATE OTHER CONSIDERA

TIONS. 

Section 3.4 of the Westinghouse Report 

indicates that in addition to the PSA 

importance measures, the following con

siderations are pertinent and should be 

considered by a designated panel of 

experts: 

- Components not modeled in the 

MOV program or PSA model; 

- Shutdown risk; 

- External events; 

- Component operating history; 

- Assessment of interchanging func

tions of MOVs; 

- Plant-specific configurations 

(equipment out of service for 

maintenance and trains or equip

ment out of service for long periods 

of time, as referred to in 10 CFR 

50.65, The Maintenance Rule); 

- MOV importance in the design 

basis analysis performed to satisfy 

the FSAR and Technical Specifica

tion requirements; and 

- Failures of MOVs that could lead 

to other accident scenarios.

81 NUREG/CR-6654
Risk Analysis Involving AOVs

"* DEVELOP COMPONENT RANKING 

WORKSHEETS. 

Worksheets are included in the Westing

house Report that are to contain the perti

nent information gathered during the 

ranking and risk evaluations. These are 

also used to assist the expert panel in 

making decisions about ranking specific 

MOVs. 

"* CONDUCT EXPERT PANEL SESSION 

FOR MOV RANKING. 

The expert panel is to evaluate the 

importance of each MOV and categorize 

them in priority ranking. 

A NRC SER was published in the final 

Westinghouse report and use of the report was 

determined to be acceptable subject to several 

conditions and limitations summarized at the 

end of the SER. Among the conditions were: 

"* Licensees must document their reasons 

for categorizing MOVs as having low 

safety significance. 

"* Accuracy of failure rate data and operat

ing conditions under which the data were 

collected are to be verified.

"* Diagnostic test intervals of up to 10 years 

are approved provided the potential for 

common-cause failures caused by mainte

nance or test activities are considered by 

the expert panel. 

"* Both risk and deterministic criteria for test 

frequencies to ensure design basis capa

bility are satisfied. 

Intersystem common-cause failure rates were 

not considered to be a pertinent issue for MOVs 

in the Westinghouse Report (as well as in the 

BWROG report described previously) because 

the implemented recommendations of GL 89-10 

and GL 96-05 precluded such occurrences. Fur

ther, implementation of the recommendations of 

GL 89-10 and GL 96-05 was relied upon within 

the industry to ensure that MOV failure rates 

were actually around to 3 in 1000 demands so 

that the assumptions in the various risk analyses 

(PSAs, etc.) could be considered accurate. It is 

noted that no such program exists at the current 

time for AOVs; therefore, the issue of accuracy 

of failure rate estimates for AOVs is significant 

in assessing the conservativism of risk analyses.

NUREG/CR-6654

82
16. CONCLUSIONS

16.1 AOV Population 

There are large populations of safety-related 

and important non-safety-related AOVs 

including risk significant AOVs, in all of the 

plants visited. AOVs generally number in the 

hundreds and SOVs number in the thousands in 

each plant. AOVs may be found in a large 

number of systems including engineered safety 

feature systems at many plants. There are large 

varieties of AOV designs and manufacturers. 

Combined choices of valve, operator, and trim 

configurations are almost unlimited. One 

licensee has installed AOVs from 38 different 

manufacturers. Licensees may not be fully aware 

of the presence of SOVs that are internal to 

equipment. The population of AOVs and the 

variety of designs present a serious challenge to 

the operators to ensure that all of their important 

AOVs have been properly selected for the 

intended application, designed, installed, set up 

properly, and thereafter maintained in operating 

condition. 

16.2 AOV Design Basis 

Demands and Margins 

One of the first tasks planned for licensees to 

perform in the proposed model AOV programs 

(Fermi 2 and Palisades) was to review and 

confirm the design bases for AOVs. 

Environmental conditions as well as load 

demands are to be reviewed. This involves a 

considerable amount of effort on the part of 

licensees because the information may be 

scattered, fragmented, or missing. Review or 

establishment of design basis demands was also 

one of the first tasks performed for motor

operated valves under the guidance in Generic 

Letter 89-10. Licensees may find, as was the case 

with motor-operated valves, that the design bases 

for AOVs include incomplete information and/or 

nonconservative assumptions (refer to Section 

12.3). Design basis demands need to be reviewed, 

or established, as a first step in order to compare 

them with the margins available in AOVs.

Licensees may not know that the design basis 

loads or environmental conditions can be met 

with acceptable margins for the important AOVs 

in their plants. Licensees that were visited are 

relying on architect/engineers or manufacturers' 

information and calculations or, more recently, 

are conducting design reviews to determine the 

design basis margins for their AOVs. Several 

manufacturers' AOV calculations or valve 

descriptions were found that included mistakes 

or inaccurate information. Some licensees have 

conducted analyses and/or tests to verify or 

determine the design bases and margins for 

selected AOVs. 

Licensees are at various stages regarding 

their determination or confirmation of the design 

basis demands on safety-related and important 

non-safety-related AOVs. Palisades and Fermi 

are lead plants in an EPRI-sponsored analysis 

project to determine design basis demands and 

thereafter calculate the ability of AOVs to meet 

those demands. The effort is restricted to 

Category 1 and 2 AOVs as described in Section 

12.3. The categorization of AOVs, in terms of 

safety significance (and thus, subsequent 

attention by the licensees) varies from plant to 

plant. There is no NRC requirement for a 

standardized categorization. 

Plant licensing documents imply that 

licensees have confirmed that positive margins 

are available for safety-related AOVs to meet 

their design basis demands. The plant licensing 

documents also imply that failures of AOVs in 

important non-safety-related systems will not 

interfere with safety-related functions 

(shutdown, remain shutdown, and mitigate 

offsite dose). A similar situation existed 

regarding the operability of motor-operated 

valves before Generic Letter 89-10 was issued. 

Licensees discovered a number of interesting 

details about their design bases and the abilities 

of motor-operated valves to meet their 

commitments. Similar efforts on the part of 

licensees regarding AOVs can be expected to 

have similar results.

83 NUREG/CR-6654
Conclusions

16.3 AOV Failures 

A number of examples were found where 

AOVs, both safety-related and non-safety

related, did not have sufficient designed margin 

to meet their design basis or operational 

demands. Problems associated with the AOVs 

and their operators included improperly defined 

diaphragm areas (LaSalle), inadequate 

diaphragm materials (Dresden, Quad Cities, 

Indian Point, et. al.), improper estimates of 

packing loads, and underestimated valve factors 

(Palo Verde). The issue of the ability of safety

related and important non-safety-related AOVs 

to perform their functions is considered to be 

safety significant. 

In addition to the operability problems 

caused by insufficient margin or by the air 

systems, AOVs are subject to all of the 

environmental conditions, maintenance 

problems, and aging mechanisms of other valves 

in nuclear power plants. NRC generic 

communications are included in Appendix A 

that describe problems encountered with 

AOVs/SOVs or air systems in the nuclear power 

industry. The events and failures included- in 

Tables 2, 3, 4, and 7 in this study represent only 

a sample of the total number of events and 

failures involving AOVs. 

At almost all plants visited, a number of 

pertinent AOV and SOV-related events and 

conditions, including common-cause failure 

events and conditions, were not included in 

Licensee Event Reports or reports to the industry 

databases that the authors had access to. 

Licensees determined that they did not have to 

report these events or conditions to the NRC if a 

safety function of safety-related equipment was 

not compromised. 

16.4 Air Systems 

High quality air is critical to ensure the 

operability of AOVs. A reliable supply of clean, 

dry, oil-free air at specified pressure is a 

necessary, but not a sufficient condition for 

operability of AOVs. Failures in air systems that

serve AOVs are a source of common-cause 

failures in AOVs and this consideration is safety 

significant. 

Every plant visited had experienced major 

problems with air quality and had taken actions, 

some quite extensive, and mostly in response to 

Generic Letter 88-14, to address their air quality 

problems. Three of the seven licensees visited 

do not definitively know, on a day-to-day basis, 

if the quality of air (or nitrogen) in their plants 

has degraded with regard to moisture 

contamination. LaSalle, TMI, Indian Point 3, 

and Turkey Point 3 and 4 are equipped with 

continuous dew point monitoring and alarming 

(Turkey Point does not have alarms in the 

control room) of the instrument air systems. The 

other licensees check moisture content in their 

air systems by observation at varying periods 

ranging from taking readings at each shift (on 

rounds) to testing semi-annually. Note that these 

plants also have other air systems containing 

AOVs of interest that are not equipped with 

devices to measure dew point. 

AOVs, positioners, and regulators can be, 

and many times have been, directly 

compromised by moisture in the air that is used 

to operate the valve. These components can also 

be adversely affected by particulate 

contamination generated by moisture-induced 

corrosion of air piping and valve parts, or can be 

damaged by hydrocarbon attack on some 

elastomers. AOV positioners can malfunction 

because of variations of air pressure in the air 

system. 

High air pressure can damage some valves, 

particularly solenoids, and can do so 

simultaneously, depending on the circumstances. 

Low air pressure can directly affect the 

performance of an AOV (or several AOVs) or 

cause it (them) to move to unexpected and/or 

unanalyzed positions. Particulate contamination 

from external sources is normally removed by 

filtration. Corrosion products can be released 

into the air system downstream of filters, as the 

result of shock, vibration, or seismic events, 

depending on the material and condition of the 

piping.

NUREG/CR-6654 84
Conclusions

Over a period of time, contaminated air from 

moisture and/or particles may be a more serious 

problem than a sudden failure of the air system 

to provide air (the usually assumed failure in 

risk assessments and other system failure 

analyses) in that the contaminants provide a 

mechanism, suddenly or over time, to cause 

multiple AOV failures or degradation leading to 

failure. This situation calls into question the 

argument that AOVs connected to high pressure 

air systems, accumulators, or instrument air 

systems will "fail-safe," i.e., move to the safe 

positions that are assumed from a loss of air. Air 

quality must be verified (continuously or very 

frequently) to be satisfactory in order to 

substantiate the "fail-safe" assumption. The 

provisions of Generic Letter 88-14 to ensure air 

quality over the long term are somewhat vague 

on this issue and some events indicated 

examples of deteriorated air quality over time 

(refer to the discussion in Section 10.1 of the 

provisions of GL 88-14 and operating 

experience since its publication). 

16.5 Accumulators 

The quality of air or nitrogen supplied by 

accumulators serving individual AOVs is 

generally not monitored or verified on a regular 

basis to ensure operational reliability. Licensees 

may get verification and certification of nitrogen 

bottles from outside vendors. Many accumulators 

do not have drains and/or convenient means to 

check them to ensure that they are clean and dry. 

Further, if an accumulator has water in it, the lost 

volume of air or nitrogen could be significant if 

repeated strokes are required by the design basis 

for a particular AOV. There is also the potential 

for contamination of the air or direct interference 

with the performance of the operator from 

moisture in the air. Since accumulators 

inherently function as drain traps in the air 

system, the potential for having moisture and/or 

particulate contamination in them is likely, 

unless the air system is "extremely dry" at all 

times and the accumulators are monitored in 

some way.

Accumulator check valve failures can cause 

partial loss of air pressure. A partial loss of air 

pressure could cause the valve to fail in an 

intermediate position or not fail safe. Also, the 

loss of air may not be readily apparent and the 

AOV could then fail when operation is 

attempted. 

Safety-related AOVs may not be supplied 

directly by non-safety-related air systems but the 

air supplied to safety-related accumulators is 

usually, but not always, supplied by a non

safety-related instrument air system. The 

potential for moisture or particulate 

contamination, partial loss of pressure due to 

check valve malfunction, or loss of accumulator 

volume through water intrusion indicates a 

direct safety concern to ensure the operational 

reliability of both the air system and the 

accumulators that serve safety-related and 

important non-safety-related equipment. 

16.6 Solenoid-Operated 

Valves 

SOVs are a source of common-cause failures 

in both safety-related and important non-safety

related AOVs because of their numbers, their 

dependence on the air system, and their similar 

characteristics and common-cause failure 

mechanisms (see Section 11). The SOV failures 

and failure mechanisms described in 

NUREG-1275, Volume 6, in 1991 are currently 

commonplace in the nuclear industry. The safety 

significance of particular SOV failures depends 

on the application, however the number of 

safety-related and important non-safety-related 

applications is large enough that attention should 

be focused on these components. 

An NRC study, NUREG/CR-5500, 

Volume 3, "General Electric Reactor Protection 

System Unavailability, 1984-1995 (draft 2)," 

indicated that about one third of BWR RPS 

unavailability was due to common-cause failures 

of the hydraulic control unit and backup scram 

SOVs in that system.

85 NUREG/CR-6654
Conclusions

The Maintenance Rule is driving efforts by 

licensees to implement many of the independent 

recommendations regarding SOVs made in 1991 

in Section 9 of NUREG-1275, Volume 6. Those 

recommendations were: 

Take corrective actions to address the root 

causes of SOV failures based on risk 

significance and determined from a plant

specific prioritization scheme. 

Review SOV design specifications, 

calculations, and operating conditions 

including temperature and pressure 

limitations. 

Conduct reviews to identify SOVs that 

may have been overlooked and to verify 

the orientation of SOVs. 

Replace or refurbish SOVs on a timely 

basis. 

Ensure that the air systems that serve 

SOVs are a reliable source of clean, dry, 

oil-free air at proper pressure. 

16.7 AOV Program Plans 

All of the seven plants visited had AOV 

action plans in varying stages of completion at 

the time of the visits. One plant was focusing on 

maintenance practices, assuming that the design 

basis operability, as defined in the original 

design, is acceptable. The AOV action plans 

appeared to be comprehensive and indicated 

awareness of the problems associated with 

AOVs, the diagnostic tools available in the 

industry, and the work of various organizations 

and utilities regarding AOVs. 

Recent licensee efforts are being driven by 

the Maintenance Rule, recent industry-wide 

correspondence, a number of events involving 

AOVs, results from their IPE and PRA 

investigations, and/or experience gained from 

the MOV programs established to meet the 

recommendations of Generic Letter 89-10. Some 

licensees have not started these efforts while 

others are well on their way, but no licensees 

visited had completed their investigations. A

similar situation exists with regard to licensees' 

determination of the margins that exist within 

their AOVs to meet the design basis demands. 

Concerns include the quality of aging 

assessments and various failure mechanisms that 

the plants may not be considering in their 

assessments of the available margins in AOVs. 

There is no uniform schedule to implement 

AOV program plans in nuclear plants. 

16.8 Industry Initiatives 

Regarding AOVs 

The nuclear power industry is aware of the 

significance of AOV operability as a logical 

extension of their work to ensure the design 

basis operability of motor-operated valves, as 

well as a result of the work done to prioritize the 

importance of equipment in nuclear power 

plants, in order to comply with the requirements 

of the Maintenance Rule. A number of industry 

organizations including the Electric Power 

Research Institute, the Air-Operated Valve 

Users Group, and the Air-Operated Valve Joint 

Owners Group (JOG AOV) recently produced 

and distributed an industry-wide program plan to 

"provide assurance that AOVs are capable of 

performing their intended safety-significant, i.e., 

risk significant functions." 

16.9 The Maintenance Rule 

Licensees are addressing maintenance 

practices to meet the requirements of the 

Maintenance Rule with respect to the operability 

of safety-related and important non-safety

related systems and equipment in their plants 

using a variety of risk related tools and 

techniques. The system of establishing goals for 

systems and equipment found to have 

operational problems appeared to be effective (at 

least as far as AOVs are concerned) in the plants 

visited. In general, licensees at the plants visited 

are using techniques developed in response to 

the Maintenance Rule to cover equipment 

outside, and in addition to, the scope of NRC's 

interest in order to increase the reliability and 

efficiency of their plants. As a result of their 

Maintenance Rule activities, licensees are 

increasingly aware of the importance of AOVs,

NUREG/CR-6654 86
Conclusions

in terms of plant safety, availability, and heat 

rate. 

16.10 Diagnostic Systems, 

Maintenance, and Testing 

of AOVs 

Diagnostic testing (computerized 

measurement and software) tools are used by all 

licensees for motor-operated valves and appear 

to be valuable for determining maintenance or 

aging-related valve or actuator problems. 

Modem AOV diagnostic tools are now 

becoming available. Extrapolating no-load test 

data to design basis conditions remains to be 

verified as acceptable, just as with motor

operated valves. 

Claims of the diagnostic system vendors 

regarding accuracy and abilities of their systems 

have not been independently verified, along with 

assumptions about calculated forces such as 

packing loads and side loads. 

The ASME Section XI stroke-timing test 

alone does not provide assurance of future 

operability of AOVs. 

16.11 Risk Analysis Involving 

AOVs 

AOVs have been examined in several risk 

studies. The findings in the study by Pickard, 

Lowe and Garrick, NSAC-128 (see Section 15.1 

herein) described the importance of the air 

system and emphasized the importance of 

providing clean, dry, oil-free air at specified 

pressure to AOVs. Recently, NUREG/CR-6644 

(see Section 15.3.2 herein) indicated that at 

some plants AOV reliability can have important 

effects on system reliability and plant risk. 

In NUREG/CR-5472, "A Risk-Based Review 

of Instrument Air Systems at Nuclear Power 

Plants," the researchers studied the IA systems 

in nuclear power plants in terms of their 

contribution to risk. In general, the conclusions

and recommendations in NUREG/CR-5472, 

support the findings in this current study. See 

Section 15.1 for a discussion of those 

conclusions and recommendations. 

An Accident Sequence Precursor analysis 

(Section 15.2 and Appendix B) indicates the 

relative risk significance of a number of events 

involving AOVs. There were 26 precursor 

events involving AOVs, 12 of which had 

CCDPs > 1E-4. For several years, AOV 

precursor events were in the top quarter of all 

precursor events. The highest CCDP for an 

AOV event was 9E-4, that being the common

cause failure of all trains of AFW at both Turkey 

Point units due to water contamination of the 

instrument air system. See Section 8.7.1 herein 

and Section 5.1.2 in NUREG-1275, Volume 2. 

The operating experience gathered as part of 

this study, and observed in the plants during the 

recent visits, indicated that the detailed reviews 

conducted by each plant (see Section 15.3.1) 

disclosed AOVs which are important and risk 

significant. Many of the AOV events, failures, 

and deficiencies were not reported to the NRC 

and, as a result, were not reflected in the data 

used for risk analyses. A number of examples 

are included in Tables 2, 3, and 4, of this study. 

Also, recent risk studies (see Section 15.3.2) 

appear to have underestimated AOV risk. They 

were prepared using models which did not 

account for air system contamination, air system 

overpressurization, or other conditions which 

could cause AOV common-cause failures. 

Many of the licensees' individualized plant 

reviews for Maintenance Rule evaluations and 

for the plants' AOV programs, included reviews 

of PRA findings by expert panels. The licensees 

found that by using their plants' operating 

experience and plant-specific probabilistic risk 

assessments, certain AOVs were found to have 

high risk importance, high risk achievement 

worth, or were important to preventing large 

early releases. 

Generally, current risk models and databases 

either do not model the air systems, or do not 

include consideration of malfunctions of the air 

systems (i.e., air systems that do not consistently 

provide clean and dry air at proper pressure).

87 NUREG/CR-6654
Conclusions

These models may also not include 

consideration of shutdown modes, not include 

(or model accurately) a number of common

cause AOV failure scenarios, or they may 

include overly optimistic nonconservative AOV 

failure rates (Section 15.5). 

Some of the PRAs lump the AOVs with 

systems or consider skid mounted equipment in 

one single failure scenario, and the dependency 

on AOVs may not be apparent in such cases. 

Non-safety-related air systems are assumed to

fail or shut down upon loss of offsite power in 

most cases and the AOVs that provide a safety 

function are assumed to move to, or remain in a 

"fail-safe" position if that occurs. Alternatively 

accumulators are provided to allow the AOVs to 

be manipulated if the air systems are unavailable 

and these accumulators are assumed to provide 

the necessary quantities of clean, dry air. These 

assumptions are predicated on the performance 

of equipment which is designed, installed, 

qualified, maintained, and tested to meet the 

required demands.

NUREG/CR-6654 88
17. SAFETY SIGNIFICANCE

The following findings and conclusions of 

this study are considered to be safety significant: 

There have been a number of events 

involving AOVs and reported conditions 

of AOVs that indicate that AOVs have 

reduced operating margins caused by such 

factors as aging, load mechanisms not 

understood or considered in the original 

design, or previously contaminated air. 

Licensees may not know if the design 

basis loads or environmental conditions 

can be met with acceptable margins for 

the AOVs in their plants. Licensees are 

either relying on manufacturers' informa

tion and calculations or, more recently, 

design reviews of the AOV design bases. 

Several manufacturers' AOV calculations 

or valve descriptions were found that 

included mistakes or inaccurate informa

tion. The knowledge (or lack thereof) of 

the design bases and margins in AOVs to 

meet the design basis demands can be a 

matter of safety significance. 

Air systems are potential sources of 

common-cause failures of AOVs. High 

quality air, i.e., clean, dry, oil-free air, at 

specified pressure, is essential to accept

able performance of AOVs and SOVs. A 

reliable supply of high quality air is 

required at all times in order to ensure that 

AOVs will be able to move to or remain 

in their fail-safe positions. 

Accumulators in the air system that sup

ply air to safety-related and important 

non-safety related AOVs can be sources 

of contamination, reduced capacity, and 

subsequent AOV inoperability unless it is 

verified that the accumulators are free of 

contamination, do not contain trapped 

water, and the accumulator check valves 

are functioning. Licensees need to know 

that the accumulators are functioning 

properly in order to justify the 

assumptions that AOVs will move or 

remain in their fail-safe positions.

" Air-operated dampers are potential 

sources of failure of emergency diesel 

generators, control room ventilation sys

tems and other safety-related systems that 

they may serve. 

"* SOVs are potential sources of common

cause failure of AOVs. Root causes of 

failure include potential contamination 

from the air system, design, qualification, 

and maintenance. These causes have been 

reported previously and need to be sys

tematically addressed by licensees based 

on the safety status and risk significance 

of the SOVs. 

"* The operating experience gathered for this 

study and observed in the plants during 

the recent visits indicates that many 

AOVs are risk significant. During the 

plant visits, we learned of many common

cause failure events and deficiencies that 

had been reported in plant condition 

reports or plant deficiency reports but 

were not reported in LERs or in other 

reports to the NRC. As a result those 

common-cause failures were not included 

in common-cause failure databases. 

Therefore, factors used in risk analyses 

for estimating the effects of common

cause failures of AOVs were nonconser

vatively affected by such under reporting. 

Incomplete modeling and quantification 

of common-cause air system failures and 

degradations appear to have resulted in 

under-estimates of the risk significance of 

AOVs. Individual AOVs may be found to 

be of low risk significance but the 

common-cause failure of two or more 

AOVs performing the same function may 

have considerable risk significance. 

Accident Sequence Precursor analyses 

indicate that there have been a number of 

risk significant events involving AOVs. 

There were 26 precursor events involving 

AOVs, 12 of which had CCDPs > 1E-4. 

For several years, AOV precursor events

89 NUREG/CR-6654
Safety Significance

were in the top quarter of all precursor 

events. The highest CCDP for an AOV 

event was 9E-4, that being the common

cause failure of all trains of AFW at both 

Turkey Point units due to water contami

nation of the instrument air system. This 

event is noted, although it occurred over 

14 years ago and before the publication of 

Generic Letter 88-14, because it demonstrates that AOVs can be important com

ponents that can significantly affect risk. 

The number and scope of NRC generic 

communications and studies of AOV and 

air systems provide an important indicator 

of the overall safety significance of these 

components and systems.

NUREG/CR-6654 90
18. REFERENCES

1. AEOD/E92-02, Insights from Common-Mode Failure Events, S. Israel, NRC/AEOD, 6/92, and 

Supplement 1, 2/93 (This report is referenced in NRC Information Notice 93-35). 

2. ANSI/ASME OM-17, Performance Testing of Instrument Air Systems in Light-Water Power 

Plants, American National Standards Institute/American Society of Mechanical Engineers. 

3. APED-5750, Design and Performance of General Electric Boiling Water Reactor Main Steam 

Isolation Valves, D. Rockwell, E. VanZylstra, General Electric Co., March 1969. 

4. BNL Technical Letter Report E-2071-T1-12-93, Safety Significance Of Inadvertent Operation Of 

Motor Operated Valves In Safety-Related Piping Systems In Pressurized Water Reactors, C. Ruger, 

J. Carbonaro, J. Higgins, W. He, Brookhaven National Laboratory, Revision 1, March 1995. 

5. BNL Technical Report A-3869-T52-10-91,"Safety Significance Of Inadvertent Operation Of 

Motor Operated Valves In Safety-Related Piping Systems In Boiling Water Reactors, C. Ruger, 

J. Carbonaro, J. Higgins, W. He, Brookhaven National Laboratory, October 1991. 

6. Elastomer Evaluation Guideline, Rev. 2, Commonwealth Edison Company, October 1994. 

7. EPRI NP-2381, Measurements and Comparisons of Generic BWR Main-Steam-Isolation Valves, 

B. Shawver, Electric Power Research Institute, July 1982. 

8. EPRI NP-5613, Procedures for Treating Common Cause Failures in Safety and Reliability Studies, 

Vol. 1, Procedural Framework and Examples, Pickard, Lowe, and Garrick, Electric Power 

Research Institute, February 1988. 

9. EPRI NP-6516, Guide for the Application and Use of Valves in Power Plant Systems, Stone and 

Webster Engineering Corp., Electric Power Research Institute, August 1990. 

10. EPRI TR-103237, Revision 2, MOV Performance Prediction Program Topical Report, Electric 

Power Research Institute, April 1997. 

11. EPRI TR-107321, Application Guide for Evaluation of Actuator Output Capability for Air

Operated Valves in Nuclear Power Plants, Electric Power Research Institute, June 1997. 

12. INEL-95/0550, GI-158: Performance of Safety Related Power Operated Valves Under Operating 

Conditions, P. McCabe and S. Khericha, Idaho National Engineering Laboratory, Draft Report 

submitted to the NRC, October 1996. 

13. Instrument Air System Review Report, 47-1165965-00, Safety and Performance Improvement 

Program (SPIP), The B&W Owners Group, December 1986. 

14. ISA Handbook of Control Valves, 2nd Ed., J.W. Hutchison, Instrument Society of America, 1976. 

15. ISA-$7.0.01-1996, Quality Standard for Instrument Air, International Standards Institute, 

November 1996. 

16. NEDC-32264-A, Class 2, Revision 2,"Application Of Probabilistic Safety Assessment To Generic 

Letter 89-10 Implementation, BWR Owners Group, October 1, 1996.

91 NUREG/CR-6654
References

17. NMAC NP-7079, Instrument Air Systems - A Guide for Power Plant Maintenance Personnel, 

V. Vama, Nuclear Maintenance Applications Center of the Electric Power Research Institute, 12/90 

(a revision, NMAC TR-108147, is in preparation). 

18. NMAC NP-7412, Maintenance Guide for Air-Operated Valves, Pneumatic Actuators, and 

Accessories, Fossil Technologies, Inc., Nuclear Maintenance Applications Center of the Electric 

Power Research Institute, July 1992. 

19. NMAC NP-7412R1, Air-Operated Valve Maintenance Guide, Altran Materials Engineering, Inc., 

Nuclear Maintenance Applications Center of the Electric Power Research Institute, November 

1996. 

20. NMAC NP-7414, Solenoid Valve Maintenance and Application Guide, Nuclear Maintenance 

Applications Center of the Electric Power Research Institute, April 1992. 

21. NRC Letter from A. Thadani, NRR to T. Tipton, NEI, Review of EPRI Topical Report TR-103237, 

EPRI MOV Performance Prediction Program, Safety Evaluation Report (attached), March 15, 

1996. 

22. NSAC-128, Pneumatic Systems and Nuclear Plant Safety, Pickard, Lowe and Garrick, Inc., 

Nuclear Safety Analysis Center of the Electric Power Research Institute, October 1988. 

23. NSAC-137, Maintaining Operability of Nuclear Plant Instrument Air Systems, MPR Associates, 

Inc., Nuclear Safety Analysis Center of the Electric Power Research Institute, February 1990. 

24. NUREG-0800, Section 3.11, NRC Standard Review Plan, Environmental Qualification of 

Mechanical and Electrical Equipment, USNRC. 

25. NUREG-1275, Volume 2, Operating Experience Feedback Report - Air Systems Problems, 

H. Ornstein, USNRC, December 1987. 

26. NUREG-1275, Volume 6, Operating Experience Feedback Report - Solenoid-operated Valve 

Problems, H. Ornstein, USNRC, February 1991. 

27. NUREG/CR-1363, Data Summaries of LERs of Valves at U.S. Commercial Nuclear Power Plants, 

1/1/76-12/31/80, C. Miller, EG&G Inc., October 1982. 

28. NUREG/CR-2770, Common Cause Fault Rates for Valves: Estimates Based on LERs at U.S. 

Commercial Nuclear Power Plants 1976-1980, C. Atwood, EG&G Inc., February 1983. 

29. NUREG/CR-3154, The In-Plant Reliability Data Base for Nuclear Plant Components: Interim 

Report - The Valve Component, R. Borkowski, ORNL, December 1983. 

30. NUREG/CR-3424, Equipment Qualification of Research Test Program and Failure Analysis of 

Class 1E SOVs, D. Paulson, Franklin Institute, November 1983. 

31. NUREG/CR-3914, BNL-NUREG-51807, Pump and Valve Qualification Review Guide, B. Miller, 

BNL, October 1985. 

32. NUREG/CR-3960, Closeout of 1E Bulletin 80-01: Operability of ADS Valve Pneumatic Supply, 

W. Foley, Parameter Inc., June 1986.

NUREG/CR-6654 92
References

33. NUREG/CR-4217, A Statistical Analysis of Nuclear Power Plant Valve Failure Rate Variability, 

Some Preliminary Results, R. Beckman, LAND, July 1985. 

34. NUREG/CR-4674, Precursors to Potential Severe Core Damage Accidents, J. W. Minarick, et. al., 

SAIC and ORNL, November 1986 and later editions. 

35. NUREG/CR-4692, Operating Experience Review of Failures of PORVs and Block Valves in 

Nuclear Power Plants, G. Murphy, ORNL, October 1987. 

36. NUREG/CR-4780, EPRI NP-5613, Volume 1, Procedures for Treating Common Cause Failures in 

Safety and Reliability Studies, Volume 1: Procedural Framework and Examples, EPRI / Pickard, 

Lowe and Garrick, Inc., February 1988. 

37. NUREG/CR-4819, Vol. 1, Aging and Service Wear of SOVs Used in Safety Systems of Nuclear 

Power Plants, Operating Experience and Failure Identification, March 1987. 

38. NUREG/CR-4819, Vol. 2, Aging and Service Wear of SOVs Used in Safety Systems of Nuclear 

Power Plants, Evaluation of Monitoring Methods, R. Kryter, ORNL, July 1992. 

39. NUREG/CR-5008, Development of a Testing and Analysis Methodology to Determine the 

Functional Condition of SOVs, R. Meininger, Pentek Inc., September 1987. 

40. NUREG/CR-5140, Value Impact Analysis For Extension of NRC Bulletin 85-03 To Cover All 

Safety-Related MOVs, J. Higgins, C. Ruger, E. MacDougall, and D. Huszagh, BNL, July 1988. 

41. NUREG/CR-5141, Aging and Qualification Research on Solenoid-Operated Valves, V. Bacanskas, 

G. Toman, S. Carfagno, Franklin Research Center, August 1988. 

42. NUREG/CR-5292, Closeout of IE Bulletin 80-23: Failures of SOVs Manufactured by Valcor 

Engineering Corporation, W. Foley, Parameter Inc., February 1989. 

43. NUREG/CR-5295, Closeout of IE Compliance Bulletin 86-03: Potential Failure of Multiple ECCS 

Pumps Due to Single Failure of AOV in Minimum Flow Recirculation Line, W. Foley, Parameter 

Inc., October 1990. 

44. NUREG/CR-5419, Aging Assessment of Instrument Air Systems in Nuclear Power Plants, 

M. Villaran, R. Fullwood, M. Subudhi, BNL, January 1990. 

45. NUREG/CR-5472, BNL-NUREG-52220, A Risk-Based Review of Instrument Air Systems at 

Nuclear Power Plants, G. DeMoss, E. Lofgren, B. Rothleder / SAIC, M. Villeran, C. Ruger / BNL, 

January 1990. 

46. NUREG/CR-5497, INEELIEXT-97-01328, Common-Cause Failure Parameter Estimations, 

F. Marshall, INEEL, D. Rasmuson, NRC, A. Mosleh, U. of Maryland, October 1998. 

47. NUREG/CR-5640, SAIC-89/1541, Overview and Comparison of U.S. Commercial Nuclear Power 

Plants, P. Lobner, C. Donahoe, C. Cavallin/SAIC, September 1990. 

48. NUREG/CR-5500, Volume 3, General Electric Reactor Protection System Unavailability, 

1984-1995 (draft 2), S. Eide, M. Calley, W. Galyean, C. Gentillon, S. Khericha, T. Wierman, 

Idaho National Engineering and Environmental Laboratory, October 1998.

NUREG/CR-6654 93
References

49. NUREG/CR-6016, ORNL-6748, Aging and Service Wear of Air-Operated Valves Used in Safety

Related Systems at Nuclear Power Plants, Oak Ridge National Laboratory, May 1995. 

50. NUREG/CR-6246, Effects of Aging and Service Wear on MSIVs and Valve Operators, R. Clark, 

ORNL, May 1996. 

51. NUREG/CR-6268, Common-Cause Failure Database and Analysis System, F. Marshall, INEEL, 

D. Rasmuson, NRC, June 1998. 

52. REGULATORY GUIDE 1.68.3, Preoperational Testing of Instrument and Control Air Systems, 

USNRC, 4/82 

53. REGULATORY GUIDE 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power 

Plants, June 1993. 

54. TRIP REPORT, AOVIMOV Users Group Combined Meeting of 12/1-5/97, LMITCO 

Memorandum MRH-01-98, from M. Holbrook, LMITCO, to T. Scarbrough, NRC/NRR, December 

1997. 

55. Westinghouse Engineering Report V-EC-1658-A, Risk Ranking Approach For Motor-Operated 

Valves In Response To Generic Letter 96-05, G. Andre, L. Ezekoye, Westinghouse Electric Co., 

Revision 2, July 1998. 

56. Joint Meeting of the AOV/MOV Users Group, December 2-5, 1997, Clearwater Beach, Florida 

(14th Annual AOV Users Group Meeting), Mike Davido, PG&E, Chairman. 

57. Letter to H. Ornstein, NRC, from 0. Rothberg, INEEL, dated December 1, 1998, Transmittal of 

Sensitivity of Nuclear Plants to Core Damage Resulting from Failures of AOVs. 

58. NRC Generic Issue 43: Reliability of Air Systems, Revision 2 (refer to the description in Chapter 3 

of NUREG-0933). 

59. NRC Generic Issue 158: Performance of Safety-Related Power-Operated Valves Under Design 

Basis Conditions (refer to the description in Chapter 3 of NUREG-0933). 

60. Letter to E. Imbro, NRC, from D. Modeen, NEI, dated July 19, 1999, Joint Owners Group Air 

Operated Valve Program Document. 

61. Joint Owners Group Air Operated Valve Program, The Joint Owners Group AOV Committee, 

Revision 0, March 9, 1999 (DE&S Doc. 575.0.0.F10.01). 

62. Letter to D. Modeen, NEI, from E. Imbro, NRC, dated October 8, 1999, Comments on Joint 

Owners' Group Air Operated Valve Program Document. 

63. Control Valves for the Chemical Process Industries, Bill Fitzgerald, McGraw-Hill, Inc., 1995.

NUREG/CR-6654 94
Tables

95 95 NUREG/CR-6654
Table 1. Populations of air-operated valves in plants visited. 

Plant Name Safety-Related AOVs Category I AOVs Category 2 AOVs Category 3 AOVs GL 89-10 MOVs

Palo Verde 1, 

2, and 3 

Fermi 2 

Palisades 

LaSalle 1 

and 2

41 + 131 = 172 AOVs 

per plant are classified 

by the licensee as safety

related. See Category 1 

and 2. 

29 AOVs in Category I 

and 34 AOVs in 

Category 2 (are 

safety-related, as well as 

370 AOVs for Scram 

inlet and outlet 

valves.(There are also 

2482 SOVs of which 

1442 are classified by 

the licensee as QAI.) 

191 AOVs. 

84 for both units. In 

addition, 370 CRD 

valves in each unit are 

classified by the licensee 

as safety-related.

41 AOVs per plant are 

classified by the licensee as 

Category 1. The licensee 

refers to active safety

related AOVs as 

Category 1. 

410 AOVs are classified by 

the licensee as Category 1. 

The licensee refers to 

AOVs having "high 

safety-significance" as 

Category 1. Included are 

370 SCRAM inlet and 

outlet valves, 29 other 

safety-related valves, and 

11 AOVs that support a 

non-safety-related, risk 

significant function. 

111 AOVs. Valves in this 

category are safety-related 

with active safety functions, 

important-to-safety based 

on their PSA risk 

significance, or included 

based on Expert Panel 

determinations. 

AOVs having high safety 

significance. Number not 

provided.

131 AOVs per plant are 

classified by the licensee as 

Category 2. The licensee 

refers to non-active 

safety-related AOVs as 

Category 2. 

84 AOVs are classified by 

the licensee as Category 2, 

including 34 safety-related 

AOVs. The licensee 

designates as Category 2 

those less safety significant 

AOVs that support safety

related functions or have 

relatively high economic 

consequences if they should 

fail. 

42 AOVs are classified by 

the licensee as Category 2. 

These AOVs are 

safety-related but of low 

risk significance or non

safety-related but used in 

"critical" applications. 

AOVs having low safety 

significance. Number not 

provided.

Approx. 2628 AOVs per plant are classified by the licensee as 

Category 3. The licensee refers 

to non-safety-related AOVs as 

Category 3. 

Category 3 AOVs are those 

"having little or no safety 

significance or economic 

consequences." (Note: The 

original 1995 rough outline for 

development of the Fermi 2 

AOV program lists a total of 

2058 AOVs, of which 598 were 

considered safety-related valves 

or dampers and 1460 were 

considered non-safety-related 

valves or dampers.) 

Approximately 561 AOVs, 

which are not Category 1 or 2 

arc classified by the licensee as 

Category 3 AOVs. 

AOVs having high economic 

significance. Number not 

provided. (LaSalle categorizes 

AOVs with no or limited 

safety/economic significance as 

Category 4.) (There are 

1575 non-safety-related AOVs 

for both units.)

There are 831 MOVs on site (3 plants) of 

which 336 are in the 

GL 89-10 program. 

147 MOVs are in the 

GL 89-10 program. 

There are 54 MOVs in 

the plant of which 30 

are covered by 

GL 89-10. 

There are 200 MOVs 

in the GL 89-10 

program, for both 

units.

C/)

z 

0'

C,'
Table 1. (continued).

Plant Name Safety-Related AOVs 

TMI 1 98 AOVs are classified 

as safety-related 

(designated "Q-class" or 

"Class 1") by the 

licensee.

z 

C)

263 AOVs are classified 

as safety-related by the 

licensee. 

The licensee classified 

191 AOVs (total for 

both units) as safety

related.

Category I AOVs 

98 AOVs are categorized as 

Class I by the licensee. 

These are AOVs with an 

active safety function. 

The licensee did not classify 

AOVs as Category 1, 2, 

or 3. (215 AOVs were 

classified by the licensee as 

being within the scope of 

the Maintenance Rule, 

10 CFR 50.65.) 

174 AOVs (98 active, 

76 passive, total for both 

units) are classified by the 

licensee as Category 1.

Categoly 2 AOVs_ ( 

328 AOVs are categorized 484 AO 

as Class 2 by the licensee. Class 3 

These are AOVs with an are AO• 

EOP function or operational or 2.The 

economic significance. 910 AO 

The licensee did not classify The lice 

AOVs as Category 1, 2, AOVs a 

or 3. (There a 

plant, th 

315 AO 

related.) 

53 (34 active, 19 passive, There ar 

total for both units) are units. It 

classified by the licensee as licensee 

Category 2. some A(

ategory 3 AOVs

Vs are categorized as 

by the licensee. These 

Vs not categorized I 

ere are a total of 

Vs at TMI- 1. 

nsee did not classify 

s Category 1, 2, or 3. 

ire 578 AOVs in the 

erefore, 578 - 263 = 

Vs are non-safety

e 836 AOVs in both 

is not known if the 

specifically designated 

WVs as Category 3.

GL 89-10 MOVs

There are 81 MOVs in 

the GL 89-10 program 

for this plant. 

89 MOVs are within 

the scope of GL 89-10. 

111 MOVs (total for 

both units) are within 

the scope of GL 89-10.

00

Notes: 

I. The category designations in the table vary from plant to plant. The use of the categories for each plant is explained with the entry. 

2. There may be SOVs in the plants that are classified as part of the AOV. Figures for SOVs were included if separate data was provided.

Indian Point 3 

Turkey Point 

3 and 4

C,...
Table 2. Selected events documented by NRC Licensee Event Reports (LERs). 

LER Event 

Plant Number' Date Description Classification2 System

900330 Operation at 100% reactor power, operators noted closure of the "C" Main 

Feedwater Regulating Valve, FCV-FW-498. The operator tried to manually 

open FCV-FW-498; however, the valve would not respond. A reactor trip on 

"SG 'C' Low Level & Feedwater Flow Low" occurred. 

The cause for the reactor trip was the closure of FCV-FW-498, in response 

to insufficient instrument air pressure supplying the valve positioner. 

Moisture in the instrument air supply had plugged the filter in the air 

regulator for the current-to-pneumatic valve positioner for FCV-FW-498, 

causing the valve to close. This moisture was present in the instrument air 

system as a result of the instrument air dryer being out of service. While the 

air dryer was out of service, the air compressors' discharge was directed 

through the instrument air bypass filters which did not remove the moisture. 

This is considered to be a common-cause failure condition.

a, c Main Feedwater 

Instrument Air

Calvert Cliffs I 31788009 880824 The unit tripped on Loss of Load when the Main Turbine tripped on high 

Steam Generator level. The Main Turbine tripped when the air line on #12 

Main Feed Regulating Valve failed and the valve failed to the full open 

position causing a high level in #12 steam generator. The air line to the feed 

regulating valve failed due to vibration and stress. 

The 1/4 inch instrument air line nipple at the junction of #12 MFRV's valve 

positioner sheared due to cyclic stress and fatigue. The improper location of 

a pressure switch in the instrument air supply line contributed to the total 

stress on the nipple. The weight of the pressure switch, along with feed 

header-induced vibrational stress, eventually caused the nipple to fail. Air 

pressure on the top and bottom of the valve diaphragm bled off 

simultaneously, and pressure beneath the valve plug from main feedwater 

forced it up, thus fully opening the valve. Normally; the valve will fail in the 

"as is" position if air pressure decreases to less than 70 psig as seen at the 

pressure switch. In this event the loss of air pressure was downstream of the 

pressure switch and thus the valve moved, unexpectedly, to the full open 

position.

c Main Feedwater

Beaver Valley 1 33490007

,

0"
Table 2. (continued).

Plant

Calvert Cliffs 1

Calvert Cliffs I

LER Event 

NumberI Date Description Classification2

z 

ON 

ON

31789005 890314 With the Unit operating at 100% power, a partial loss of Instrument Air (IA) c

occurred due to a failed check valve. The low IA event was ended when the 

Air Compressors were started and restored IA pressure. The event was 

caused by the wrong check valve type installed in the IA system causing the 

valve internals (flapper and swing hinge) to fail due to improper wear 

pattern. 

The repair and leak testing of the similar replacement check valve showed 

that the valve type was inappropriately chosen for its intended application. 

The existing check valve was primarily designed to function with a higher 

header pressure (i.e., about 250 psig instead of an actual operating pressure 

of 100 psig) and uses a hard seat and disk. The check valve needs the higher 

differential pressure and rapid flow reversal to ensure an airtight closure. 

This valve was to have been replaced with a different check valve better 

suited for its intended use. Therefore, the root cause of this event was the 

valve type being inappropriately chosen for its intended application. 

This was an ASP Program Event. See Appendix B. 

31789018 891106 A condition was discovered that could have prevented the fulfillment of 

certain systems to remove residual heat and control the release of radioactive 

material after a Loss of Coolant Accident (LOCA). During the performance 

of a test to satisfy specific requirements in Generic Letter 88-14, it was 

discovered that many air-operated control valves and piston-operated 

ventilation dampers which utilize safety-related air accumulators would not 

have performed as expected after a loss of normal non-safety-related 

instrument air. The root cause of the event was identified as a lack of an 

adequately documented design basis combined with inadequacies in the 

testing and preventative maintenance program for the Instrument Air 

System. This is considered to be a common-cause failure condition.

System

Instrument Air

6 Different 

Systems 

Instrument 

Air

C

0 

0

(tq
Table 2.- (continued).

Classification2

41397002 970506

46190004 900703

Catawba I Feedwater Containment Isolation valve I CF-51 was inoperable due to c

actuator low nitrogen pressure. Each Steam Generator Feedwater 

Containment Isolation Valve is a pneumatic-hydraulic controlled gate valve 

whose safety related function is to terminate flow in either direction. The 

valve actuator is a Borg Warner #37981 that uses nitrogen pressure to close 

the valve. Per the Borg Warner instruction manual, the valve is inoperable 

when nitrogen pressure is below 2050 psig decreasing signal. The alarm 

pressure switch setpoint is calibrated at 2100 psig (+ or - 50 psig) and 

provides alarm indication through the Operator Aid Computer (OAC). On 

April 3, 1997, between the hours of 0608 and 2319, valve ICF-51 was 

inoperable due to low nitrogen pressure. This inoperability time exceeded the 

10-hour time limit allowed by Technical Specifications. 

The root cause of this event was inadequate information in the OAC Alarm 

Response. The Alarm Response did not indicate the need for immediate 

action and did not include setpoint information that would have led the 

operators to recognize the close proximity of the alarm setpoint to the 

minimum allowed nitrogen pressure for valve operability. 

Investigation of information notice 88-24 revealed that the maximum 

operating pressure differential (MOPD) of 73 solenoid-operated valves 

(SOVs) supplying air to active safety-related air operated valves (AOVs) and 

dampers was less than the maximum instrument air (IA) system pressure 

(max IAP). The SOVs would be subjected to max IAP in the event of air 

regulator (AR) failure. Loss of air had been previously considered as a 

failure mode; over pressurization had not. Further investigation identified 

that AOV components and other safety-related end-use devices also have the 

potential for over pressurization if their upstream ARs fail open. The over 

pressurizations (AOV components, SOVs, and other end-use devices), may 

result in failure of safety-related devices to reposition to their safety 

positions. The potential for over pressurization was the result of errors by 

vendors, the nuclear station engineering department, and the architect 

engineer (AE), and miscommunication with the AE. Corrective actions 

included installing rupture disks and/or high MOPD SOVs, and replacing 

ARs with safety-related, seismically qualified, ARs. The potential over 

pressurization of the AOVs and two end-use devices was determined to be 

reportable in accordance with the provisions of 10 CFR Part 21

Feedwater 

Containment 

Isolation Valve

Miscellaneous Active 

Safety-Related 

Systems

This is considered to be a common-cause failure condition.

LER N.T.. 1 1

b, c

Event ~T_% 1fl~crintion System

E

Clinton

7z 

m'

Table P'lant 2. 1Nume Ua D•I Ldte Y_• "... (continued).
Table 2. (continued).

LER Event 

NumberI Date Description Classification2

z 

61 a, 

C)

Comanche Peak 1 44595005 950831 Engineering personnel identified nonconservatism in the calculation that 

determined (1) leakage rates for accumulator check valves associated with 

various air operated valves and the nitrogen accumulators for the pressurizer 

Power Operated Relief Valves (PORVs), and (2) the pressure switch alarm set 

points for these accumulators. Engineering personnel performed evaluations 

which revealed that, with the exception of the PORVs, the valves associated with 

these accumulators were operable. The PORV accumulator low pressure alarm 

set points would still ensure operability during Modes 1, 2 and 3. However, for 

Modes 4, 5 and 6, the set points would not ensure operability for all conditions. 

TU Electric believed that the cause of the event was nonconservative design in 

the original calculation used to determine the low pressure alarm set point for the 

PORV accumulators and the accumulator check valve leakage rates. 

The PORV nitrogen accumulator low pressure alarm set points were to be raised 

to 90 psig. A review was to be performed for all safety-related accumulators used 

for valve actuation to verify that the appropriate set points are being used. 

This is considered to be a common-cause failure condition.

c Pressurizer 

Power Operated 

Relief Valves

Clinton

Plant

46198009 980203 Engineering personnel determined during their review of a I0CFR21 

notification by Engine Systems, Incorporated, that the air start solenoid pilot 

valves for the emergency diesel generators (EDG) would not operate as 

required by the design basis. The solenoid valves, which are used to valve air 

to the air operated main valves that supply starting air to the start motors on 

all three EDGs, would not operate reliably at the low end of the design basis 

air start system pressures and DC voltages. The apparent cause of the 

inadequate design of the solenoid valve was a failure of the design engineer 

to consider the full range of design basis design pressures and DC voltages 

when changing the spring size of the solenoid pilot valve. Corrective action 

for this event included: changing the spring size in the solenoid valve for the 

Division III EDG air start System; replacing the Division I and II EDG air 

start system solenoid valves, or otherwise modifying the system to meet the 

design basis requirements; and revising the annunciator procedures for the 

EDG air start system receiver low pressure alarm to require that the EDGs be 

declared inoperable when supply air pressure drops below 200 psig. 

There are 335 SOVs, distributed among about 17 nuclear plant owners, listed 

in the Part 21 notice. Point Beach LER 26698008 also refers to Engine 

Systems Part 21 Notification # 1998120 dated January 26, 1998. 

This is considered to be a common-cause failure condition.

b, c

H

System

EMDs
Table 2. (continued).

LER 

Plant Number' 

Comanche Peak 1 44598001

Cooper

Event 

Date Description

980110 During performance of a partial stroke test of feedwater isolation valves 

(FWIVs), the Plant Equipment Operators (utility, non-licensed) discovered 

that Feedwater Isolation Valves I -HV-2134 and 2-HV-2137 had low 

nitrogen pressure. The Shift Manager declared that both FWIVs I-HV-2134 

and 1-HV-2137 were inoperable. Maintenance personnel proceeded to 

charge the accumulators with nitrogen. The low nitrogen pressure for FWIV 

1-HV-2134 was caused by a leak in a nitrogen solenoid valve and on the 

nitrogen cylinder head. The low nitrogen pressure for FWIV 1-HV-2137 was 

caused by a leak in a nitrogen solenoid valve. Immediate corrective action 

was to restore the nitrogen pressure to declare the FWIVs operable. The leak 

rate on 1-HV-2134 was reduced and work planned to replace elastomer parts 

for this valve. The solenoid valve on 1-HV-2137 has been replaced to 

prevent recurrence. The elastomer seal on FWIV I -HV-2134 was degraded 

and caused the nitrogen to leak. 

The LER does not describe the root cause for the SOV failure or the 

mechanism for the elastomer seal degradation.

29894013 940712 A Reactor Scram and Group 2 (Shutdown Cooling), 3 (Reactor Water 

Cleanup), and 6 (Reactor Building Ventilation) Isolations occurred and 

Standby Gas Treatment started as a result of a spurious decrease in the 

indicated Reactor vessel water level on the "B" channel level instruments. 

The cause of the spurious actuation was leakage through a solenoid valve 

due to wear. This solenoid valve provides injection of Core Spray to backfill 

the "B" channel reference leg of the Reactor vessel water level indication 

system. The wear was due to pressure transients from valve surveillance 

testing in the Core Spray system, which had not been anticipated. The SOV 

(NBI-SOV-SSV739) is rarely operated, and the wear found would not be 

expected from normal operation. Root cause investigation indicated that the 

condition of the valve was due to pressure transients from the surveillance 

testing of CS-MOV-MO26B with the Reactor depressurized. The solenoid 

valve is designed to use process system pressure to assist in obtaining a tight 

shutoff, a condition which is not attainable with the Reactor depressurized. 

The selection of this valve design for the system did not appropriately 

anticipate the spectrum of operational conditions to which it could be 

exposed when the equipment was installed in 1989.

Classification2 System

c Feedwater

c RCS, Core Spray

C) 

nO 

('A a-E
Table 2. (continued).

Plant

Crystal River 3

LER Event z Number' Date Description 4C 

7m 

C-) 

aN ON

Classification 2

c

System

Containment

34687015 871207 A low pressure instrument air pressure alarm was received in the control 

room. The low instrument air pressure caused several valves in the steam and 

feedwater systems to open and dump steam and condensate to the condenser. 

The Integrated Control System responded by increasing feedwater flow and 

pulling control rods out. Reactor power increased and reached the high flux 

trip on the Reactor Protection System. The loss of instrument air pressure 

was caused by direct venting of the Instrument Air header to atmosphere 

when a solenoid valve failed on Instrument Air Dryers 1-I and l-2.The air 

dryers were isolated and bypassed in order to re-establish normal instrument 

air header pressure. No information was provided in the LER concerning 

how long the dryers were out of service or what the short or long-term 

effects on air quality might have been. According to the LER, the dryers 

were subsequently replaced. 

This is considered to be a common-cause failure condition.

b, c Main Steam 

Main Feedwater 

Instrument Air

30297015 970612 During a review of the differential pressure calculation for the Letdown Line 

[CB] Inboard Containment Isolation Valves, FPC discovered that the 

evaluation of the maximum differential pressure (d/P) that these valves could 

be subject to was in error. These valves are rated to close against a maximum 

d/P of 1800 psi, but could be subjected to a d/P in excess of 2000 psi in the 

event of a letdown line rupture downstream of outboard containment 

isolation valve (MUV-49), concurrent with a failure of MUV-49 to close and 

operator action in accordance with Emergency Operating Procedure (EOP) 

3. Outboard containment isolation valve MUV-49 would not be capable of 

closing if subjected to 2000 psi d/P. Isolation at Penetration 333 requires 

either MUV-49 or MUV-40/41/ 505 to close. The cause of this event was the 

use of inappropriate assumptions in the calculation for the determination of 

maximum valve diP. FPC planned to install a new inboard containment 

isolation valve prior to restart of CR-3. In addition, the air operator on MUV

49 was to be modified to allow closure against the predicted d/P. 

The closure of these valves is required to mitigate the effects of a Makeup 

System Letdown Line Failure Accident and in response to a Reactor 

Building Isolation signal. Consequently, the valves were outside of their 

design basis and their failure to close in the described scenario during 

previous operating periods, could have created an non-isolatable loss of 

reactor coolant to the Auxiliary Building. 

This is considered to be a common-cause failure condition.

0

Davis Besse
Table 2. (continued).

LER Event 

Number' Date Description

34688007 880304 Several AOVs were declared inoperable due to the potential for moving from 

their fail-safe position upon Loss of Instrument Air following a Safety 

Features Actuation System Initiation. Service Water Valve SW 1434 and 

four other valves of similar design require air pressure to hold them in their 

"fail-safe" position. These valves are SW 1424, SW 1429, CC 1467, and 

CC 1469 which are the Temperature Control Valves for two of the 

Component Cooling Water Heat Exchangers and Decay Heat Removal Heat 

Exchangers respectively.

Classification2

C

System

Decay Heat

Decay Heat Removal 

Instrument Air

Each valve operator is similar and is provided with two safety-grade 

accumulators. One accumulator is piped to the bottom of the valve actuator 

piston and provides the motive force to open the valve when air is vented 

from the top of the actuator piston. The second accumulator is piped to the 

top side of the actuator piston through two solenoid valves and provides the 

motive force to close the valve. The cause of the valve drifting from its open 

position was air leakage from the valve's accumulator system. The cause of 

the accumulator leakage was not described in the LER.A mechanical locking 

device was to have been added which, on the loss of air, will automatically 

lock its associated valve in the fail-safe position subsequent to the valve's 

receiving an SFAS initiation signal.

31597026 970925

This is considered to be a common-cause failure condition. This was an ASP 

Program Event. See Appendix B. 

Due to a lack of overpressure protection on the 85, 50, or 20 psig control air 

headers, if a non-safety-related air regulator failed open it would result in an 

overpressurization of a control air header. This would result in the potential 

for common mode failure of both trains of safety related equipment. The lack 

of overpressure protection on the control air headers due to a regulator 

failing open had not been identified as a mechanism that could 

overpressurize the low pressure headers. As a result, single failure of a non

safety related component affecting both trains of safety related equipment 

was not identified. 

There would have been no significant effects for the 85 and 50 psig headers, 

however, overpressurization of the 20 psig header could have resulted in the 

degradation of the RHR system and the partial opening of the Unit 2 Steam 

Generator (SG) Power Operated Relief Valves (PORVs) for the duration of 

the overpressure event. Due to a single failure being identified that could 

have potentially resulted in the degradation of both trains of RHR this event 

could have been significant. (continued next page)

c RHR HX Outlet 

Valves 

Steam Generator 

PORVs

Plant

Davis Besse

D. C. Cook

7, 

Cz 

m 

C) 

ON

Table 2. (continued).
Table 2. (continued). 

LER 

Plant Number' 

D. C. Cook 31597026 

(continued)

Event 

Date

970925

23787023 870717

23788012 880517

L7 Description

.2:, 

tm 

,-) 

QN

Classification 2 System 

RHR HX Outlet 

Valves 

Steam Generator 

PORVs

c

The cause of the lack of overpressure protection on the Control Air System 

was the fact that a regulator (non-safety-related) failing open was not 

identified as a mechanism that could overpressurize the low pressure 

headers. As a result, single failure of a non-safety related component 

affecting both trains of a safety related system was not identified. 

Four component failures involving Copes-Vulcan P-200-12 and 

B&W-B-209-12 feedwater regulating valves, as well as booster valve and 

positioner failures led to a scram. Subsequent investigation revealed a 

number of problems, including a broken handwheel stop, air restrictions in 

an SOV due to wear, a positioner out of adjustment, and improper booster 

relay valve diaphragm seating caused by either age or wear. 

The Component Failure Data section of the LER indicated that an 

industry-wide search revealed 81 failures in one year of Copes-Vulcan 

Model P-200 AOVs due to positioner operating abnormalities. The B&W 

operators had 6 failures, 5 of which were SOV related. 

This is considered to be a common-cause failure condition. 

All eight Main Steam Isolation Valves (MSIVs) were declared inoperable 

due to their failure to fully close on a loss of pneumatic supply. The root 

cause of the MSIVs failure to fully close has been attributed to high drag 

forces exerted on the valve stem by the valve packing. 

This is considered to be a common-cause failure condition.

Main Feedwater

Main Steam 

Isolation Valves

c

Dresden 2

,.. 

rj

ON

Dresden 2
Table 2. (continued).

DescriDtion Classification 2 Plant 

Dresden 2

There are over 1900 Copes-Vulcan D100 valves in service in U.S. nuclear 

power plants. This is considered to be a common-cause failure condition. 

24993004 930116 At 83% power, an Instrument Air Header Pressure low alarm was received in 

the Control Room. Instrument Air Header pressure was observed to be 

decreasing rapidly. Loss of Instrument Air was attributed to mechanical failure 

of the 3A Instrument Air Compressor dryer inlet valve to close during the 

dryer purge cycle with concurrent failure of the backup Service Air to 

Instrument Air Cross-tie valve to promptly open on a low instrument air 

pressure signal. 

Based on bench testing it was determined that inlet valve binding in the open 

position was the root cause for dryer blowdown. Binding was attributed to 

excessive friction in the valve ball. 

The root cause for the cross-tie valve slow response time was attributed to 

malfunction of its air pressure regulator, due to binding of internal parts. It was 

also noted in the LER that a solenoid valve needed to be relocated. Prior to 

returning the 3A Instrument Air Compressor to service the following 

corrective actions were taken: 

- replacement of the 3A dryer inlet and exhaust valves, 

- replacement of the inlet and exhaust solenoid valves, and 

- replacement of the cross-tie valve air pressure regulator and solenoid 

relief valve.

b, c Scram Header 

Instrument Air

980128

Event 

Date

LER 

Number' 

23798003 

(Event 

No. 33620, 

dated 

1/28/98 

followed up 

by Morning 

Report 

H-98-0045, 

dated 

3/6/98 

referred 

to this 

condition.)

Failures in diaphragms of D-100 Copes-Vulcan AOVs were caused by wear 

attributed to inadequate rubber covering over the fiber in the diaphragms. The 

root cause of this event was manufacturer's design deficiency of the valve 

operator's diaphragm (NRC Cause Code B). The diaphragm which failed has 

the embedded reinforcing fiber layer too close to the surface on the lower side 

of the diaphragm. This significantly increases the susceptibility of the 

diaphragm to a mechanical abrasion failure because of rubbing between the 

diaphragm and the actuator plate. Essentially, the diaphragm wears out 

prematurely. This failure mechanism was also documented by the CoinEd 

System Materials Analysis Department (SMAD) for three other failed 

diaphragms of the same make and model from the Quad Cites Nuclear power 

-lant during detailed laboratory examinations. In 1996, the valve manufacturer 

(Copes-Vulcan) upgraded the diaphragm design by increasing the amount of 

rubber on the side of the diaphragm in contact with the diaphragm plate. This 

same valve (A03-2301-64, HPCI Turbine Stop Valve above seat drain valve) 

failed at Dresden 2 in March 1995 for what appears to be the same reason 

(LER 24995004). LER 24997003 described another failure of AOV 

A03-2301-64 due to misadjustment.

c

System

HPCI

..

Dresden 3

z: 

C• 

0•
Table 2. (continued).

LER Event 

Plant Number' Date Description 

Dresden 3 24993005 930126 An air leak occurred on the fail-safe accumulator system for Drywell Vent 

Valve 3-1601-63. Air was identified to be leaking out of a weep hole on the 

two way mechanical Versa Valve.

Duane Arnold 33191005 910622

The apparent cause for the leakage out of the weep hole in the two way 

Versa Valve was due to degradation of the O-ring seals caused by aging and 

fatigue of the O-ring material. 

With the reactor at approximately 100% power, a single outboard Main 

Steam Isolation Valve (MSIV) closed, resulting in a high flux automatic 

reactor scram.

Classification2

c

c

System

Drywell 

Environmental 

System

Main Steam 

Isolation Valves

The cause of the MSIV closure was determined to be a non-safety related 

pipe joint failure. The two-inch nitrogen supply pipe that supplies the 

outboard MSIVs' control accumulators separated sufficiently at a soldered 

coupling to reduce supply pressure. Although check valves are installed to 

maintain control pack accumulator pressure, two smaller fitting leaks on the 

control pack for the 'B' outboard MSIV slowly reduced the pressure to the 

nitrogen-operated MSIV position control valve. This caused the control 

00 valve to slowly change position, porting actuating nitrogen to close the 

MSIV.

34194004 940822 Four air-operated, spring closing isolation valves may not have been capable 

of performing their Reactor Coolant Pressure Boundary (RCPB) function of 

isolating against Reactor Coolant System (RCS) pressure at their as-found 

actuator spring preload settings. The cause of this event was inadequate 

control of the valve actuator settings which resulted in insufficient preload 

being maintained on the actuator closing springs. The spring preload settings 

were to have been increased to a value sufficient to ensure adequate valve 

closure prior to plant startup. 

Two industry events during 1993 involving the failure of air-operated, spring 

closing valves to adequately isolate against full RCS pressure prompted a 

review of similar valves at Fermi 2. Four, 3/4-inch Rockwell-Edwards globe 

valves with Fisher air actuators were Identified with a similar problem. 

These are normally open, spring to close valves which provide automatic 

containment isolation for reactor recirculation pump seal purge flow from 

the Control Rod Drive Hydraulic (CRDH) system. Isolation occurs on either 

a High Drywell Pressure or a Reactor Vessel Low Level 2 signal. The valves 

are oriented with normal system flow from the CRDH system over the seat, 

to the reactor recirculation pump seals which are at normal reactor pressure. 

(continued next page)

C Control Rod 

Drive Hydraulic 

System 

Reactor Coolant 

System

ON, 

4N

0•

Fermi 2
Table 2. (continued). 

LER Event 

Plant Numbers Date Description Classification 2 System 

Fermi 2 34194004 940822 In the event of an accident with a concurrent loss of CRDH system pressure, c Control Rod 

(continued) the valves could be subjected to a back pressure (under seat) differential of Drive Hydraulic 

up to about 1200 psid, which may either prevent the valves from fully System 

closing or allow them to lift off of their seats after they have closed. The Reactor Coolant 

original specification for these valves was to ensure closure in the CRDH System 

flow direction at a pressure of 1750 psig. This resulted in a setting by the 

vendor sufficiently high to meet both the containment isolation as well as the 

RCPB isolation requirements for these valves. However, the nameplate 

supplied by the vendor incorrectly identified a lower setting. The valves 

were reset during installation to this lower setting using the nameplate data. 

The resulting setting was too low to ensure adequate closure of the valves for 

meeting their RCPB isolation function. 

This condition had existed since initial operation of the plant. 

This is considered to be a common-cause failure condition.

28587018 870504 The valve operator system for HCV-385 and HCV-386 had design 

deficiencies which could have prevented them from performing their safety 

function. These are the isolation valves for the containment spray (CS) and 

High and Low Pressure Safety Injection (HPSI/LPSI) pump minimum 

recirculation line. They were designed such that a single failure in 

conjunction with a Loss of Coolant Accident (LOCA) and loss of the non

safety grade instrument air system would have prevented these valves from 

going shut. This failure would have allowed primary coolant to be pumped to 

the Safety Injection and Refueling Water Tank (SIRWT) which vents to the 

Auxiliary Building; thus, violating containment integrity. HCV-385 and 

HCV-386 rely on a safety grade accumulator to provide the motive force to 

close the valves. Two design deficiencies were identified. First, both valves 

shared a single accumulator. Thus, a single failure could disable the ability to 

close the valves. Second, the valves' required time to close is 45 seconds to 

ensure that the coolant from the containment sump does not reach the valves. 

If highly contaminated coolant reaches the two valves they could be 

inaccessible for up to 1000 hours, and the air accumulators would have been 

required to hold the valves shut for this full time period. If the valves close in 

less than 45 seconds, they must hold only long enough for the operators to 

locally isolate them by handwheel. Test data indicated that closure time had 

been at or near 45 seconds. This valve operator/air accumulator system 

deficiency was an original design deficiency present since the plant started 

operation in 1973. This is a common-cause failure condition.

Containment Spray 

High Pressure 

Safety Injection 

Low Pressure 

Safety Injection

Fort Calhoun

z 

0N 

ON 

U'

c
LER Event 

Plant Number' Date Description Classification2 System 

Fort Calhoun 28587033 870706 Clarified water entered the Instrument Air System during a surveillance test a, c Instrument Air

28588002 880125

of the diesel generator room dry pipe sprinkler system. The interfacing check 

valves between the Instrument Air and Fire Protection systems were 

prevented from completely closing by foreign material. Additionally, 

inadequate procedures and inadequate operator training on this unique dry 

pipe valve contributed to this event. 

This event occurred because: 

I. Instrument air check valves IA-575 and IA-576 were prevented from 

closing by foreign material. 

2. The operator performing the test failed to properly reset the dry pipe 

valve as a result of inadequate procedures and inadequate training on the 

unique dry pipe valve. The air maintenance device was bypassed thus 

removing another check valve and orifice that could have prevented 

and/or restricted flow of water into the air system. During the reset 

process, as performed, a flow path existed to the Fire Protection System. 

The Fire Protection System pressure is approximately 30 psi greater than 

Instrument Air pressure. Thus, water flowed into the Instrument Air 

System. 

This event was the cause of failure of the DG-2 exhaust damper and 

shutdown of DG-2 on September 23, 1987, which is addressed in 

LER 28587025 in Table 3. This is considered to be a common-cause failure 

event. 

The High Pressure Safety Injection alternate discharge header isolation 

valve, HCV-2987, was declared inoperable due to failure to meet the design 

criteria for operability. The purpose of HCV-2987 is to ensure long term core 

cooling through hot leg injection. It is a normally open valve that fails as is 

on a loss of instrument air. 

HCV-2987 uses an air intensifier to convert instrument air received at about 

40 psig to supply the valve receiver reservoir air pressure at 320 psig. A 

check valve prevents depressurization of this reservoir on a loss of 

instrument air. The valve then can be cycled closed one time without the use 

of instrument air. This is the valve's required position for simultaneous hot 

and cold leg injection for long term core cooling if the "B" HPSI pump is 

operating. The equivalent of closing HCV-2987 could be obtained by 

(continued on next page)

,-. 

Cl)-

Fire Protection 

System

c High Pressure 

Safety Injection

0

Fort Calhoun
Table 2. (continued). 

LER Event 

Plant Number' Date Description 

Fort Calhoun 28588002 880125 closing the four HPSI isolation valves located down• 

(continued) close to meet the single failure criteria for hot leg inj

Classification2 System 

stream. HCV-2987 must c High Pressure 

ection. Safety Injection

28588004 880311 

28588009 880406

It was determined that the air intensifier had failed (for reasons not included 

in the LER). A temporary mechanical jumper was installed to bypass the 

valve's air intensifier. A nitrogen gas bottle supplying 320 psig pressure to 

the valve's receiver reservoir was used to supply sufficient pressure for valve 

operations. 

During a LOCA, a potential leakage path through isolation valve, PCV-1849, c 

for the Instrument Air (IA) System Containment Penetration M could exist if 

the IA system pressure is not maintained above the containment pressure. 

PCV-1849 is designed to close on receipt of a low IA system pressure in 

conjunction with a Containment Isolation Actuation Signal (CIAS). 

However, PCV- 1849 is an air operated valve that has an actuator which 

would allow the valve to open on loss of air. During a Loss of Coolant 

Accident (LOCA) with a concurrent Loss of Instrument Air System pressure, 

PCV- 1849 may not have been capable of being closed, thus containment 

integrity could not be assured. This resulted in the determination that the 

actual design was outside the USAR. This is not consistent with the USAR 

Appendix G Criterion 53 and therefore, containment integrity could not be 

ensured during a LOCA with concurrent Loss of Offsite Power. 

Although the LER does not say it, this appears to be an original design 

defect.

Fort Calhoun 

Fort Calhoun

Instrument Air 

High Pressure 

Safety Injection 

Component 

Cooling Water

During a self-conducted Safety System Functional Inspection (SSFI) of C

valve operators that have installed accumulators, the capability of certain 

valves to perform their design function during a design basis accident with a 

concurrent loss of instrument air was questioned. The valves of concern are 

LCV-383-1&2, Safety Injection and Refueling Water Tank (SIRWT) 

isolation valves; HCV-438B & D, Component Cooling Water (CCW) to 

Reactor Coolant Pump (RCP) seal cooler isolation valves; HCV-238 & 239, 

charging pump header to Reactor Coolant System (RCS) isolation valves; 

and HCV-240 auxiliary pressurizer spray isolation valves. In general, the 

concern was that the accumulators for these AOVs, as originally designed 

and configured, did not have sufficient capacity to ensure that air pressure 

would be available for the time needed to complete the design basis 

functions. 

This is considered to be a common-cause failure condition. 61 0z 

ON CD 

Cn
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

28588028 881020 A check valve on the air accumulator providing back up air supply to steam 

isolation valve YCV-1045A was discovered inoperable during function 

testing. Immediate corrective action was required to repair the check valve. 

Upon disassembly, it was discovered that a metal particle, which had lodged 

in the valve seat, rendered the accumulator inoperable.

a Main Steam 

Isolation Valves

Fort Calhoun

Haddam Neck

28590025 900929

21393005 930518

Conditions were identified involving the Component Cooling Water (CCW), 

Raw Water (RW), and Containment Spray (CS) system which placed the 

unit outside its design basis for post-accident containment cooling. The 

CCW conditions involved the potential for degradation of containment air 

cooler performance and/or loss of CCW system operability following loss of 

Instrument Air. The CCW/RW interface valves would fail open on a loss of 

instrument air (IA) assumed as part of the post-LOCA accident scenario. The 

IA system is not a safety-related system; therefore its availability cannot be 

credited in a post-accident situation. Although these valves are equipped 

with backup air accumulators, the accumulators are not qualified as safety

grade items and also cannot be credited to operate. In addition, valves to 

many essential and nonessential heat exchangers served by CCW fail open 

upon a loss of instrument air. The resultant system flow distribution would 

deprive the containment air cooling coils of their design CCW flow by 

allowing CCW flow to nonessential equipment. This would result in heat 

removal performance by the containment air coolers below that assumed in 

the design basis. The primary cause is attributed to deficiencies in the 

original system design. 

This is considered to be a common-cause failure condition. This was an ASP 

Program Event. See Appendix B. 

A containment isolation valve for the reactor coolant letdown line 

(LD-TV-230) could not be closed from the control room using the control 

switch. LD-TV-230 is a normally open air operated valve located inside the 

containment. The problem was traced to a failure of the air pilot solenoid 

valve. The cause was corrosion inside the valve as the result of water 

intrusion from the containment control air system due to a malfunction of the 

containment control air dryer. 

Corrective actions included replacement of the failed solenoid valve, a 

blowdown of the air system to remove any remaining water, and repair of the 

dryer. It was unknown how long this valve was inoperable prior to 

discovery, but is believed to have been longer than the action time allowed 

by the plant Technical Specifications. The failed SOV was an ASCO Model 

(continued on next page)

C Component 

Cooling 

WaterRaw 

WaterContain

ment Spray 

System

a, b, c Instrument Air 

Reactor Coolant 

Letdown

Fort Calhoun

CD V)
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

Haddam Neck

Haddam Neck

Haddam Neck

21393005 

(continued)

930518

21393007 930525

21394005 940219

NP 8320A1 82V. The cause of the water in the air system was a malfunction 

of the containment control air dryer. The dryer is a Model 25HA4 

manufactured by Pall Pneumatic Products Corp. The malfunction involved a 

failure of the desiccant towers to switch from drying to regeneration mode 

properly. It was not known how long this condition may have existed. 

This is considered to be a common-cause failure condition. See 

LER 21393007. 

While performing a pressure decay test of the pressurizer Pilot-Operated 

Relief Valves (PORVs) emergency air supply system, it was determined that 

the pressure decay exceeded the Technical Specification acceptance criterion 

of 0.3 psi/hr. The problem was traced to a leak in the diaphragm assembly of 

one of the PORVs (PR-AOV-568). This leak was caused by both the 

inadequate sealing of the PORV diaphragm assembly and the failure of the 

PORV air supply pressure regulating valve (CA-PRV-836A).The failure of 

the air pressure regulating valve was caused by corrosion inside the valve 

due to water intrusion from the containment control air system. The moisture 

in the containment control air system was due to a malfunction of its air 

dryer. The diaphragm of the failed brass regulating valve was covered with a 

powdery blue-green corrosion product when it was inspected. This corrosion 

caused a leak in the diaphragm allowing air to vent through the regulator 

eventually causing the pressure to equalize on both sides. 

This is considered to be a common-cause failure condition. See 

LER 21393005.This was an ASP Program Event. See Appendix B. 

With the plant in Mode 5 (Cold Shutdown) while performing a stroke test of 

the pressurizer Pilot-Operated Relief Valves (PORVs), it was determined 

that the valves would not fully open. The problem was traced to a leak in the 

diaphragm assembly of the PORVs (PR-AOV-568 & 570). The leaking 

diaphragms were caused by loose diaphragm cover bolts. Both PORV 

diaphragms were replaced during the 1993 refueling outage with a new style. 

The principle change was the substitution of a longer lasting material 

(EPDM) for the old Buna-N material. The manufacturer also changed the 

shape of the diaphragm somewhat although this was never communicated to 

the licensee. This change resulted in some difficulty installing the 

diaphragm. To overcome this, a commonly used lubricant (Moly 55) was 

applied to aid installation. The PORVs were subsequently retested 

satisfactorily. After the February 1994 failures, an in-depth discussion with 

the manufacturer on the possible causes for failure revealed several aids to 

overcome installation problems. (continued on next page)

a, b, c Instrument Air 

Reactor Coolant 

Letdown

a, c

c

Instrument Air 

Pressurizer 

Pilot-Operated 

Relief Valves

Pressurizer 

Pilot-Operated 

Relief Valves

z 

C-) 

(A

,.-
Table 2. (continued).

LER Event

Pla N.m.er. Dateito

Classification2

Haddam Neck

Haddam Neck

21394005 

(continued)

940219

21396012 960611

The most significant was the use of a sealant around the diaphragm's bolt 

circle. It is believed that the presence of lubricant instead of the sealant 

allowed some extrusion of the diaphragm from between the base and cover 

and away from the bolt holes. This extrusion also led to small tears at several 

diaphragm bolt holes, allowing the bolts to loosen over time. Also, air 

regulators supplying both PORVs were set too low (77.3 PSIG and 75.1PSIG 

versus the required 85 PSIG). 

Program Event. See Appendix B. This event was reported as a common

mode failure. This was an ASP event. 

An engineering analysis determined that the main feedwater regulating 

valves (FRV) would not fully isolate feedwater flow as required for a main 

steam line break accident in containment. It was determined that the 

differential pressure across the valves would overcome the valve spring's 

closing forces. In a design basis steam line break analysis, the feedwater 

motor operated valves (MOV) are required to isolate; however, in the event 

of a single failure of the MOV the FRV is credited with isolation. The failure 

to isolate feedwater for a steam line break in containment could result in 

exceeding maximum containment design conditions. The cause of this 

condition was an erroneous assumption in the accident analysis that the 

FRVs would isolate against a high differential pressure. The differential 

pressure across the valve would overcome the valve spring's closing force. 

This is considered to be a common-cause failure condition. See 

LER 21396018. 

(Event No. 30619 dated 6/11/96 applies.)

c

C

Pressurizer 

Pilot-Operated 

Relief Valves

Main Feedwater

Plnnt

,-

Descrio~tion System
Table 2. (continuedV

LER Event 

Plant Numberi Date Description Classification2 System

Haddam Neck

Hatch I

21396018 960822

32192003 920102

An engineering analysis revealed that the feedwater regulating bypass valves 

would not fully isolate feedwater flow as required for a main steam line break 

inside containment. This condition was discovered during a follow-up to a 

similar problem with the main feedwater regulating valves (LER 21396012). 

The failure to isolate feedwater for a steam line break inside containment could 

result in exceeding maximum containment design conditions. This event did 

not involve any actual equipment failures. 

The cause of this condition was an erroneous assumption that the feedwater 

bypass valves would close and isolate against the differential pressure 

experienced between the steam generator feed pump and a faulted steam 

generator. Additionally, credit was taken for isolation of the bypass valves 

from the control room ten minutes after an accident, however the control air 

system which operates the valves is not a credited system. The bypass valves 

(FW-HICV-1301-1,2,3,4) are 1-1/2 inch, air to close, spring to open, 

manufactured by Masoneilan. They are normally closed during full power 

operation. On an auxiliary feedwater actuation signal they go full open until 

manual operator action is taken at the main control board to throttle flow. 

This is considered to be a common-cause failure condition. See 

LER 21396012. 

Valves 1P41-F039A and B, air operated cooling water supply valves to 

Emergency Equipment Room coolers 1T41-B002A and B, failed to open 

automatically as required during routine operability testing. These valves are 

designed to open automatically to provide cooling water to the room coolers to 

maintain the temperature below 148°F when the Core Spray and/or Residual 

Heat Removal pumps are in operation. With both the normal and standby 

coolers for this room inoperable, Core Spray pump I E2 I-COO IA and RHR 

pumps lEI 1-C002A and C were declared inoperable. An LCO was invoked 

and a temporary modification was implemented to place valves 1 P41 -F039A 

and B in the open position to ensure a supply of cooling water to the 

Emergency Equipment Room coolers. An unanticipated breakdown of a 

material used in the manufacturing process may have been a root cause. The 

solenoid operated valves (SOVs) in the air supply lines to valves I P41 -F039A 

and B failed to reposition as required. Consequently, valves 1P41 -F039A and 

B could not open. It appeared that the SOVs failed because their solenoid cores 

stuck to the top of the core housings, perhaps as a result of the gelling of a 

lubricant (Dow Coming 550) used in the assembly process. Corrective actions 

include replacing the SOVs, increasing the cycling of SOVs of this type, and 

changing these valves to another type of SOV. 

This is considered to be a common-cause failure condition.

C

b, c

Main Feedwater

RHR Core Spray

z 

tiI 

4x.

c"
Table 2. (continued).

LER Event

Descrintion

Z 

al

Classification2

b, c

System

10 Different 

Systems 

Instrument Air

35494017 941110

35497020 970828

Two room cooler isolation valves failed to open during a quarterly test of the 

Safety Auxiliary Cooling System. The valves are associated with the room 

coolers for two emergency diesel generators. These are Anchor-Darling flex

wedge gate AOVs with Hiller actuators. The most probable cause of failure 

was identified as binding associated with the stem packing and a 

nonconservative valve friction coefficient used in calculating the actuator 

size. (Lessons learned from the testing of motor-operated valves in 

accordance with the guidance in Generic Letter 89-10 led to this conclusion.) 

The licensee determined that this was a common-mode failure attributable to 

a design deficiency. 

A similar condition (referenced in the LER) was reported in Hope Creek 

LER 35493006 in October 1993 for these AOVs. In that event, excessive air 

pressure caused the gates to bind in the seat. This condition resulted when 

the packing style was changed which reduced the packing drag on the valve 

stem. The design change package did not account for the reduced packing 

drag and did not lower the air supply pressure to the actuator. The additional 

seating force and gate travel reduced the ability of the spring in the actuator 

to drive the valve open. 

Design deficiency (failure of instrument air) could have resulted in the loss 

of Emergency Diesel Generators and Emergency Core Cooling System. See 

Hope Creek Event #32836 in Table 4.

c

C

Diesel Generator 

Room Cooling

Safety Auxiliary 

Cooling System 

(SACS)

35486063 860828 Following an extensive investigation the determination was made that twelve 

air operated valves had solenoid valves installed which had an operating air 

supply pressure rating less than the maximum expected air supply header 

pressure. Ten of the valves in question had non-Q pressure regulators 

installed to limit the pressure to within the design value. Exceeding the 

pressure rating may result in failure of the solenoid valve to open. The root 

cause of this event was design inadequacy in that solenoid valves of 

incorrect pressure rating were specified for use. 

The subject valves were installed in the Containment Atmosphere Control 

(CAC) System and the Safety Auxiliaries Cooling System (SACS). 

This is considered to be a common-cause failure condition.

nt. Plan 

Hope Creelk

,.

Hope Creek

Hope Creek

Numberl Date Descrintion
Table 2. (continued).

LER Event 

Number' Date Description C Plant 

Indian Point 2

"lassification 2 System 

Main Steam 

Power Operated 

Relief Valves 

Auxiliary 

Feedwater 

System

Indian Point 3 28688009 881025

This is considered to be a common-cause failure condition. 

Automatic Switch Company (ASCO) indicating a potential problem existed 

with ASCO NP8314 series solenoid valves (SOVs) failing to shift to a de

energized position following extended periods of being energized. ASCO 

identified the failure mechanism as the solidification of a lubricant used in 

the initial manufacturing processes that had migrated into critical surfaces of 

the valve's subassembly and which subsequently caused the SOVs to fail to 

shift when called upon. The plant identified six normally energized NP8314 

series SOVs in use. Two of the normally energized NP8314 series SOVs 

were replaced with improved series NP8314 SOVs from ASCO. The two 

that were removed were disassembled and tested with no lubricant 

solidification problems noted. 

Refer to NUREG- 1275, Volume 6. ASCO employees used the lubricant, 

without authorization, to make assembly easier. 

This is considered to be a common-cause failure condition.

b, c Reactor Coolant 

System 

Liquid Waste

24793010 930818 It was determined by an engineering analysis that regulating valves installed c 

in the nitrogen backup to the Instrument Air System for the Main Steam 

Power Operated Relief Valves and the Auxiliary Feedwater System 

(AFWS), were not fully capable of performing their function in the manner 

set forth in the Final Safety Analysis Report (FSAR). 

Specifically, the nitrogen backup regulator is described in the FSAR as 

capable of automatically supplying nitrogen in the event of loss of 

instrument air. The nitrogen regulating valves were designed and installed 

with an extraneous equalizing line between the upper spring chamber vent 

and the downstream air line. This equalizing line compromised the ability of 

the nitrogen regulating valves to automatically modulate (or regulate) the 

nitrogen pressure and flow. The nitrogen regulating valves would therefore 

only pass nitrogen to the extent that it leaked past the valve seat or the valve 

was preset to a given opening. The nitrogen regulating valves could still have 

been operated manually. The valves are CASHCO, model HP 1-32-45-S36, 

0.5 inch size. Thus, the cause of the failure of the nitrogen backup regulator 

was an incorrect design application of the regulator by the licensee that was 

not caught by testing.

0 

0'
z 

70~ 

00•

Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System 

Indian Point 3 28693035 930916 An existing design flaw could allow an automatic Safety Injection actuation c WCCPPPS, IA 

signal combined with a loss of offsite power to result in the loss of the Weld and SA 

Channel and Containment Penetration Pressurization System (WCCPPS). 

The signal would cause the system's pressure control valves to fail in a 

closed position which cuts off the air supply. The design flaw is outside the 

design basis of the plant and had existed since initial plant startup (August 

1976). 

The IA system supplies compressed air at approximately 100 psig through a 

check valve to the WCCPPS header which supplies the four WCCPPS zones. 

Each zone has an air receiver upstream of a pressure control valve (PCV) 

that is sized to maintain the air supply for at least four hours in the event the 

IA and SA are lost. The PCVs are designed as air-to-open valves, meaning 

that air pressure must be supplied to the dome of the PCV in order for the 

valve to open and maintain the downstream air pressure at about 46 psig. 

The PCVs rely upon the pressure in the IA system to remain open since there 

are check valves between the WCCPPS air receivers and the connections to 

IA. An automatic SI System actuation signal with a loss of offsite could have 

caused a loss of IA because the safety related electrical busses are stripped 

and the IA and SA compressors (CMP) are not reloaded. Following a loss of 

the compressors, the IA system would retain sufficient pressure to operate 

the PCVs for about 1/2 hour. When the air pressure is insufficient to hold 

open the PCVs, they fail closed and all air flow to the WCCPPS stops. 

Manual action to regulate WCCPPS air pressure using the PCV bypass lines 

is feasible (each zone has a pressure relief valve set at 49 psig to prevent 

overpressurization), but is not considered in the system's design basis. The 

backup supplies would not prevent this event. The primary source of backup 

is the nitrogen supply which has sufficient capacity for 24 hours. When the 

pressure downstream of any PCV drops to 43 psig, the zone's pressure 

controller (PC) actuates a nitrogen supply valve to deliver nitrogen to a 

connection just upstream of the PCVs. This gas supply will not operate the 

PCVs because of intervening check valves. The secondary source of backup 

air is the plant's SA system which is cross-connected to the IA system. SA 

would also be lost when a. automatic SI actuation signal occurs because the 

compressors are stripped from the safety related busses and not reloaded. 

The last backup is a permanent cross connection to the Con Ed IP2 SA 

system. Operation of this backup requires manual action not considered in 

the system's design basis. 

This is considered to be a common cause failure condition.

H 

0

cjTable 2. (continued).

Event 

n"at f T3e.•crintion Classification2 Plant 

Indian Point 3 b, c

System

Service Water 

System

28693053 931202 With the plant in a cold shutdown condition, flow control valves 

SWN-FCV- 1176 and SWN-FCV-1 176A were found to be inoperable during 

a post-work test. Operations declared all three emergency diesel generators 

inoperable because at least one valve must be operable to allow service water 

to the generators. 

The solenoids on AOVs used to control the EDG service water effluent for 

all three EDGs (two parallel valves in a common header line) were replaced 

and then tested as part of a normal work package on a safety-related system. 

The solenoids were incorrectly connected and the valves would, therefore, 

not stroke open. A series of miscommunications and mistakes (some caused 

by inadequate procedures) led to the condition where all three emergency 

diesel generators were inoperable for about 4.5 hours. 

Although the LER focuses on the mistakes and procedural difficulties, this 

event illustrates the (unintended) interdependence of critical systems on air

operated valves which are common to parallel trains of safety-related 

equipment. This is considered to be a common-cause failure condition. 

See Indian Point 3, Event No. 26449 in Table 4.

28693050

LER 

lX-T t

Indian Point 3 b, c Emergency 

Diesel 

Generators

z 

4zx

,

o"~

NUIIIUI ý1L .. .. C . .

931115 Operations issued a deviation event report to identify the potential 

inoperability of four Service Water System solenoid operated valves 

affecting emergency diesel generator operation. The condition resulted from 

solenoid operated valves rated at less than the instrument air pressure supply 

and a failure mode that could cause unregulated instrument air. 

Subsequently, 109 additional valves subject to overpressure and one failure 

of a valve to vent were identified. The condition resulted from an original 

design deficiency and a subsequent failure to correct the error in response to 

regulatory feedback. 

The system was designed so that the air operated valves fail safe when the 

instrument air supply is lost. The failure mode of the SOVs due to 

overpressure from air regulator failure or high setpoint and the effect on the 

AOVs was not recognized. 

The licensee did not evaluate GL 91-15 when received. They did not request 

an adequate review of GL 91-15 because the GL did not require a response 

to the NRC. There were no procedural requirements in place to request such 

a review. 

This is considered to be a common-cause failure condition.
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

al Indian Point 3

(TI 

4UC

0-.i

Containment 

Isolation 

Diaphragm 

Valves

Indian Point 3 28696008 960320 The System Engineer for the Isolation Valve Seal Water System (IVSWS) 

determined that the minimum IVSWS nitrogen tank pressure required to 

support proper system operation was in question. Deviation Event Report 96

824 was written to document this determination. A subsequent evaluation 

determined that the nitrogen supply (three bottles on a header) had been less 

than that required for the IVSWS to operate as designed. This placed the 

plant in a condition prohibited by Technical Specifications. The cause of the 

event was the lack of original design basis documentation to identify the 

requirements for the nitrogen tanks.

c Isolation Valve 

Seal Water 

System

28696004 960215 Two air operated vapor containment isolation diaphragm valves in series 

were found to be inoperable which violated Technical Specification 3.6.A. 1. 

The original isolation valve design would close at system pressure when 

there was a differential pressure or accident design pressure, but not against 

system pressure with no pressure differential. The design and specification of 

the valves during initial design and construction was not adequate to ensure 

the valves met their design requirement for containment isolation. 

RC-AOV-519 and RC-AOV-552 (the valves are designed to positively seal 

with a differential pressure) would close against a differential pressure of 

150 psi (the Primary Water System design pressure) but would not have 

closed with a line pressure greater than about 120 psig when there is no 

differential pressure. The original specification for these valves was for a 

maximum differential pressure of 200 psi, with no reference to a minimum 

pressure differential or constant line pressure requirement. The valve design 

had not been modified since the original plant design and construction. 

The condition could have existed when closing isolation valve 

RC-AOV-560, downstream and in series with RC-AOV-519 and 552, in a 

sequence that maintained a line pressure greater than 12 psi. There would 

then be no differential pressure during stroke testing. Stroke testing in these 

conditions created another problem if the limit switches are adjusted after the 

valves are shut but do not fully close. The valves were assumed to be fully 

closed during the adjustment so subsequent failures to close would be 

masked and could allow a control room indication that the valves were fully 

closed when they were not. There was no direct external indication on the 

valves to show if they were fully closed. 

This is considered to be a common-cause failure condition.

0
Table 2. (continued).

LER Event

Plnr Descriotion Classification2

37396011 960928 While developing an Air operated Valves (AOV) preventative maintenance 

program, inconsistent testing data were obtained for valves with 

WKM 70-13-1 pneumatic actuators. The inconsistent results appeared to be 

related to incorrect effective diaphragm areas (EDA) for the AOV actuators. 

36 WKM AOVs were addressed in the LER (18 per unit). 13 AOVs in each 

unit are part of the primary containment isolation system (PCIS) and 5 are in 

the reactor core isolation cooling system (RCIC). Two problems associated 

with the EDA of the actuators of the WKM valves were identified. The first 

was related to the actual versus the manufacturer's published EDA of the 

actuator. If the actual EDA is less than what the manufacturer publishes, then 

the closing (or opening) forces installed in the valve (via spring/spring 

adjustment) will be less than required. The second problem was stretching of 

the diaphragm during valve travel resulting in a reduced EDA. 

This is considered to be a common-cause failure condition.

c Containment 

Isolation System 

Reactor Core 

Isolation Cooling

37495005 950218 The hand switch for the "D" Outboard Main Steam Isolation Valve (MSIV), 

2B21-F028D, was placed in the CLOSED position and the valve failed to 

close. The hand switch was cycled several times from AUTO to CLOSED to 

AUTO with the same results. The valve was eventually closed by placing the 

hand switch in the OPEN SLOW TEST position and depressing the Slow 

Test pushbutton. This slow-closed the valve, and the pilot air was isolated. 

The cause of the MSIVs failure to close was sticking of the solenoid pilot 

valve. This sticking prevented the pilot valve from changing state, and did 

not allow the pilot air to vent. The failure of the pilot air to vent resulted in 

air ported to the under side of the MSIV operating piston, holding the MSIV 

in the open position. 

The sticking solenoid pilot valves were disassembled and inspected. Foreign 

material was observed on several internal parts of the solenoid pilot valve. 

The interfacing surface of the core assembly and plug nut of the "B" 

solenoid appeared to have a thin coat of foreign material. When the core 

assembly and plug nut were pressed together, which is the normal 

configuration when the solenoid is energized, the film acted as an adhesive. 

This adhesive was strong enough that when the plug nut was lifted from the 

bench, the core assembly adhered to it and was also lifted. The foreign 

material was determined to be Nyogel 775A. 

According to ASCO, this material is used during assembly as a thread 

lubricant on the solenoid cover and at the interface of the solenoid base and 

housing assembly. This is considered to be a common-cause failure.

b, c

LaSalle I

System

LaSalle 2 Main Steam 

Isolation Valves

z: 

ON

CD 

V)r

Number' Date Descrintion L
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification 2 System

Maine Yankee

33689011 890906

33697011 970402

z 

(7 

•0 

6N

c 

b, c, 

Personnel Error

30996003 960213 Maine Yankee was operating at 2440 MWt (90.3% power) when the reactor Main Feedwater

automatically scrammed due a loss of load trip from a high steam generator 

#3 water level. The high water level was due to a faulty positioner on the 

steam generator #3 Main Feedwater Regulating Valve (MFRV). The faulty 

positioner caused the main feedwater regulating valve to go to the full open 

position which resulted in overfeeding the #3 steam generator and the 

resultant trip on high water level. The faulty positioner on #3 MFRV has 

been replaced. The positioner on #2 MFRV was also replaced as a 

precautionary measure, since it had been in service since July 1992. The 

positioner on #1 MFRV had been replaced in January 1996. 

An instrument air check valve in the air supply to the service water supply 

isolation to one of two Turbine Building Closed Cooling Water headers, was 

found incorrectly located. The location of the check valve would have 

caused the accumulator air supply to bleed down on a loss of the normal 

instrument air supply. 

The root cause of the condition was the incorrect re-assembly of the 

instrument air lines to the solenoid valves and the accumulator. It is assumed 

that the check valve has remained incorrectly located since the disassembly 

and reassembly of this air line for service water piping replacement. 

The closing force for multiple dual function (two separate pressure isolation 

functions) valves had been improperly set, resulting in the valves being 

incapable of closing to a leak tight condition against normal operating 

system pressure (NOSP). Eleven of the 23 valves tested were not capable of 

providing an adequate closing force. This deficiency could have resulted in 

the potential for a release of radioactive materials to the Auxiliary Building 

greater than analyzed in the facility Final Safety Analysis Report (FSAR). 

The closing forces were incorrectly set during the period between October 

1986 and March 1997.

The cause of this event was an insufficient program to ensure that facility 

procedures clearly addressed all related design basis functions. The affected 

valves were to have been adjusted to ensure they properly close against 

containment design pressure and NOSP. The appropriate procedures were to 

have been revised to ensure that proper valve control parameters are 

specified and verified after any maintenance activities are performed that 

could affect dual function valve closing forces. 

This condition affected 11 valves in three systems. This is considered to be a 

common-cause failure condition. 

Event No. 32070 referred to this event.

c

Turbine Building 

Closed Cooling 

Water 

Instrument Air

Containment, 

CVCS, RC 

Sample System 

Liquid Radwaste 

System

CD 

V)..-

Millstone 2

Millstone 2
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

Millstone 3 42396013 960515

42396028 960916

An engineering evaluation determined that a design deficiency in the 

Residual Heat Removal System (RHS) was a condition that was outside the 

design basis of the plant. A loss of control air supplied from the non-safety

related instrument air system could cause the RHS control valves to fail 

open. If this condition occurred during the initial phase of a plant cool down, 

the Reactor Plant Component Cooling Water System (CCP) temperatures 

could exceed 125°F. This is the design temperature used in the system stress 

analysis. If RHS heat exchanger operation was initiated at 350°F RCS 

temperature, as assumed, then the RHS heat exchanger CCP outlet 

temperature could be as high as 250'F if the valves failed open. Under the 

resultant conditions the CCP piping would not meet the ASME Section III, 

Appendix F stress criteria.

The original plant design did not consider that if the RHS flow control valves 

failed open on a loss of air, it could create unacceptably high RHS heat 

exchanger CCP discharge temperatures. 

This is considered to be a common-cause failure condition. 

An engineering evaluation identified a failure scenario in which a loss of 

Instrument Air (IAS) to temperature control valves in the Charging Pump 

Cooling (CCE) system serving the charging pump lube oil coolers, 

coincident with 33°F Service Water (SWP) temperature could result in 

overcooling of both trains of the charging pump lube oil system and 

challenge charging pump operability. Failure of the air-operated CCE valves 

to the full open position due to a loss of the non-safety related IAS system 

would adversely affect both trains of the charging pumps by allowing 

excessive cooling of the CCE system which cools the lube oil system. This 

condition alone could have prevented the fulfillment of the safety function of 

the system. The cause of the charging pump inoperability was inadequate 

original design. This condition would result from overcooling of the lube oil 

system from a failure of the non-safety related Instrument Air system 

coincident with a worst case minimum SWP temperature and maximum flow 

and heat exchanger cleanliness. Under these conditions, the air-operated 

CCE valves would fail open and excessive cooling of the lube oil system 

would occur. This particular combination of conditions was not considered 

in the initial design. A short term corrective action was committed to in the 

LER to install a temporary modification to limit the failure position of the 

three way CCE temperature control valve to ensure sufficient bypass flow 

(continued on next page)

c Residual Heat Removal System

c Charging System

.-

Millstone 3

0 

ON 

LA 

.P.
Table 2. (continued).

LER Event 

Number' Date

42396028 

(continued)

960916

z 

0

Description Classification

2 Plant 

Millstone 3 

Millstone 3

c

42396031 960906 b, c

around the SW heat exchanger to maintain CCE temperature above 45'F. 

Other long-term actions were also discussed. This is considered to be a 

common-cause failure condition. 

A design engineering review concluded that specific safety related control 

valves could fail due to exceeding the manufacture's maximum operating 

pressure differential pressure rating of the Solenoid Operated Valves (SOVs) 

installed on the control valves. SOVs could fail to perform safety related 

functions because of excessive operating pressure differentials. This can 

result from failures of non-qualified air regulators installed in the Instrument 

Air (IA) system upstream of the SOVs. The failure of an air regulator would, 

in turn, result in full IA system pressure being applied to the SOV. The SOV 

can potentially fail to operate properly since they are not rated for full IA 

system pressure. A total of 48 SOVs which perform safety-related functions 

had been originally identified in various systems which would be susceptible 

to such a failure. The cause of this condition was the failure to consider the 

potential for pressure regulator failure in the original design and selection of 

SOVs. 

This is considered to be a common-cause failure condition. 

The valves were in the following systems: CVCS-5 valves; HVG - 4 

valves; HPSI-3 valves; LPSI-2 valves; AFW-2 valves; Reactor Gas 

Drains-2 valves; Aux. Steam-2 valves; Chg. Pump Cooling-2 valves; 

Nitrogen Sys.-1 valve; Reactor Plant Vent-] 2 valves; Main Steam-3 

valves; Primary Grade Water-I valve; Turbine Drains-8 valves; Quench 

Spray System-I valve.

System

Charging System

13 Different 

Systems
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System 

Millstone 3 42396036 960926 The High Pressure Safety Injection (SIH) and Low Pressure Safety Injection c High and Low

42396040 961024

(SIL) systems would have been subject to degraded performance due to 

possible mispositioning of normally closed safety related air operated valves. 

Mispositioning of these 21 valves can be postulated to occur under post

accident harsh environmental conditions due to failure of non-qualified 

power and control circuits. As a result the potential diversion of SIH and/or 

SIL flow under accident conditions could have been more than the margin 

allowed within the Loss of Coolant Accident analysis. 

Seventeen additional safety-related air operated and solenoid operated valves 

were subsequently identified where failures of non-qualified control circuits 

could degrade performance of a safety system function. These additional 

valves are located in the following systems: Reactor Plant Component 

Cooling Water (CCP), Containment Vacuum, Reactor Plant Sampling (SSR), 

Post Accident Sampling (SSP), and Main Steam to the auxiliary feedwater 

steam turbine. 

The cause of the reported conditions was a design mistake. The initial plant 

design did not adequately consider the potential mispositioning of these 

valves under harsh environmental conditions or active failure. 

This is considered to be a common-cause failure condition. 

An engineering evaluation determined that a failure scenario for the Reactor 

Plant Component Cooling Water (CCP) system had the potential for a loss of 

system safety function. The failure scenario involves a loss of the non

category 1 Instrument Air System, which would allow CCP valves to 

reposition to a maximum cooling configuration. Coupled with a low heat 

load and minimum Service Water (SWP) inlet temperature, the CCP system 

could reach temperatures lower than values for which they are analyzed, 

thereby rendering the CCP system, and systems that it serves, potentially 

inoperable. This was.caused by improper initial design of the CCP system. 

The failure mode described in the LER was a design oversight by the plant's 

architect engineer that occurred during the original design process. The 

design basis analysis for the CCP system focused on high CCP heat load 

conditions. The designers did not analyze for extremely low CCP heat loads 

concurrent with very low SWP temperatures. 

This is considered to be a common-cause failure condition.

c

Pressure Safety Injection

Reactor Plant Component 

Cooling Water

CD 

V)

Millstone 3

z 

0\ 

C-r1 

ýO 

4ab
Table 2. (continued).

Plant

Nine Mile Point I

LER Event 

Number) Date Description Classification

2

z 

ni 

ON 

ON

c

System

Main Feedwater

27093002 930610 An evaluation of air leakage from backup accumulators on AOVs associated 

with containment integrity revealed that containment integrity could be lost 

during a postulated Loss of Instrument Air coincident with a Small Break 

Loss of Coolant Accident. Seal Return Block valves, which are air operated 

valves outside containment whose purpose is containment isolation for the 

RCP Seal Return lines, may return to the open position due to accumulator 

leakage. Due to leakage in the backup accumulator system associated with 

the valves, the accumulator may not have maintained them closed during a 

postulated event. 

In response to Generic Letter 88-14 and other industry guidance, selective air 

operated valves were tested. During the preparation of the initial test 

procedures emphasis was placed on ensuring that these valves failed to their 

required position on a loss of instrument air. It was also recognized that 

leakage of the accumulator system could result in these valves opening; 

however, the appropriate acceptance criterion was not established at the 

time.

c

22096004 960520 Feedwater flow control valve oscillations, caused high reactor water level, 

which tripped the turbine and resulted in a reactor scram. The cause of the 

event was a degraded actuator for the #13 feedwater flow control valve 

(FCV). Upon disassembly of the actuator, worn o-rings and bushings as well 

as a misaligned stem and bushing were found. 

Two contributing factors were identified. The first was that the design of the 

current configuration of the pneumatic controls left little margin for 

degradation of the actuator and caused the valve to cycle frequently. This 

cycling may have caused increased wear on the actuator bushings and rubber 

goods which could lead to volume booster needle valve clogging and cause 

instability in the control system. The second is the misalignment of the 

actuator stem. This misalignment caused binding which could also have 

caused erratic operation. The probable cause of the misalignment was 

incorrect factory assembly of the actuator, resulting in a misalignment of the 

internal bushings. 

The control system design was to have been evaluated to determine if 

changes were needed to increase reliability. Potential recommended actions 

included modifications to increase pneumatic volume of sub-components to 

decrease the gain of the valve controls and limit cycling of the feedwater 

control valve. This would allow the booster to be less susceptible to wear 

material and reduce the number of service cycles on the boosters.

tO

Oconee 2 RCP Seal Return
I dUIII . opuIILIIIUvU). 

LER Event 

Plant Number' Date Description Classification2 System 

Oconee 3 28791007 910703 A reactor trip occurred from 100% power on a loss of main feedwater. Five a, b Main Feedwater

Oyster Creek I 21985012 850612

condensate (in parallel) demineralizer valves failed closed when the master 

controller for these valves failed. The master controller failed when particles 

from a degraded seal clogged the Instrument Air flow path. Demineralizer 

bypass valves could not open to compensate because an Operator had failed 

to return them to automatic control. This is considered to be a common-cause 

failure condition. 

The loss of condensate flow resulted in the trip of condensate booster pumps 

due to low suction pressure, which then caused a trip of the main feedwater 

pumps, followed by the reactor trip. 

Emergency feedwater pumps started, but a solenoid valve failed, requiring 

manual operator action to control flow to one steam generator. The SOV was 

a Valcor V-70900-21-3. This model SOV was used in several other 

applications at the plant and had failed to operate properly in the past. 

Because it is normally energized and operates at an elevated temperature 

(approximately 250'F), the licensee believed that the temperature caused 

degradation leading to the valve sticking open when de-energized. 

This was an ASP Program Event. See Appendix B. 

As the result of a reactor scram all rods inserted but one of the two Scram 

Discharge Volumes did not fully isolate. The resulting flow of hot water 

through the Scram Discharge Volume caused steam and paint fumes to 

discharge in the reactor building. This, in turn, activated the deluge fire 

system on one level of the reactor building. 

One SOV drain valve that isolates the Scram Discharge Volume bottomed 

out before the valve was fully seated because the stroke adjustment was 

improperly set. The actuator spring on a second drain valve was improperly 

sized and opened slightly when pressure from the first valve was applied to 

its seat. With both valves slightly open, the leak was established. 

Failure of the SDV to fully isolate would be most severe in the event of a 

loss of offsite power, which would cause an immediate reactor scram. Feed 

pumps would not be available and the leak through the SDV would lower the 

reactor level until Core Spray would be initiated. Operator action to reduce 

reactor pressure by manual action would be necessary. 

This was an ASP Program Event. See Appendix B.

Condensate 

System 

Instrument Air 

Scram Discharge 

Volume c

-.-

C)
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

z 

ON 

tOA 

-01

a 

a 

a

HShutdown 

Cooling (RHR) 

Shutdown 

Cooling (RHR) 

Main Feedwater 

Instrument Air

25592007 920205 The main steam isolation valve (MSIV) actuator solenoid valves, could have 

been rendered inoperable by a main steam line break outside of containment 

because the solenoids were all served by the same power supply. 

The original plant design placed the MSIV actuator solenoid valves in the 

same room as the MSIVs. During the 1974 refueling outage a second 

redundant set of MSIV actuator solenoid valves were installed in the turbine 

building (a non-harsh environment). However, the redundant solenoid valves 

located in the non-harsh environment received power from the same 

electrical circuit as the valves located in the harsh environment. Following a 

main steam line break outside of containment, a short of the solenoid valves 

in the harsh environment could short the fuses in the control scheme. These 

same fuses were in the circuit for the redundant solenoid valves in the 

non-harsh environment. Thus, a failure of the component in the harsh 

environment could also have prevented the redundant component from 

performing its safety function. (continued on next page)

b, c

Palisades 25578003 780108 With the reactor in cold shutdown, CV-3025 (shutdown cooling system HX 

outlet valve) failed closed, causing a loss of shutdown cooling flow and 

allowing PCS temperature to rise to approximately 215°F. Condition was 

attributed to moisture contamination of the air supplied to the AOV. This 

event is discussed in Section 8.3.2 of this study. 

25581030 810718 CV-3025 failed closed and isolated shutdown cooling system. Condition was 

again attributed to moisture contamination of the air supplied to the 

AOV.This event is discussed in Section 8.3.2 of this study. 

25587018 870620 During a power reduction, the "A" train main feedwater regulating valve, 

failed to close. Shortly after, when the turbine generator was removed from 

service, a moisture separator and reheater control valve failed to close. 

The cause of the failure was attributed to a plugged "close" port in the valve 

positioner. The particle that had plugged the port was not recovered; 

however, it was believed to have originated from the upstream carbon steel 

manifold or materials still present from post air system problems. The LER 

did not indicate the nature of the "air system problems." The failure of the 

moisture separator reheater control valves to trip closed was attributed to set 

point drift or in an improper setting within the temperature governed control 

circuit of the auto control feature. This auto control feature provides for slow 

closure of the valve for high temperature protection of the low pressure 

turbine.

Palisades 

Palisades

00

Palisades Main Steam 

Isolation Valve
Table 2. (continued).

LER 

' r_ _ L I Event

flp•co, rnt inn INUmDer Date F

25592007 

(continued)

920205

25592023 920327

25594004 940209

Plant 

Palisades 

Palisades

Classification2

b, c

b, c

System

Main Steam 

Isolation Valve

High Pressure Safety Injection

The main steam line isolation valve actuation solenoid valves were to have 

been relocated to a non-harsh environment so that a main steam line break 

outside of containment would not render the MSIVs closing solenoid valves 

incapable of performing their design basis function. This modification was to 

have been completed prior to startup from the then current refueling outage. 

This is considered to be a common-cause failure condition. 

It was discovered that previously unidentified junction boxes in the electrical 

circuits for solenoid valves SV-3084 and SV-3085 and position switches 

POS-3084 and POS-3085 which control the hot leg injection letdown valves 

CV-3084 and CV-3085 contained environmentally unqualified electrical 

connections. (Note that the position switches were not the correct type.) 

Further, the Air Circuit Assembly (including junction boxes) was found to be 

not environmentally qualified. 

The causes of this event included: 

1. The failure of the plant designers to ensure that the procurement and 

installation of the Air Circuit Assembly met the requirements of 

IOCFR50.49 

2. The lack of proper administrative controls lbr procurement and 

documentation of environmentally quali fied equipment. 

A previously unidentified single failure mechanism was discovered, 

involving the dependence on non-safety-related instrument air to provide 

motive power to safety-related AOVs. This condition could have affected the 

operability of the Engineered Safeguards System (ESS) pumps. The 

Palisades Engineered Safeguards Systems (ESS) pumps are equipped with 

seal and bearing cooling. Pump cooling is required for the long term 

operability of the pumps. 

ESS pump cooling is provided by: diverting a small portion of each pump's 

discharge flow through a heat exchanger, where it is cooled by Component 

Cooling Water (CCW); directing it through cooling jackets around the 

pump's seal areas; and then returning it to the pump suction. CCW is also 

directed to cooling jackets around the pump bearing areas. The CCW flow 

for all pump cooling is controlled by common supply and return AOVs, 

CV-0913 and CV-0950. (The Palisades CCW system has redundant trains of 

pumps and pump controls, but uses a single, common, piping system.) These 

two AOVs are opened automatically on a Safety Injection Signal (SIS). The 

normal configuration for the CCW to the pump seal cooling had been for 

(continued on next page)

C Engineered Safeguards 

Systems (HPSI, 

LPSI, CS CCW)

I'.)

Palisades

z 

C) 

ON 

(-A 

41,
Table 2. (continued). 

LER Event 

Plant Number] Date Description Classification2 System

z 

C

C

C. 

Wr

Engineered 

Safeguards 

Systems (HPSI, 

LPSI, CS CCW)

Palisades 25594004 

(continued)

940209 both CV-0913 and CV-0950 to be closed. Both of these AOVs are spring 

loaded to the open position, requiring availability of both instrument air and 

electrical power to hold them closed. Other AOVs, which supply service 

water (SWS) as backup cooling, are also normally closed, but are spring 

loaded to the closed position, thus requiring availability of both instrument 

air and electrical power to open them. If either CV-0913 or CV-0950 failed 

to open on demand following a Loss Of Coolant Accident (LOCA) (possibly 

due to binding of the valve stem or similar failure), cooling flow to the ESS 

pumps would not be available from the CCW system. Conceptually, 

recovery from such a failure would simply require the operators to initiate 

the alternate cooling by closing whichever CCW valve remained open and 

opening the SWS valves. However, instrument air is not a safety grade 

system and therefore might not be available to close either CCW valve or 

open the SWS valves. The use of non-safety grade instrument air to open the 

backup (SWS) cooling AOVs implies that credit should not be taken in the 

design basis for operation of backup cooling. Therefore, if either CV-0913 or 

CV-0950 should fail to open on demand, all (safety grade) cooling to the 

ESS pumps, and consequently the functioning of the pumps themselves, 

could eventually be lost. These conditions lead to the possibility of a single 

active (common-cause) failure disabling the ESS pumps. Failure of the ESS 

pumps would be a matter of considerable safety significance. The LER did 

not explain if or how the SWS backup function would still be an operational 

option. 

Palisades decided that operating with the CCW valves normally open meets 

plant design and licensing requirements and reduced the overall plant risk 

compared with having the CCW valves normally closed. Accordingly, the 

CCW valves were aligned so that their normal position is to be open. This is 

considered to be a common-cause failure condition.
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

Palisades C HP Air System, 

ECCS

52889005 890412 Atmospheric Dump Valves at PVNGS would not perform their intended 

safety function because the bonnet pressure was too high. In addition, due to 

the design of the valve/operator, these AOVs were subject to oscillations that 

could damage the valve. These conditions are also described in a PVNGS 

ATMOSPHERIC DUMP VALVE ANALYSIS REPORT, dated 

March/April 1989. 

(LER 53089001 described the original performance problems of ADVs 

during a transient event at Palo Verde 3 that led to the investigation of the 

ADVs in all three units. This was an ASP Program Event. See Appendix B.) 

This is considered to be a common-cause failure condition.

25598006 980305 Event No. 33843 for Palisades in Table 4 was originally recorded for this 

event. The LER was subsequently issued and is discussed here. 

The original design bases, and subsequent design reviews, for the HP Air 

System address the Large Break Loss of Coolant Accident (LBLOCA), yet 

did not address, in detail, the SBLOCA. As a result, the impact of a loss of 

the HP Air compressors on the ability to supply air needed to align valves for 

sump recirculation during a SBLOCA, and the need to incorporate manual 

operator actions in EOPs to assure HP air reliability were not considered. 

As a result of the design basis reconstitution for the HP Air System, the 

licensee identified the need for procedural guidance to direct manual 

operator actions to restore the HP Air System during a SBLOCA. The HP 

Air System supports ECCS realignment to the recirculation mode by 

providing air required to operate control valves needed to switch the suction 

of the ECCS pumps from the Safety Injection Refueling Water Tank 

(SIRWT) to the containment sump. During a LOCA with LOOP, the HP Air 

System compressors are load shed from safety-related electrical buses and 

must be manually repowered at their respective motor control centers 

(MCCs). 

When the air compressors are without power, the receiver tanks are not being 

charged. During this time, the HP Air System pressure experiences a gradual 

decay due to air bleed-off from pressure regulators and air leakage through 

seals in piston-driven control valve actuators. 

The LER focuses on the procedures needed to ensure restoration of HP air; 

however, the effects of gradual loss of air on the failed position on valves 

that are supplied by HP air had not been investigated.

Palo Verde I

z 0 

rri 

U'

C Containment 

Main Steam

,-. 

07
Table 2. (continued).

LER Event 

Number' Date Description

z 

7O 

(61

Classification2

C 

c

Peach Bottom 3 27891017 910924 During routine preventative maintenance on the Main Steam Relief Valve 

(MSRV) solenoid valves (SOV) the MSRV SOV wiring insulation was 

discovered to be degraded. Improper installation of the MSRV thermal 

insulation caused a high temperature environment around the solenoid valves 

and associated wiring. 

The MSRV thermal insulation had been improperly installed during the 

Refueling Outage in 1989. This resulted in an unusually high temperature 

environment in the immediate vicinity of the SOVs and associated wiring. 

This high temperature condition caused the MSRV SOV wiring insulation to 

degrade. The temperature increase from the insulation installation error 

caused the expiration of the Environmental Qualification (EQ) life of the 

components after approximately three days of operation. The ADS MSRVs 

comprise 5 of the 11 MSRVs and three of the ADS MSRVs were affected by 

this condition. 

This is considered to be a common-cause failure condition. This was an ASP 

Program Event. See Appendix B.

Plant 

Palo Verde I 

Palo Verde I

52894009 941118 Seal leakage was reported on one of the Fisher air-operated 

letdown/containment isolation valves reported to be undersized in 

LER 52895007. 

52895007 950512 Three Fisher air-operated letdown/containment isolation valves were found 

to be undersized and the bench settings were also set too low. These valves 

were provided by the system vendor (ABB-CE) and the design basis 

evaluation was found to be at fault. Modifications included spring 

replacement, reducing the stroke length of the actuator, and modifying the 

limit switches. Similar conditions were found at Units 2 and 3. 

The system vendor (ABB-CE) procured valves that had undersized air 

actuators and bench sets which were too low to provide desired valve seating 

force for the differential pressure which would be present across the valves 

during a letdown line break. They could not determine whether or not the 

procurement of the valves was approved by the Architect Engineer or the 

owner. The root cause of the design deficiency was attributed to the absence 

of a detailed design basis evaluation for air operated valves as part of an 

AOV program. This is considered to be a common-cause failure condition.

System

CD 

V)

Containment 

Containment

Personnel 

Error, b

Main Steam 

Relief Valve

Table 2. (continued).
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System 

Perry 44087009 870227 Failure of two control air solenoid valves rendered both emergency diesel b, c Emergency 

generators inoperable. The valves were subjected to temperatures near the Diesel 

upper end of the qualified operating range and were continuously energized. Generators 

These factors are believed to have eventually caused degradation of the valve 

poppets and failure of the valves. The valve manufacturer was contacted and 

responded that failures seldom occurred and that the most common failure 

developed when the poppets, made from Buna-N, were subjected to an 

incompatible lubricant or excessive heat. These valves are subjected to 

temperatures near the upper end of the qualified operating range and are 

continuously energized. These conditions were believed to eventually cause 

degradation of the material within the valve resulting in air leakage and 

eventual failure. Both leaking solenoid valves had previously been identified 

for replacement due to leakage, with work requests initiated but not yet 

implemented. In addition, several Surveillance Tests had been run 

successfully subsequent to identifying the leaking valves. The cause of the 

immediate diesel failures was the accelerated degraded condition of the 

. leaking solenoid valves. At discovery, the conditions were evaluated not to 

require immediate action, thus expeditious replacement did not occur. 

Subsequently, it was determined that the control air pressure regulators 

which supply air to these solenoid valves were malfunctioning and may have 

contributed to the diesel failure. This is considered to be a common-cause 

failure condition. This was an ASP Program Event. See Appendix B. 

Perry 44090021 900709 During cooldown of Unit 1, in preparation for the second refueling outage, b, c Main Steam 

directions were given to close the main steam isolation valves (MSIV's) in 

order to maintain control of the reactor cooldown rate. All eight of the 

MSIV's slow closed properly; however, two of them (I B21 -F022C and 

-F028B) failed to remain closed when their control switches were placed in 

the 'close' position. These valves later closed on their own, with the control 

switch left in the 'close' position. 

Centerior energy letter PY-CEI/OIE-0327 L to the NRC, dated September 

11, 1990, documented this event. Similar events occurred on October 29, 

Z 1987, November 3, 1987, and November 29, 1987. These earlier events were 

caused by degraded ethylene propylene diene monomer (EPDM) elastomers 

in the ASCO 3-way dual coil solenoid valves (October 29, 1987 and 

November 3, 1987), and by a sliver of EPDM inside a solenoid causing it to 

stick (November 29, 1987). The elastomers degraded because of locally high 

temperatures resulting from steam leaks. The EPDM elastomers were 

replaced with Viton in a complete solenoid valve changeout during 

(continued on next page)
LER Event Ha

Plant Number' Date Description Classification2 System 

Perry 44090021 900709 Refueling Outage 01. The cause of the September 7, 1990 ,events was b, c Main Steam

determined to be failure of the ASCO 3-way dual coil solenoid valves to 

change position after being de-energized. Specifically, the disc holder seat 

elastomer failed to shift from the exhaust port.

z 

0 

8N 

ON 

(Ji

C Containment 

Isolation System

29389004 890127 An Unusual Event was declared and a shutdown began because two primary c 

containment air operated valves were inoperable. The air operated valves 

were declared inoperable because of the pressure drop rate of the stored air 

supply for the valves. The air operated valves were in the closed position at 

the time of the event. The cause for the pressure drop rate was collective air 

leakage at some of the air supply connections and leakage past the seat of the 

pressure relief valve(s) installed in each air supply (Trains 'A' and 'B) to the 

air operated valves. The air supply had been modified, tested and placed in 

service just prior to this event using acceptance criteria that established 

minimum operability without sufficient margin to accommodate increased 

leakage. 

This is considered to be a common-cause failure condition.

TnhI9 (cntinnAV

29389002 890110 A shutdown was initiated by the Nuclear Engineering Department upon 

notification to Station Management of a potential problem with the air 

supply (air accumulators) for two primary containment air operated valves. 

The discovery resulted from an analysis of the non-safety-related Instrument 

Air System conducted per NRC Generic Letter 88-14. Redundant check 

valves were capable of ensuring the primary containment function. The 

cause is attributed to insufficient capacity of the original accumulators dating 

back to original plant construction. 

This is considered to be a common-cause failure condition.

(continued)

Pilgrim

Pilgrim Containment 

Isolation System
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System 

Pilgrim 29397025 971123 A Technical Specification (TS) required shutdown was completed because c Main Steam, 

two main steam isolation valves (MSIVs) in separate main steam lines were Containment 

inoperable. 

The root cause of the failure of MSIVs AO-203-2B and -IC to close via their 

push button was main closure spring relaxation leading to loss of closure 

force at the end of the closure stroke. A contributing cause of the failure of 

MSIV AO-203-IC to fully close was increased friction between the spring 

plate and actuator stanchions due to improper adjustment or loosening of the 

cam rollers. A contributing cause of the failure of MSIV AO-203-2B to fully 

close was improper installation of the setscrew that locks the stem threaded 

connection together in the hydraulic dashpot. The setscrew was able to back 

out and was broken off during valve stroking. As a loose part in the 

hydraulic dashpot, there was evidence of some minor scraping between the 

hydraulic piston and the cylinder wall that could have increased overall 

assembly resistance. 

Corrective action taken included replacement of some closing springs and 

overhaul of actuators. Corrective action planned includes replacement of 

closing springs on the remaining MSIVs, evaluation of alternate spring 

designs, and revision of maintenance procedures. This is considered to be a 

common-cause failure condition. 

The main steam isolation valves (MSIVs) form part of the nuclear system 

primary containment. The MSIVs isolate the main steam pipelines to limit 

the reactor coolant loss and radioactive material release. There are two 

MSIVs per main steam line. The inboard valves are located just inside 

primary containment (AO-203- 1 A, -I B, -I C, - ID), and the outboard valves 

are immediately outside of primary containment (AO-203-2A, -2B, -2C, 

-2D). Either valve is capable of isolating the line. The control systems for the 

inboard and outboard valves are independent of each other. The MSIVs are 

twenty inch, Y-pattern valves manufactured by the Atwood and Morrill 

Company, Model 20849-H. An accumulator is connected to each MSIV 

between the pneumatic supply and the air cylinder valves. The accumulator 

Z functions to store pneumatic energy for closing the MSIV if the supply of 

pneumatic energy (air or nitrogen) is lost. The accumulator volume provides 

M full valve movement through one-half cycle, open to close. The MSIVs are 

designed to close via spring force (only) if pneumatic power is not available 

C 2 to assist closing the valves. . 

See Event No. 33360 in Table 4.
Table 2. (continued).

LER Event Number' Date Descriotion

z 

'ri 

n0

Classification 2

c

System

Main Feedwater

Point Beach 1 26697014 970321 Engineers discovered a condition that alone could have prevented the 

Auxiliary Feedwater (AFW) System from automatically performing its 

safety-related function during design basis accidents involving a loss of 

instrument air and reduced steam generator pressures. A loss of instrument 

air during the accident would cause both motor-driven AFW pump 

(MDAFWP) flow control valves to fail open. The postulated event scenario 

is the result of a latent characteristic of the original AFW system design. The 

original design did not provide ample assurance that the MDAFWPs would 

automatically function during all design basis events. The expected fix was 

to furnish a reliable pneumatic supply to the control valves. Event No. 31995 

refers. This is considered to be a common-cause failure condition.

Pilgrim

Plant

29397026 971206 A Scram occurred while ascending to power from the outage described in 

LER 29397025. The cause of the scram was the failure of the "A" feedwater 

regulating valve (FRV) in the full open position due to the misalignment of 

the valve clip inside the pilot valve assembly of the positioner. Corrective 

action taken included replacing the valve positioner and performing 

calibration adjustments. 

The Scram root-cause was traced to the opening of the "A" FRV (FV-642A). 

The opening of FV-642A was caused by the inadvertent misalignment of the 

pilot valve clip inside the pilot valve assembly of the Bailey positioner for 

the "A" FRV. The pilot valve clip was misaligned when the positioner was 

opened during the MSIV forced outage (LER 29397025). After the scram, 

when the positioner for the "A" FRV was opened, the pilot valve clip was 

found to be displaced. This caused the positioner's pilot valve stem to move 

downwards, porting air from the top of the "A" FRV actuator, allowing the 

valve to go full open. The feedwater regulating valve, FV-642A, is a Copes

Vulcan fourteen inch, 900 psi, double poppet, balanced, hydraulically 

dampened, diaphragm operated control valve equipped with a DI00-160 

actuator and a Bailey AVI positioner. 

See Event No. 33360 in Table 4.

CD' 

,.W

c Auxiliary 

Feedwater

Pln Number'...... Dat
Table 2. (continued).

LER Event

Nu.mberI Date Descrition

Point Beach I

z 

(-I'

26698008 980203 A 10 CFR 21 Notification (No. 1998-12-0) was received from Engine 

Systems, Inc., which indicated that the air start motor SOVs used on 

Emergency Diesel Generators (EDGs) do not meet the minimum DC voltage 

requirements when inlet pressures below 200 psig are applied to the SOV. 

Seven of the SOVs were sold to Wisconsin Electric. The valves were found 

on all four EDGs. Calculations performed to assess available DC voltage at 

the air start SOVs indicated that only one of the EDGs (G03) had sufficient 

post-accident voltage. Since G03 was only connected to one of the 

safeguards buses, the plant was considered to be outside the design basis. 

The SOVs rely on system air pressure to assist the coils in overcoming the 

force of the springs and thus operating the valve plungers. The original valve 

design assumed air pressure at 200 psig and, therefore, had a weaker spring 

than (later) supplied with the 275 psig rated valve. The SOVs operate with 

air pressure under the seat at a minimum of 165 psig for GOI and G02 

(Train A) and for G03 and G04 (Train B). The valves are supplied with DC 

control power. The voltage at the valve is dependent on the battery load 

during an accident scenario that uses the battery alone under assumed 

accident load conditions. In the case of these valves, the recommended 

minimum voltage for the valves was 105 Volts DC. Therefore, the effect of 

control power voltages less than that recommended by the vendor would not 

ensure the valve would open due to the stronger spring in the 275 psig rated 

valve if the air pressure at the valve was less than 200 psig. Initial 

calculations for the expected at each air start assuming accident DC load 

conditions resulted in voltages less than 105 V DC for GOI, G02 and G04. In 

summary, the cause of the condition was the use of a 275 psig solenoid valve 

which does not meet the minimum DC voltage available during an accident, 

which requires the station batteries for control power. The SOV relies on 

system air pressure to assist the coil in overcoming the force of the spring 

and thus operating the valve plunger. Reduced system pressure combined 

with reduced coil voltage results in the inability of the valve to operate 

satisfactorily. This problem is not applicable to the original 200 psig valve 

(identical to the 275 psi valve except for the spring) because it has a weaker 

spring, and therefore, less air pressure is needed to overcome the lower 

spring force. There are 335 SOVs, distributed among 17 nuclear plant 

owners, listed in the Part 21 notice. Clinton LER 46198009 also refers to 

Engine Systems Part 21 Notification #1998-12-O.This is considered to be a 

common-cause failure condition.

Plant

K c

System

EDGs

Descriotion Classification 2
Table 2. (continued).

LER Event 

Number' Date Description

45889022 890502 It was identified that the safety related air dryer outlet isolation valves might 

not perform their design function on the loss of the non-safety- related air 

dryer or piping. The root cause of this condition was an engineering design 

oversight in that these valves were installed in the direction of normal air 

flow (flow above the disc), instead of in the direction required to isolate the 

post accident safety-related air supply from the non-safety related dryer and 

associated piping.

Classification2

C

System

Instrument Air

26194002 940425

27291030 910920

This LER was included as an example of common-cause failure in 

AEOD/92-02, "Insights From Common-Mode Failure Events." This was an 

ASP Program Event. See Appendix B. 

During Main Steam Isolation Valve (MSIV) operability testing, all three 

MSIVs were declared inoperable when they failed to meet the Technical 

Specification (TS) required closure time. 

The original MSIV actuator design incorporated minimum design margin. 

The potential for failure of the MSIVs to meet their TS stroke time 

requirements undetected was caused by failure of the surveillance and 

modification testing program to account for the minimal design margin of 

the actuators. 

The accumulators were minimally sized to close the valves within five 

seconds under hot, no load conditions, and to maintain the valves closed if 

non-safety-related instrument air was not available to the valve actuators. 

This is considered to be a common-cause-failure condition. 

Both Pressurizer Pressure Operated Relief Valves (PORVs) failed to open 

during a functional check. The cause was degraded Buna-N diaphragm 

material causing air leakage from the valve actuators. The initiating cause of 

both PORVs failing to open was loosening of the fasteners in the valves' 

actuator diaphragm enclosures. The Buna-N diaphragm material incurred 

"creep" where the diaphragm changed from its original geometry under load 

and over time. Elevated ambient temperature degraded the Buna-N 

diaphragm material resulting in the material taking a permanent set, a loss of 

resilience, and extrusion of the material from the clamped region. Permanent 

thinning of the material in the clamped region caused loss of fastener 

preload. This was observed as loosening of the diaphragm fastener cap 

screws/nuts, which allowed a leakage pathway for the diaphragm control air. 

This is considered to be a common-cause-failure condition. This was an ASP 

Program Event. See Appendix B.

c

c

Main Steam 

Isolation Valves

Pressurizer 

Pressure 

Operated Relief 

Valves

z 

1-fl 0

Plant

River Bend

Robinson

CD 

M"

00

Salem I

Plant.. ....
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System 

Salem 1 27296012 960723 On July 23, 1996, a review determined that the keys/keyways on the c RHR 

actuators for the Residual Heat Removal (RHR) flow control valves 

(6 AOVs per unit) are subject to failure. The valves were made by Fisher 

Controls International, Model Type 656/7600. 

When using the simplified Fisher Catalog 14 methodology, the calculated 

maximum stem torque exceeds the vendor specified allowable torque. 

Preliminary calculations of the shaft torque and resulting average shear 

stresses in the valve stem key and keyway were also performed. The 

calculations showed that for normal operating conditions, the average shear 

stress in the key may exceed the material yield stress. The calculations also 

showed that although the average shear stress in the valve stem was 

estimated to be less than the minimum material yield stress, the fatigue life 

of the shaft keyway appeared to be limited to a low number of cycles. Key 

failures in the past are attributed to an overload during normal operation of 

these valves. A review of the original design revealed that these valves were 

installed with little or no design margin and the keys were likely to fail due 

to low cycle fatigue with stress levels exceeding yield strength. Corrective 

action was to replace the valves and review Fisher Model 7600 valves for 

similar concerns. This is considered to be a common-cause failure condition. 

San Onofre 2 36196011 961216 On December 10, 1996, the licensee, in developing a test program for air c PCIS 

operated valves, applied test equipment to containment isolation valve 

2HV0513. It was likely that the actuator settings for valve 2HV0513 would 

not generate sufficient closing force to overcome internal pressure and 

packing drag under design basis conditions. Edison concluded that valve 

2HV0513 had been inoperable when Unit 2 was in Mode 1. 

The licensee concluded that either of two separate errors could have caused 

valve 2HV0513 to have insufficient closing force: (1) vendor setpoint 

methodology error; or (2) a deficiency in the reassembly procedure. 

The licensee committed to reset and retest valve 2HV0513 prior to returning 

Unit 2 to Mode 4. They also were to upgrade its maintenance procedures to 

z conform to the current vendor manual. They performed an analysis on all air

operated containment isolation valves and confirmed that all other Units 2 

and 3 air operated containment isolation valves have sufficient actuator 

closing force to remain operable. The licensee also was to verify actuator 

settings for Unit 2 air operated containment isolation valves which may not 

have had their actuator settings established in accordance with the vendor 

recommendations. Similar actions were to have been completed for Unit 3 

during its next refueling outage.
Table 2. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

32792018 a Main Feedwater 

Essential Air 

Non-essential 

Air System

Sequoyah I

Shearon Harris

32797012 970801

40098001 980109 

(Event 

No. 3513 

refers)

While implementing a modification to improve the materiel condition of the 

control and Service Air System the control air header pressure dropped 

rapidly after modifications personnel cut into-a six-inch control air header 

located inside an equipment clearance boundary. The loss of control air 

pressure resulted in a runback of the Unit I and Unit 2 Turbines and 

instabilities in Unit 1 secondary side flows sufficient to warrant a manual trip 

of the Unit I reactor. The immediate cause of the loss of control air header 

pressure was corrosion products (rust debris) that inhibited full closure of 

one of the six-inch gate valves used as a clearance boundary. The corrosion 

products were assumed to have resulted from water intrusion that had 

occurred about five years earlier. 

A potential failure mechanism existed where a leak in the non-safety 

Instrument Air System could result in the inoperability of the Steam 

Generator (S/G) Pre-heater Bypass Isolation Valves (1AF-64, IAF-102, and 

I AF-8 1. These valves are safety-related containment isolation valves that are 

required to automatically shut in 10 seconds or less upon receipt of a Main 

Feedwater Isolation Signal (MFIS). These valves are each positioned by a 

pneumatic piston-operated actuator and are opened and closed by high 

pressure air (approximately 150 psig) from the (continued on next page)

c

c

Control and 

Service Air 

System

Steam Generator

Sequoyah I 921026 Steam generator (SG) No. 3 was experiencing low level because of irregular 

feedwater regulating valve (FRV) operation. The unit operator placed the 

flow indicating controller in manual for SG No. 3 FRV to attempt to 

gradually raise the level. The automatic-manual switching relay (K-1) for 

this flow indicating controller did not change state, causing the manual 

control circuit to be electrically inoperable. This condition resulted in the 

FRV going to the full open position and not responding to the manual 

changes to the controller input, thereby causing a reactor trip because of a 

turbine trip that resulted from a high-high feedwater level in the No. 3 SG. 

Approximately 1,000 gallons of water was entrained in the nonessential air 

system which supplies the essential air system. The water saturated the 

essential air dryers. The saturation of the air dryers caused the nonessential 

control air system pressure to drop, resulting in an automatic isolation of the 

essential control air system. Some water was carried over to end-use devices. 

The first components affected by the water entrainment were the FRVs. The 

FRV controllers malfunctioned because of the entrained water. 

This is considered to be a common-cause-failure condition.

C. 

Cn

0
Table 2. (continued).

LER Event 

Plant Number' Date Description Classification 2 System 

Shearon Harris 40098001 980109 actuator's accumulator. The accumulator is maintained at a higher air c Steam Generator 

(Event pressure by an air intensifier pump. 

No. 3513 However, an air leak was postulated in the non-safety-related IA piping that 

refers) could possibly reduce the air inlet pressure to just low enough to affect 

(continued) proper operation of the actuator's 3-way and 4-way pilot valves and not be 

detected by the pressure switches in the main header of the Instrument Air 

System. If this occurred, the pilot valves would shuttle, causing the 

accumulator pressure to bleed off, which would prevent the valves from 

closing as required. Operations personnel would have no indication of . 

accumulator low pressure other than local observations made by an auxiliary 

operator and possibly dual valve indication in the main control room due to 

the valves cycling slightly. This potential scenario was reported to the NRC 

via the emergency notification system on January 9, 1998. (Refer to Event 

No. 33513.) 

The cause of this condition was inadequate design control during 

development of a plant modification implemented in August 1984, which 

"was prior to issuance of the Harris Plant Operating License. Specifically, 

NRC Information Notice 82-25, "Failure of Hiller Actuators Upon Gradual 

Loss of Air Pressure," stated that on a gradual loss of control air, the 

pneumatic control valves may assume some intermediate position and cause 

the stored air in the accumulator to vent to atmosphere and prevent the 

actuator from performing its safety function of closing. 

The investigation for this event also revealed additional missed opportunities 

to identify this design deficiency in the Instrument Air System. These 

included: (1) HNP's response development for NRC Generic Letter 88-14 in 

1989; (2) Adverse Condition Report #91-314 initiated in June 1991, which 

identified a leak in the supply air regulator for IAF-81 that significantly 

lowered accumulator pressure; and (3) development of PCR-6158 and its 

associated evaluation (PCR-6066) in 1992, which implemented a 

modification to the actuator air circuitry to enhance the air intensifier pump 

z and upgrade portions of the control air piping to safety-related. In each of 

C these cases, the "smart" leak scenario described in the LER was either not 

identified or not considered credible due to the incorrect assumption that the 

tCl system design met the single failure criteria. This is considered to be a 

common-cause failure condition that existed for 13 years. 

&I. 

.CD
Table 2. (continued).

LER 

Plant Numberl 

Three Mile 28986007 

Island I

Event Taerriflttflfl

.ate '- I F Classification

860326

Turkey Point 3 & 4 25085020 850723

to.

Classification 2

The TMI-1 FSAR states in Subsection 9.10.3.2 that the Two-Hour Backup c 

Air Supply System meets the single failure criteria. As a result of a Safety 

System Functional Inspection by the Performance Appraisal Team, a 

concern was developed that the installed Two-Hour Backup Air Supply 

System may not meet the single failure criteria. 

If the Backup Air Supply were lost, air pressure to control Emergency 

Feedwater System Valves and Atmospheric Dump Valves would result, 

requiring manual action by the operators to control the affected valves. 

Upon further investigation by GPUN, it was determined that indeed the 

single failure criteria was not met. The root cause for the above was an error 

in the original system design. The design verification performed did not 

identify that the final design was not in accordance with the System Design 

Description requirements.

Turkey Point Units 3 and 4 were notified by the Power Plant Engineering 

Department of a 10 CFR part 21 deficiency concerning the ability to close 

the main steam isolation valves (MSIVs). Each MSIV is a check valve 

installed in the reverse direction. MSIV closure is assisted by instrument air, 

a partial travel spring and steam flow. Under low steam flow conditions and 

a loss of non-safety-related instrument air pressure, the accumulator air 

volume may not be sufficient to close the MSIVs. The inability to close the 

MSIVs during an uncontrolled steam release could be postulated to result in 

a loss of the steam generators as a secondary heat sink. The design of the 

MSIVs was to have been upgraded to ensure that each MSIV will meet the 

Final Safety Analysis Report closure criteria without steam flow assistance. 

This is considered to be a common-cause failure condition. 

Refer to NUREG-1275, Vol. 2, for additional information.

System

Two-Hour Backup Air 

Supply System 

EFW 

Atmospheric 

Dump Valves 

Main Steam 

Isolation Valves

CD 

rA

C
Table 2. (continued).

LER Event 

P1 Number' Date Description

Turkey P

Classification 2

a

System

Auxiliary 

Feedwater 

Instrument Air

Vermont Yankee 27198025 981211 

(Rev. 2)

On 12/11/98, with the plant at 100% power, it was determined that one 

Scram Discharge Volume (SDV) drain valve (CRD-LCV-33B) did not meet 

the stroke time requirements of the In-Service Test (IST) Program. Both the 

North and the South SDV's have two drain valves in series, 

CRD-LCV-33A/C on the North and CRD-LCV-33B/D on the South. The 

drain valve, CRD-LCV-33B, was subsequently declared inoperable. 

The SDVs are used to limit the loss of and contain the reactor vessel water 

from all control rod drives during a scram. These volumes are provided in 

the scram discharge header. During normal plant operation the volumes are 

empty with all the drain and vent valves open. Upon receipt of a scram 

signal, the vent and drain valves close. While scrammed, the control rod 

drive seal leakage continues to flow to the discharge volumes until the 

discharge volume equals reactor pressure. Following a scram, when the 

scram signals are cleared and the Reactor Protection System (RPS) logic is 

manually reset, the vent and drain valves are opened and the SDV drained. 

(continued on next page)

ant

oint 3 25085021 850722 The "B" SG bypass feedwater control valve failed to open following a 

reactor trip. About 3 hours later, the "C" SG bypass feedwater control valve 

failed open. 

After making repairs to correct the problem of the reactor trip, operability 

checks of AFW flow control valves CV-3-2832 ("B" SG) and CV-3-2833 

("C" SG) for train 2 showed the valves would not close and were declared 

inoperable. 

The presence of moisture in the instrument air system caused the formation 

of corrosion products which, along with the moisture, were supplied to the 

valve actuators and related control equipment. 

This event was the common cause of a number of related, simultaneous 

AOV and AOV-control equipment failures. Refer to Section 5.1.2 of 

NUREG-1275, Vol. 2 for a detailed discussion of the conditions. The 

licensee has since resolved these conditions satisfactorily, primarily by 

dramatically improving instrument air quality. This was an ASP Program 

Event. See Appendix B. 

This is considered to be a common-cause failure condition.

c

C-il

Control Rod 

Drive

,-
T~ihlci 9 (~ntiniiidh

LER Event

I 1 t NI. M.D te' Descri. tion

Classification 2

Vermont Yankee 27198025 

(Rev. 2) 

(continued)

981211 According to the LER, the root causes of this event were inadequate actuator 

sizing calculations (performed by the vendor) under the design conditions 

specified in the procurement specification, and inadequate manufacturing 

Quality Assurance controls to ensure specification requirements were 

maintained. Contributing causes included conflicting information in the 

design package that was not identified by Vermont Yankee, and closing 

forces of the actuator at either end of the valve stroke were apparently not as 

designed.

The actuators and valves were installed under Engineering Design Change 

EDCR 97410 in April of 1998 and were sized in accordance with the 

vendor's recommendations. These drain valves are required to go shut on a 

scram signal to isolate the SDV and act as a primary containment isolation 

valve when the scram valves are open. This issue was addressed by the 

installation of larger actuators, on December 21, 1998, via the Minor 

Modification process, which were properly sized to operate the valves under 

any design conditions at Vermont Yankee. The licensee stated that the LER 

constituted a Part 21 notification in accordance with IOCFR21.2(c) and 

NUREG-1022, Revision 1. Further, the licensee informed NRC by 

Event 35150 dated December 14, 1998, that this event was considered to be 

a common-cause failure. 

Notes: 

I. The NRC Licensee Event Report (LER) Number consists of the three-digit NRC Docket Number for the plant at which the event occurred, the last two digits of the year in 

which the LER was generated, and a three-digit sequential number of the LER. This is consistent with the NRC's Sequence Code Search System (SCSS) database designation. The 

LER system is described in 10 CFR 50.73. 

2. Events in the table are classified as follows: 

a. Failures of AOVs or events resulting from contaminated (not clean and dry) air; 

b. Failures of AOVs or events resulting from solenoid failures; and/or 

c. Failures of AOVs or events resulting from design deficiencies, material failure, or insufficient margin.

Plont

c

System._

Control Rod 

Drive

Tcahlda 9 (,-nnt;n"o-AI

Descrintion tJj
Table 3. Selected events documented by NRC Licensee Event Reports (LERs) involving air-operated dampers or ventilation components that are 

powered from air systems. 

LER Event 

Plant Number' Date Description Classification2 System

Browns Ferry 2 26095008 951004 The inboard and outboard reactor zone isolation dampers failed to close 

during a weekly routine switch of operating equipment, resulting in the loss 

of secondary containment. Six SOVs serving six valve operators and 

dampers were involved. The cause was traced to failure of the solenoid 

valves which control the dampers. The SOVs failed because of the presence 

on a black sticky residue (adhesive) that caused the solenoid core to stick to 

the plugnut. The source of the residue and its composition were under 

investigation by ASCO when the LER was transmitted to the NRC. 

Additional SOV failures were discovered after the LER was first transmitted. 

Browns Ferry 3 LER 26992003, which refers to failures of the outboard 

containment exhaust damper to close due to unknown-cause SOV failure, is 

referred to in this LER. 

The LER indicated that the event was not reportable to the industry's 

database. 

This is considered to be a common-cause failure condition.

Calvert Cliffs I 31789018 891106 A condition was discovered that could have prevented the fulfillment of 

certain systems to remove residual heat and control the release of radioactive 

material after a Loss of Coolant Accident. During the performance of a test 

to satisfy specific recommendations in Generic Letter 88-14, it was 

discovered that many air-operated control valves and piston-operated 

ventilation dampers which utilize safety-related air accumulators would not 

have performed as expected after a loss of normal non-safety-related 

instrument air. 

The root cause of the event was identified as a lack of an adequate 

documented design basis combined with inadequacies in the testing and 

preventative maintenance program for the Instrument Air System. 

This is considered to be a common-cause failure condition.

C Instrument Air

b, c Containment

CD 

Cjn
Table 3. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

31597023 970916 A design change to the bypass dampers for the Engineered Safeguards 

Features (ESF) ventilation system installed between December 1996, and 

August 1997, introduced the possibility of a single failure which could result 

(The event in the loss of both trains of the ESF ventilation system. The loss of the 

covered by 85 psig air header without concurrent loss of the 20 psig air header would 

this LER result in the ESF ventilation trains being unable to meet their design 

was function.

c ESF Ventilation

The ESF vent system charcoal inlet and bypass dampers originally both used 

a 20 psig air header. The charcoal bypass dampers were normally open and 

were intended to fail closed. The charcoal inlet dampers were normally 

closed and intended to fail open. The licensee installed new bypass dampers, 

which required higher pressure to operate and were therefore transferred to 

the 85 psig air header. If 85 psig air was lost, the bypass dampers would 

reposition to the closed position. The inlet dampers would remain closed and 

this would result in dead-heading of the filter trains and subsequent loss of 

cooling to ECCS equipment. 

The root cause of the event is the failure of the design change process to 

identify the potential adverse impact on the ESF ventilation system created 

by the modification of the control air supply to the bypass dampers. 

This is considered to be a common-cause failure condition.

33389004 890309 Engineering identified a design deficiency which originated during plant 

construction. The deficiency would result in loss of area cooling for parts of 

both safety divisions of safety-related and non-safety-related electrical 

distribution systems as a result of loss of instrument air to the cooling system 

temperature control valves. 

The event was caused by an error during design and construction of the 

plant. Review of the original design and procurement documentation 

indicates that the designer selected, from the vendor specification sheet, a 

valve described as a fail-open valve. It appears that the vendor specification 

sheet used by the designer contained an error or the designer misinterpreted 

the specification sheet. The actual valves ordered and installed were of a fail

closed design. 

A corresponding 10 CFR Part 21 report was not found regarding this 

condition.

C Instrument Air 

Electrical 

Cooling System

Cook 1

originally 

reported as 

Event 

No. 32939)

,..

Fitzpatrick
Table 3. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

Fort Calhoun a Diesel Generator 

Cooling System

Indian Point 2 24788017 881102 An engineering analysis of the capability of the ventilation system to 

maintain the EDG building internal temperature at or below the continuous 

rated temperature of the electronic equipment associated with EDG operation 

was conducted with unsatisfactory results. The ventilation system is 

pneumatic operated (i.e., the intake and exhaust dampers are operated by 

pneumatic actuators and the fans are energized via pneumatic switches). It 

was postulated that if one of two pressure regulator valves in the supply line 

from the safety-related air supply to the pneumatic controls failed, it could 

prevent the ventilation system from operating as intended. 

This is considered to be a common-cause failure condition.

c EDG Ventilation 

System

28587025 870923 Diesel Generator #2 automatically shut down on high coolant temperature 

due to inoperability of the air operated exhaust damper for the diesel 

radiator. Approximately 14 minutes into a test, DG-2 automatically shut 

down due to high coolant temperature. The air operated exhaust damper for 

the diesel generator radiator may not have fully opened automatically as 

designed when the diesel was running, thus restricting the required air flow 

through the radiator. The cause of the damper malfunction was postulated to 

be the presence of residue (see below) causing the pilot valve that directs air 

flow to sometimes stick. 

Previously (6/87), water had been introduced into the instrument air system 

during surveillance test on the fire protection system dry pipe valve for the 

diesel generator rooms (refer to LER 28587033 in Table 2). An extensive 

program was undertaken blowdown air operated devices including air 

operated valves and to cycle those valves as allowed during power operation. 

After the trip of DG-2, the pilot valve was inspected and cleaned and the 

accumulator drained. Similar actions were taken for DG-1. This was an ASP 

Program Event. See Appendix B. 

Note that the contamination of the air system caused AOV failures after 

cleanup and after some time had passed. This is considered to be a common

cause failure condition, based on contamination of the instrument air system.

M 

0~ 

61
Table 3. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

Indian Point 3

z 

4r1

c

H

Central Control 

Room (CCR) 

Ventilation 

System

Indian Point 3 28696002 960120 During operator actions to secure Emergency Diesel Generator (EDG) 32, 

room ventilation did not operate as required. EDG 32 was declared 

inoperable resulting in a condition prohibited by technical specifications 

(i.e., two inoperable EDGs). The EDG 32 room ventilation malfunction was 

caused by debris in the pressure regulator used to supply air to the ventilation 

control system. The regulator was disassembled and debris was observed, 

particularly in the valve seating area. Also the valve/stem assembly was 

found to be loose. The engineers attributed the debris to internal scaling of 

the carbon steel piping which is connected to the inlet of the regulators. 

On January 22, 1996, a Deviation Event Report (DER) was issued for the 

pressure regulator of EDG 33 room ventilation system because it required 

constant adjustments in order to maintain the output to 100 psig. The 

regulator was replaced and the old regulator disassembled. The engineers 

observed debris in the valve seating area and the valve/stem assembly was 

disengaged from the diaphragm. 

The EDG 32 regulator that had been replaced on January 21, 1996, was 

removed and bench tested. The test showed that the replacement regulator 

for EDG 32 was unable to maintain pressure. 

Upon disassembly, the replacement regulator for the EDG 32 valve/stem 

assembly was found disconnected but with no debris and it was replaced 

with another new regulator. A failure analysis was to be performed on the 

failed EDG 32 regulator but the results are unknown. 

The EDG 31 regulator was removed, tested and inspected. Although the 

regulator maintained air pressure, debris was noted in the internals and the 

valve/stem was loose. The pressure regulator for EDG 31 was replaced. 

(continued on next page)

28693045 931023 An investigation identified the loss of Instrument Air as a failure mode of the 

Central Control Room ventilation system damper actuators that would result 

in loss of system function. Loss of this function is a condition prohibited by 

technical specifications. This event was caused by personnel error of an 

indeterminate origin during original design. 

The architect engineer did not evaluate this failure mode during initial design 

or when upgrading the system from a non safety system to a safety system 

prior to initial operation. 

This is considered to be a common-cause failure condition.

00

a Emergency 

Diesel 

Generators
Table 3. (continued). 

LER Event 

Plant Number' Date Description Classification 2 System

Indian Point 3

Indian Point 3

Indian Point 3

28696002 

(continued)

960120

28693013 930414

28693036 930915

The regulators are commercial grade and dedicated for Category I service. 

New York Power Authority (NYPA) reported the loose valve/stem assembly 

to the manufacturer. 

NYPA replaced the carbon steel piping from the starting air receiver tank to 

the regulator, added a filter to the inlet of the regulators, and added a drip leg 

to piping upstream of the regulators to help trap any debris before it enters 

the regulator. 

This is considered to be a common-cause failure condition. 

While performing System Operating Procedure SOP-CB- 10, "Fan Cooler 

Unit Operations," it was observed that closing the dampers via their 

respective Central Control Room (CCR) control switches did not result in 

actuation of the close indication lamps for A and B dampers on 32, 34 and 

35 Fan Cooler Units. A mechanic reported that the damper linkages and 

bearing housings were dirty for all dampers. The pins connecting the 

linkages were rusty and paint chips were noted on the arms where the 

linkages connect to the louvers. The rust was cleaned from the pins and dirt 

and rust was cleaned from the linkages. The linkage arms and pistons were 

lubricated with Mobil AW-2 Grease. The dampers were operated from the 

Control Room and found to perform as intended. 

This is considered to be a common-cause failure condition. 

A Nuclear Regulatory Commission inspector posed questions regarding the 

stop nuts installed on damper (DMP) actuators in the Central Control Room 

(CCR) ventilation system. The inspector also questioned the proper 

mounting of damper actuators. The Heating, Ventilating and Air 

Conditioning (HVAC) system engineer produced documentation which 

indicated that the damper actuators were inadequately mounted to meet 

seismic qualification. The CCR HVAC system design requires that the 

system withstand seismic events. The system engineer determined that the 

inadequate seismic mounting of the damper actuators had existed since 

initial criticality, which took place on April 6, 1976. 

This is considered to be a common-cause failure condition.

a

a

c

Emergency 

Diesel 

Generators

Central Control 

Room (CCR) 

Ventilation 

System

Central Control 

Room (CCR) 

Ventilation 

System

z 

t'n 

&I 

LA 4ý1.

0D 

Wb
Table 3. (continued). 

LER Event 

Plant Number1 Date Description Classification2 System

34197007 970325 One of the two operating Reactor/Auxiliary Building Ventilation System 

(RBHVAC) exhaust fan discharge dampers closed, resulting in high reactor 

building pressure and a loss of secondary containment integrity. The cause 

was age and service related mechanical failure of the SOV that controls 

opening air to the discharge damper for the exhaust fan. The exhaust port 

seating material on the SOV was completely disintegrated and allowed air to 

leak through the exhaust port.

b Reactor Building 

Ventilation 

System

37389007 890210

37397046 

and 

37398007 

(Event 

No. 33434 

also 

referred.)

971216 

and 

980319

Review of the Control Room HVAC System drawings identified one 

potential situation where an undetected single failure (a hot short circuit) 

could result in a failure of the normally closed, fail closed, maximum 

exhaust air isolation damper, OVC14YA/B in the open position on the 

operating train. The postulated hot short circuit would cause the damper to 

move to the fully open position and the Control Room Operators would have 

no direct indication (e.g., alarm or valve position indication) of the fact that 

the damper was open. Under this set of circumstances, it is possible that 

smoke or radioactivity could be drawn into the Control Room in spite of the 

initiation of the emergency make-up train. The reviewers concluded that the 

postulated sustained hot short was beyond the design requirements of 

LaSalle County Station. 

(LER 37398007) An investigation of the allowable closure time assumptions 

for the turbine building high energy line break (HELB) check dampers 

(#1/2VT79YA/B/C) led the licensee to reconsider the original builders 

assumptions and calculations for the reactor building exhaust isolation 

dampers (1VR05YA and B). 

(LER 37397046) In the event of a Main Steam High Energy Line Break, the 

Turbine Building Ventilation isolation dampers, I(2)VTO79YA, B & C, 

would not close fast enough to prevent the pressure from exceeding the 

pressure retaining capability of the walls, floors, and ceilings that separates 

the VT exhaust tunnel from the safety related High Pressure Core Spray 

(HPCS) electrical switchgear room. The apparent causes included an invalid 

calculation assumption dating from the original design. 

Dampers 1VR05YA and B had malfunctioned and were rebuilt in 1985 (see 

LER 37385008 and LER 37385011). The licensee indicated that they revised 

the calculations and assumed a 0.4 second instrument time delay, and a 0.075 

second solenoid valve response time. The plant was licensed in 1982.

b 

b, c

Ut 

0.1

Fermi 2

LaSalle I 

LaSalle I

Control Room 

HVAC System 

Auxiliary 

Building
Table 3. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

37489018 891216 The 250V battery was declared inoperable due to low battery electrolyte 

temperatures. Since this battery serves as a power supply to the Reactor Core 

Isolation Cooling system, the RCIC system was also declared inoperable. 

The cause of this event was the failure of two Division I Switchgear Heat 

Removal System damper actuators coupled with sub-zero outside air 

temperatures. The two dampers involved failed to fully close, resulting in 

sub-zero outside air admission to the switchgear ventilation system.

c Division I 

Battery Room 

Heat Removal 

System

42393002 930207 The A Train Control Room Pressurization System failed its 18-month 

pressurization surveillance test. On February 6th the B Train initially failed 

its surveillance test. Two common-cause failures were identified by the 

licensee in the LER, and both trains were declared inoperable pending 

investigation. 

The first potential common-cause failure was freezing of the pressure 

regulating valve caused by moisture in the air banks. The second common

cause failure was pressure oscillations within the control room envelope 

which caused the differential pressure to periodically fall below the 

acceptable 0.125 in. Hg. for brief periods of time. 

The air banks were purged and refilled with dry air to reduce the dew point. 

Samples of the air in the air bottles of both trains tested at dew points around 

minus 40'F at atmospheric pressure, which corresponds to approximately 

70'F at 2250 psig. Excess moisture could have entered the system due to 

improper blowing down of the condensate traps or from purifying cartridges 

which exceeded their useful life. The system design did not include drying 

capabilities other than the moisture removal capacity of the compressors.

c Control Room 

Pressurization 

System

LaSalle 2

Millstone 3

Z;;

z
Table 3. (continued). 

LER Event 

Plant Number' Date Description Classification 2 System

Prairie Island 1

z 

C) 70~ 

70' Cjb 

t.P

c

(

M-"

Room Chilled 

Water System 

Instrument Air

25492028 921016 During a review of a Standby Gas Treatment (SBGT) Tech Spec revision, it 

was revealed that on a loss of instrument air (IA) the Control Room Dose 

would exceed General Design Criteria (GDC)- 19 limits. This conclusion was 

made assuming that the pneumatically controlled SBGT heaters would fail to 

start, the air operated flow control valve would fail open, flow would 

increase to 5100 cfm, and the Control Room Air Filtration Unit (CR AFU) is 

started in 110 minutes. 

The apparent cause of the event was an inadequate design for the Standby 

Gas Treatment System. The pneumatic flow instruments and the Flow 

Control Valves rely on the non-safety-related Instrument Air system, to 

function properly. The operation of these components upon a loss of 

instrument air had not been thoroughly evaluated. 

The condition was discovered in 1992. This condition had existed since 1972 

when the plant commenced commercial power operation. 

This is considered to be a common-cause failure condition.

c

28295013 950927 During review of the effects of the loss of instrument air and the resultant 

plant response, it was determined that the control room chilled water system 

would not function without instrument air (a non-safety-related system). 

With the chilled water system not functional, the temperature in the control 

room and relay room would exceed equipment qualification temperatures in 

approximately 30 minutes unless operator action was taken. 

Original plant design took credit for instrument air availability after the onset 

of an event. The instrument air system was assumed to be operable because 

it is powered by the safeguards diesel generators. Pre-operational testing of 

the instrument air system demonstrated that all valves fail to their safe 

position, but did not consider the integrated plant response upon the loss of 

instrument air. Also, the need for the control room chilled water system to 

function as essential support equipment was just recently identified. 

This was reported in 1995. This condition had existed since 1973 when the 

p!ant commenced power operation.

t.' Quad Cities I

to)

Control Room
LER Event 

Plant Number' Date Description Classification2 System

890406 It was observed that six solenoid operated valves (SOVs) were oriented to 

isolate normal air flow into the safety-related instrument air system (IAS) 

accumulator. Since these SOVs are directional, they may not have 

maintained air in the accumulators as required when in the closed position. 

This is considered to be a common-cause failure condition.

31197002 970213 

(Event 

No. 31783 

refers)

Four pressure switches (ABS/PDS) on Salem Unit 2 could not be verified as 

seismically qualified. In addition, one of the four non-seismically qualified 

pressure switches was found to have its sensing lines reversed. The pressure 

switches control the position of backdraft dampers that are boundaries for 

environmentally isolated contiguous zones which are vented to the 

atmosphere. Failure of pressure switches could cause failure of damper 

operators and thus could subject equipment to a harsh environment from 

steam. In addition, because one of the pressure switches had its sensing lines 

reversed, its respective isolation damper would not have closed in response 

to a HELB. 

The pressure switches were replaced in 1986 under a routine maintenance 

recurring task. The original pressure switches were seismically qualified but 

the replacement pressure switches were not. The reason for replacement with 

non-seismic pressure switches was unknown. The sensing lines for the 

mechanical penetration area pressure switch were reversed by design in 

1978. The design change incorrectly assumed the mechanical penetration air 

supply duct would be exposed to high energy as a result of a HELB; 

however, the HELB analysis identified that the mechanical penetration area 

would be the high energy area.

b, c Control Building HVAC 

Auxiliary 

Building HVAC 

Fuel Building 

HVAC 

Auxiliary 

Building 

Ventilation

b, c

River Bend 45889024

Salem 2

z 

i-ri 

ON 

ON

,.-

T•hla 't /.-^,•tin,,.rl•
Table 3. (continued). 

LER Event 

Plant Number' Date Description Classification2 System

39587019 870730 The Licensee identified a design deficiency in the instrument air system 

which could cause loss of control room ventilation on a loss of instrument 

air. The air supply to each inlet damper actuator is provided with an air 

accumulator and check valve to ensure that the dampers remain in the open 

position to perform their safety-related function in the event of a loss of the 

instrument air system. While evaluating the system design, it was determined 

that the instrument air supply check valve for each train of outside air inlet 

dampers was not designed to provide a leak tight seat under the system air 

pressure. The event was caused by two improperly designed instrument air 

check valves installed in the IA system which may not have seated properly 

at the line pressure within the instrument air system.

c Instrument Air 

Control Room 

Ventilation

Notes: 

1. The NRC Licensee Event Report (LER) Number consists of the three-digit NRC Docket Number for the plant at which the event occurred, the last two digits of the year in 

which the LER was generated, and a three-digit sequential number of the LER. This is consistent with the NRC's Sequence Code Search System (SCSS) database designation. The 

LER system is described in 10 CFR 50.73. 

2. Events in the table are classified as follows: 

a. Failures of AOVs or events resulting from contaminated (not clean and dry) air; 

b. Failures of AOVs or events resulting from solenoid failures; and/or 

c. Failures of AOVs or events resulting from design deficiencies, material failure, or insufficient margin.

Summer

CD 

t-,y-
LI. A l,......A n t-thpr th!n I cpnp FvAnt Renorts. * dUI.9. OWLLU VIILLZU.LI1U...U iii -.

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designations] Date Description Classification2 System 

Beaver Valley I Event 980422 Design Change Package (DCP) 2313 was issued for installation of a c Pressurizer 

No. 34118 backup nitrogen supply for PORV PCV-456 after discovering that none 

existed even though required by (at least) 1982 Updated FSAR 

commitments, although PCV-456 is not assigned to mitigation of LTOP 

events. PORVs PCV-455C and -455D had been provided with nitrogen 

backup and are assigned to mitigation of LTOP events.

On March 26, 1998, and again on April 7, 1998, it was discovered that: 

(1) The installation of the existing tubing to the air regulators associated 

with PORVs PCV-455C & -455D was too rigid in that it did not 

allow for sufficient thermal expansion/contraction and seismic 

movement associated with stroking the PORV under all design 

conditions; 

(2) The mounting plate for the Solenoid Operated Valves (SOVs) on the 

PORVs was inadequately designed; 

(3) The SOVs were not specifically qualified for excitation resulting 

from PORV discharge; 

(4) The existing Nitrogen tubing was field routed and required a level of 

rework to reestablish seismic adequacy; and 

5) The PORV actuators were not adequately protected from a failure of 

the instrument pressure regulators in that the PORVs could fail in a 

nonconservative manner by opening in a faster time than what was 

evaluated in the system transient pipe loading analysis when the 

Reactor Coolant System (RCS) is at operating temperature and 

pressure. 

The licensee stated that the apparent cause of these events was loss of 

design and licensing basis configuration control which subsequently led 

to inadequate installation, application and implementation of TS 

•z Surveillance Requirements. 

(LER 33498011 was subsequently issued to cover this event and 

following developments described herein.)
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designations' Date Description Classification2 System

Big Rock Point Event 

No. 32226

Event No. 

33832

Failure 

Evaluation of 

ASCO 

Solenoid 

Valves 

Related to 

DERs 

97-1202, and 

97-1200

Z 

eri 

ON 

ON c Control Rod 

Drive

970425 Vent Valve on Control Rod Drive (CRD) System could lift under high 

reactor pressure condition. 

Based on a review prompted by NRC Information Notice 96-68, 

"Incorrect Effective Diaphragm Area Values in Vendor Manual Result in 

Potential Failure of Pneumatic Diaphragm Actuators," it was discovered 

that the vent valve on the CRD system (CV-NC I1A) would not remain 

closed on a system pressure in excess of 1585 psig. This determination 

was based on control air pressure supplied to the valve and the size of the 

air diaphragm in the valve. The licensee was to modify the system (no 

details provided) to ensure that the air-operated valve remained closed. 

No LER was found for this condition. 

This is considered to be a common-cause failure condition. 

980304 Pressurizer PORV 2-NRV-152 could have been rendered inoperable 

because of a leaking air system check valve. It was concluded that the 

emergency backup air supply for the PORV could have been inoperable 

by a leaking air system check valve. Details of the time that the check 

valve was leaking were not provided. 

An LER could not be found to cover this condition. 

970917 Analysis of failed solenoid valves which control safety-related AOVs (18 

failures in a population of 66 SOVs) indicated problems with lubricants 

and thread-locking compounds that caused the valves to stick (the core 

adheres to the wall). The Fermi 2 engineers prepared a comprehensive 

SOV failure analysis report as part of the resolution of the DER. The 

failure analysis report included what they refer to as a common-mode 

failure investigation. The conclusions were: 

- Sufficient material controls and use instructions were not in place to 

reduce the possibility of contamination in the plant's pneumatic 

system. 

- Definition and understanding was needed for the various licensee

defined categories of materials (in terms of use restrictions) in the 

plant. 

(continued on next page)

Pressurizer

a,b,c Various 

Safety-related 

Systems

Cook c

Fermi 2
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designations' Date Description Classification2 System

Failure 

Evaluation of 

ASCO 

Solenoid 

Valves 

Related to 

DERs 

97-1202, and 

97-1200 

(continued)

DER 93-0045 

DER 94-0568 

DER 94-0406

970917 - NUREG-1275, Volume 6 findings and recommendations were not 

thoroughly reviewed for applicability at Fermi 2 and compared to 

their (then) current practices. 

- The corrective and preventive maintenance processes did not 

adequately use actual equipment performance data for SOVs when 

determining the necessary preventive maintenance schedules. (This 

was noted to be a problem for other components, such as pumps, 

valves, and heat exchangers, as well.) 

This is considered to be a common-cause failure condition. 

In the SOV failure analysis report, it was also noted that some SOVs 

were supplied with 125 or 130 volt AC power over their service 

lifetimes, although the equipment qualification program assumed 120 

volt AC power. The report called for further investigation of this 

consideration.

930119 

941015

Moisture in the air system (flasks) resulted in corrosion contamination 

and subsequently caused valve problems. 

Precoat valves were found to be installed backwards since the plant 

started operations. During investigations that followed, it was found that 

the air-to-open, spring-to-close valve actuators (G410OF210A and B) 

were undersized.

940822 The DER referred to Fisher Information Notices (FIN) on valve 

problems. FIN 94-02 and Supplement I notified users of piston rod 

extension problems. FIN 94-03 regarding Fisher 3570 positioners with 

Fisher 472 or 473 spring return actuators indicated problems with return 

to the fail-safe position. FIN 94-04 identified problems with 18 models 

of piston actuated AOVs. Fermi 2 had 8 valves of these type models. 

This is considered to be a common-cause failure condition.

a,b,c

a 

c

Various Safety-related 

Systems

Various Safety related and Non

safety-related 

Systems 

Various Safety

related and Non

safety-related 

System

Fermi 2

t.A

Fermi 2 

Fermi 2 

Fermi 2

C) 

70 

OA

0' 

(b
Z 

C) 

ON

Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designations] Date Description Classification2 System 

Hope Creek Event 970828 During a review of the Safety Auxiliaries Cooling System (SACS) c Safety 

No. 32836 design basis, PSE&G identified a potential design deficiency associated Auxiliaries 

with the safety-related Control Area Chilled Water System chiller units Cooling 

that could have prevented the fulfillment of a safety function. In the 

event of an accident resulting in a loss of instrument air, concurrent with 

low Ultimate Heat Sink temperature, all of the safety-related chillers 

could trip. Specifically, the SACS control valves which regulate SACS 

cooling flow through the chillers are designed to fail open upon loss of 

(non-safety-related) instrument air. These valves control room cooling 

for the Emergency Diesel Generators and the Emergency Core Cooling 

Systems. 

PSE&G was to evaluate actions to correct the condition as part of its 

corrective action program. Corrective actions included an operability 

determination, revising the engineering evaluation, and evaluating and 

implementing a design change. 

This condition had existed since initial plant operation (1986). The cause 

of the failure to recognize and correct this condition was attributed to 

human error in the original design and in several design reviews. LER 

35497020 also referred to this event. 

This is considered to be a common-cause failure condition. 

Indian Point 2 Morning 970203 The plant shut down due to a feedwater regulating valve failure. Main c Main Feedwater 

Report feedwater regulating valves were stuck open due to serious galling in the 

No. 1-97-0011 cages and plugs. The system was contaminated by abrasive grit used to 

clean the turbines. 

This is considered to be a common-cause failure condition.
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designations] Date Description Classification2 System 

O..A.A r ,-N..•,. l . .;tnPn ennlv is insufficient to meet the reauirements of c Various

Indian t'oint L the over-pressure protection system design basis. The licensee 

determined that the available supply of nitrogen may not meet the OPS 

system design basis of 200 operating cycles in 10 minutes. Additional 

nitrogen supplies were provided. 

No LER could be found for this event.

Safety-related and Non-safety

related System

Indian Point 3

The event was not directly related to AOV failures or deteriorated 

conditions, but was included as an example of inadequate design of a 

pneumatic supply system that was discovered recently. 

Event 931202 All three emergency diesel generators (EDGs) were inoperable for 

No. 26449 approximately four hours. (Originally, the diesels were reported 

inoperable for approximately five minutes but this was later corrected to 

four hours.) 

The solenoids on AOVs used to control the EDG service water effluent 

for all three EDGs (two parallel valves in a common header line) were 

replaced and then tested as part of a normal work package on a safety

related system. The valves were stroke-time tested after the solenoids 

were installed and then energized. The solenoids were incorrectly 

connected and that the valves would, therefore, not stroke open. 

LER 28693053 (see the description in Table 2) also refers to this event. 

See Millstone 2 Event No. 26187 for a similar event. 

This is considered to be a common-cause failure condition.

z 

0, 

&I

..

No. 34103

b, c Emergency Diesel 

Generators
z 

61 

0>

Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designationsl Date Description Classification2 System 

Limerick 2 Event No. 971206 Six stuck steam flooding HVAC dampers were caused by sticking b, c, d MSIVs 

31739 solenoids. Excessive friction in either the solenoid or the associated 

bushing appears to be the cause of the failures. The dampers are located 

in 3 ventilation duct penetrations in the walls surrounding the outboard 

MSIV room. Each wall penetration has two steam flooding HVAC 

dampers arranged in series. The dampers are required to close following 

a HELB in the outboard MSIV room to protect equipment in the 

adjoining room. 

The licensee had no analysis for an HELB with steam entering the 

adjoining room. The adjoining room contains safety-related equipment 

such as the scram solenoids, scram hydraulic control units, and several 

ECCS MOV electrical breakers. Shutdown capability could have been 

adversely affected. The licensee had not determined temperature, 

pressure, or humidity to be expected from the scenario. 

This is considered to be a common-cause failure condition. An LER was 

not submitted for this event. 

Millstone 1 Event 970307 The emergency service water system was declared inoperable because of b, c Emergency 

No. 31910 potential failure of a non-qualified pressure regulator in the air system Service Water 

that supplies air to the SOVs that actuate the blowdown valve for the 

ESW strainers. 

The failure of this regulator would subject the SOVs, which are rated at 

85 (psig) to full air system pressure at 100 (psig). Failure of the SOVs 

could prevent the ESW strainer blowdown valves from opening thus 

preventing fulfillment of the ESW safety function. Both trains of the 

ESW system are affected because the regulator provides a common air 

supply to both SOVs. Other SOVs were to have been investigated for the 

same or similar problem. 

This is considered to be a common-cause failure condition.

0D 

CA
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designations Date Description Classification System 

1 .. I M17 l (',t-týhpr '7 1 Q03 at 1830 hours with the unit in Mode 4 (Hot b, c Emergency

No. 26187 Shutdown), the 'B' emergency diesel generator (EDG) was declared 

inoperable because the solenoid valve 2-DG-95B had a lower rated 

differential pressure than that produced by the air start system pressure. 

The "A" EDG was also inoperable at this time due to a failed 

surveillance test. 

The solenoid valves 2-DG-96A and 2-DG-96B were ASCO Model 

206-381-2RF, rated for a MOPD of 200 psid. These valves were 

evaluated as adequate for their intended design since they are the 

equivalent to commercial solenoid valves rated at 300 psid. The nuclear 

grade solenoid model was derated for the purposes of satisfying the more 

rigorous seismic requirements. The "B" diesel generator solenoid valve 

2-DG-95B was an ASCO Model 206-381-3RF with a differential 

pressure rating of 150 psid which is below the air pressure it has to work 

under. The equivalent commercial solenoid for the Model 206-381-3RF 

has a pressure differential rating of 200 psid. The solenoids were 

replaced and the EDGs were returned to service about 19 hours later. 

LER 33693011 refers. See Indian Point 3 Event No. 26449 for a similar 

event.

Diesel 

Generators

H

Ivillstone L

ON1

z 

Mr
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designations' Date Description Classification2 System

Oconee 1,2, and 3

z 

q0 

C')

Inspec- The purpose of these inspections was to follow up on an open item 

tions on regarding potential emergency feedwater (EFW) system design 

Nov. 2-6 vulnerabilities. The discussion of the design and operation of C-187, a 

and 12-inch AOV in one of the parallel 20-inch lines from the upper surge 

16-20, tank (UST) to the condenser hotwell was of particular interest in this 

1998, and study of AOVs. In 1989, the licensee upgraded the EFW seismic 

Jan. resistance capabilities. As part of that upgrade, normally closed AOV C

11-15, 187 was declared safety-related and became the single EFW boundary. 

1999 The licensee incorrectly left AOV C-I187 designed to open on a low 

condenser hotwell level (that would result from a break in any of the 

non-seismic pipes connected to the hotwell or a break in the main 

feedwater line) and dump the UST water to the condenser hotwell, thus 

failing the EFW system. In 1993 and 1994, the licensee modified the 

controls for AOV C-187 to automatically close at a low UST level; 

however, the EFW was still left vulnerable to a single failure of the 

single solenoid valve actuator for C-187. Also, AOV C-187 was still 

relied upon as a single EFW seismic boundary valve. The licensee's 

probabilistic risk assessment recognized that failure of AOV C- 187 was 

one of the top contributors to a potential EFW system failure. The PRA 

stated: "If a main feed line break is assumed, the UST could be drained 

into the hotwell, thereby failing EFW's initial suction source." The issue 

remained open as of the publication of this report.

NRC 

Inspection 

Report 

50-269/99-10, 

50-270/99-10, 

and 

50-287/99-10. 

(Refer to NRC 

Accession 

Number 

9902090072)

,--

Emergency 

Feedwater

ON

C
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designationsi Date Description Classification2 System

Oyster Creek Event No. 

(also Browns Ferry, 32196 

Fitzpatrick, V t M

Monticello, Peach 

Bottom, and Quad 

Cities)

ON..

970422 

and 

970429 /- YKi . V-I . 

32240, Part 21 

from ASCO, 

applies 

See Morning 

Report Nos: 

H-97-0065, 

dated 6/4/97 

and 1-97

0031, dated 

4/29/97 

E-Mail from 

David Skeen, 

dated 5/6/97, 

12:35 pm

Thirteen of 36 scram solenoid pilot valves (SSPVs) at Oyster Creek 

exhibited excessive air leakage during performance maintenance testing. 

The leakage resulted from hardening of the core disks in the valve pilot 

heads. ASCO later issued a Part 21 report which indicated that nearly 

1000 units were fabricated with Buna-N (nitrile) material rather than the 

specified nuclear grade Viton (fluorocarbon) material. GE indicated that 

the air leakage would cause control rods to drift closed but blockage of 

the SSPVs was not considered to be a credible event, and backup scram 

valves are available. No LERs were found that cover these events; 

however, LER 27198025 and NRC IN 94-71 refer to similar problems at 

Vermont Yankee and WNP-2. Event #27149 refers to a similar event at 

Pilgrim. Contrary to GE's analysis in the cases of Pilgrim/WNP-2, the 

licensees indicated failures that did result is blockage of SSPVs that 

could have rod insertion. 

The plants and number of SSPVs affected were: 

Browns Ferry - 5 

Fitzpatrick - 51 

Monticello - 260 

Oyster Creek - 300 

Peach Bottom - 10 

Quad Cities - 372 

This is considered to be a common-cause failure condition.

b, c Scram

z 

ON 

(A 

-4.

,-
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designations' Date Description Classification2 System

z 

rri

Letter from 

P. Flenner, 

Consumers 

Energy, to 

R. Schaff, 

USNRC, dated 

4/23/97, with 

attachments 

Includes 

Condition 

Report 

C-PAL-97

0404 

Palisades 

Engineering 

Analysis EA

AOVSYS

ESS-01, 

Revision 3, 

dated 9/11/97

a Various Safety

related and Non

safety-related 

Systems 

Engineered 

Safeguards and 

Shutdown 

Cooling

Palisades 970315 The letter includes condition reports that describe corrosion in air lines 

which led to degradation or failure of multiple air regulators. 

This was considered to be a common-cause failure event caused by a 

high pressure air system that did not meet the required cleanliness 

criteria, i.e., the system had been contaminated with moisture and 

resulting corrosion products. 

The conditions were not reported in an LER. NRC/AEOD was informed 

of the situation in an event (overnight) report. The attachments to the 

letter include the licensee's rationale for not reporting the occurrence. 

See Section 8.3.1 of this report. 

970911 This entry in the specified report refers to Design Basis Review 

(calculations and specifications) for AOVs in the Engineered Safeguards 

System (ESS). These calculations refer to the CV-3025 AOV discussed 

in Section 8.3.2 of this study. 

The CV-3025 AOV has a design basis function to open during entry into 

the shutdown cooling mode after a small-break LOCA. This valve has a 

function to be throttled to adjust cooling flow after a small-break LOCA, 

but the licensee considers this to be beyond the design basis. 

See LER 25578003 and LER 25581030 in Table 2 for a description of 

the conditions and events involving the CV-3025 AOV.

NA

(A

ON

Palisades
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designations' Date Description Classification 2 System

980305 The original design bases, and subsequent design reviews, for the HP Air c 

System address the Large Break Loss of Coolant Accident (LBLOCA), 

yet did not address, in detail, the SBLOCA. 

HP Air System testing demonstrated that the HP Air receivers can supply 

sufficient air for one hour to operate the control valves required to 

realign ECCS to the recirculation mode when the Recirculation 

Actuation Signal (RAS) occurs. This bounds the expected time frame for 

RAS initiation following LBLOCA events. However, SBLOCA events 

can be postulated, wherein RAS initiation occurs beyond the one hour 

time frame. The HP Air System compressors must be returned to service 

within one hour for postulated events where RAS initiation occurs more 

than one hour into the event. 

LER 25598006, described in Table 2, was subsequently issued to cover 

this event.

Event 971206 (LER 29397025) On November 23, 1997, at 2215 hours, a shutdown was 

No. 33360 completed as required by Technical Specifications, because two main 

(LERs steam isolation valves (MSIVs) in separate main steam lines were 

29397025 and inoperable. The direct cause of the failure of the two MSIVs to close was 

29397026 the relaxation of the closing springs. Corrective action taken for the two 

2372 laterMSIVs included replacement of the closing springs and overhaul of the 

were later 

actuators. The cause for the failure of MSIVs AO-203-2B and -IC to 

idenferied as close via their push button was main closure spring relaxation leading to 

rhisefeingto aloss of closure force at the end of the closure stroke. A contributing cause 

nothdis even aof the failure of MSIV AO-203-IC to fully close was increased friction 

nedription thbetween the spring plate and actuator stanchions. This is considered to be 

Dsri a common-cause failure condition.

c

Various 

Safety-related 

and Non-safety

related System

MSIVs, Main 

Feedwater

(LER 29397026) During power ascension from the shutdown described 

above, a scram occurred as the result of a turbine trip due to high reactor 

water level. The scram was caused by the failure of the "A" feedwater 

regulating valve (Copes-Vulcan D 100-160 diaphragm operated AOV) in 

the full open position. This was caused by the misalignment of a pilot 

valve clip in the positioner when the positioner was opened during the 

forced outage to fix the MSIVs. 

CD

Event No. 

33843

Palisades

Pilgrim

column)

M 

011 (A
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designationsi Date Description Classification2 System

Valcor Engineer'g 

Corp. (and 

Susquehanna)

Valcor Engineer'g 

Corp. (and 

Susquehanna)

z 

t'rl 

4x.

Event 

No. 34262 

See also 

Susquehanna 

Condition 

Report 

No. 98-2296 

(continued)

Event 

No. 34262 

See also 

Susquehanna 

Condition 

Report 

No. 98-2296

980519 This event report describes a 10 CFR Part 21 report (1998-43-1) of a c

deficiency in Valcor Model V70900-65- 11 solenoid-operated pilot 

valves. 

Following 6 to 18 months of continuous energized service, 3 Valcor 

Model V70900-65-1 1 air pilot valves failed to stroke closed in service 

immediately upon de-energization. Delays in closing ranged from one to 

five minutes. Despite their best efforts, the vendor reported that closing 

delays have not been replicated outside of the plant systems. 

The vendor reported that a potential cause of the delayed closing was the 

possible susceptibility of this particular model to the effects of residual 

magnetism. Delayed closing could occur when the air gap between the 

plunger and the stop becomes too small or if the plunger makes contact 

with the stop. Extended periods of continuously energized service and 

subsequent compression setting of the 0-ring seat may induce a 

condition in which the air gap is too small. As a result of the small air 

gap, the valve may then be subject to residual magnetism effects, which 

would tend to prevent closure of the valve when it is de-energized. 

Opening operation of the valve was not affected. (continued on next 

page) 

980519 Further bench testing found additional similar failures. The licensee's 

failure analysis projected that, in the as-found condition, 59% of the 

installed valves would fail to perform their design function before the 

end of their first cycle of operation in the plant (18 months). 

Forty-six units of this particular model have been delivered, all to 

Susquehanna. Similar valves sent to Duke Power had the potential for 

similar failures. 

All of the valves have been, or were scheduled to be repaired by Valcor. 

An LER was not found for this condition. 

This is considered to be a common-cause failure condition.

CD

Scram

c Scram
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designationsi Date Description Classification2 System

Vermont Yankee 

(10 CFR Part 21 

from AV)

Waterford 3

Event 

No. 32253

Letter 

WF3-97-0107 

dated May 6, 

1997 (NRC 

Accession 

Number 

9705120329)

970425 Vermont Yankee informed the Automatic Valve Co. (THE 

AUTOMATIC VALVE CO. IS DESIGNATED BY THEMSELVES AS 

AV, AND IS NOT AVCO AS ERRONEOUSLY NOTED IN MUCH 

OF THE RELATED CORRESPONDENCE), in September 1996, that a 

scram solenoid pilot valve (SSPV) was making a "buzzing" noise. On 

April 24, 1997, during scram time testing, the control rod associated with 

the noisy SSPV had a slower (but within Tech Spec limits) time than the 

other control rods being tested. AV determined that the cause of the slow 

SSPV was the accumulation of debris from the stainless steel plunger 

resulting from the plunger cycling at 60 Hz for some time. Poor fit and 

finish between the plunger and guide, and very small tolerances, were 

thought to be root causes. 

AV supplied about 90 SSPVs to Vermont Yankee, and Vermont Yankee 

is the only plant with AV SSPVs. AV supplied SOVs for BWR MSIVs, 

which they described as similar to the SSPVs, to as many as 25 U.S. 

nuclear power plants. 

A search of the SCSS database using the words "scram solenoid" 

produced 62 LER records, mostly related to slow rod insertion times or 

failures to insert. 

961123 NRC Inspection Report 50-382/96-24 identified an unresolved item 

concerning the closed safety function of certain containment isolation 

valves that receive an open ESF actuation signal and subsequently fail to 

open on loss of air.

b, c

c

Containment isolation criteria required that containment integrity be 

maintained for at least 30 days post-accident. Some containment 

isolation valves are air operated and fail open on loss of air. Since the IA 

system, which supplies the AOVs is non-safety-related, it cannot be 

relied upon post-accident. The AOVs were supplied from safety-related 

air accumulators but the accumulators did not have sufficient capacity to 

ensure that the AOVs would remain closed for 30 days. 

Thirteen AOVs were of this type. One other AOV, CVC-209, charging 

system outside isolation valve, did not have an accumulator nor was it 

equipped with a Class IE SOV.

Scram

z 

tri 

4JContainment
Table 4. (continued). 

Event Report 

Number and 

Other Related Event 

Plant or Vendor Designations' Date Description Classification2 System

z 

7O 

0o 

o' 

tU'

961123 A backup nitrogen supply was to have been provided. 

No LER could be found for this condition. 

This is considered to be a common-cause failure condition.

Letter 

WF3-97-0107 

dated May 6, 

1997 (NRC 

Accession 

Number 

9705120329) 

Event 

No. 34237 and 

MR Number 

4-98-0028 that 

updated the 

Event Report 

LER 

38298010 also 

refers to this 

event

C

C

H" 

0o

Containment

ECCS, MFW

The licensee addressed this condition by opening a normally shut manual 

isolation valve to place a relief valve with a capacity of 1419 scfm in 

service, thus providing a relieving capacity of 1419 + 170 = 1589 scfm, 

which is greater than the 879 scfm capacity of the pressure regulator. 

a. The NRC numbers are 5-digit sequential numbers assigned by the NRC. These events are usually reported in response to the requirements described in 10 CFR 50.72. 

b Events in the table are classified as follows: 

a. Failures of AOVs or events resulting from contaminated (not clean and dry) air; 

b. Failures of AOVs or events resulting from solenoid failures; 

c. Failures of AOVs or events resulting from design deficiencies, material failure, or insufficient margin; and/or 

d. Failure of AO damper(s) or events related to damper failures..

980514 Failure of the main nitrogen regulating valve, NG147, a non-safety

related valve, could have affected both trains of several safety-related 

systems. The configuration of the nitrogen system includes a pressure 

regulator installed at the outlet of the liquid nitrogen tank which reduces 

pressure from 2300 to 750 psig at a flow rate of 879 scfm. The design 

system pressure rating downstream of the pressure regulator was 

800 psig with a relief valve installed at a setpoint of 940 psig, at a flow 

rate capacity of 170 scfm. 

The nitrogen system provides motive force for operation of valves in the 

ECCS, EFW, and other systems. Failure of the pressure regulator could 

have over pressurized the downstream piping and thus, could have either 

interrupted pneumatic power to the AOVs if the piping failed, or result in 

damage to the AOVs if the piping did not fail. This is considered to be a 

common-cause failure condition.

Waterford 3

Waterford 3

0T 

00
Table 5. Licensee Event Reports cited in Tables 2, 3, and 7 listed in numerical order. 

Plant LER Number' Event Date 

Haddam Neck 21393005 930518

Haddam Neck 

Haddam Neck 

Haddam Neck 

Haddam Neck 

Oyster Creek I 

Nine Mile Point 1 

Dresden 2 

Dresden 2 

' Dresden 2 

Indian Point 2 

Indian Point 2 

Dresden 3 

Dresden 3 

Turkey Point 3 

C__ and 4 

Turkey Point 3 

Quad Cities 1

21393007 

21394005 

21396012 

21396018 

21985012 

22096004 

23787023 

23788012 

23798003 (Event No. 33620, dated 1/28/98 

followed up by Morning Report H-98-0045, 

dated 3/6/98 referred to this condition) 

24788017 

24793010

Classification' 

a, b, c

930525 a,c

940219 

960611 

960822 

850612 

960520 

870717 

880517 

980128 

881102 

930818

c 

c 

c 

c 

c 

C 

c 

c 

c 

c

24993004 930116 b,c

24993005 

25085020 

25085021 

25492028

930126 

850723

c 

c

850722 a 

921016 c

System 

Instrument Air 

Reactor Coolant Letdown 

Instrument Air 

Pressurizer Pilot-Operated Relief Valves 

Pressurizer Pilot-Operated Relief Valves 

Main Feedwater 

Main Feedwater 

Scram Discharge Volume 

Main Feedwater 

Main Feedwater 

Main Steam Isolation Valves 

HPCI 

EDG Ventilation System 

Main Steam Power Operated Relief Valves 

Auxiliary Feedwater System 

Scram Header 

Instrument Air 

Drywell Environmental System 

Main Steam Isolation Valves

Auxiliary Feedwater 

Instrument Air 

Control Room
Table 5. (continued) 

Plant 

Palisades 2 

Palisades 2 

Palisades 2

Palisades 

Palisades 

Palisades 

Palisades 

Browns Ferry 2 

Robinson 

. Point Beach 1 

0 

Point Beach I 

Oconee 2 

Vermont Yankee 

Salem I 

Salem I 

Peach Bottom 3

Prairie Island I 

Fort Calhoun

LER Numbera

5578003 

5581030 

5587018

25592007 

25592023 

25594004

25598006 

26095008 

26194002 

26697014 

26698008 

27093002 

27198025, Revision 2 

27291030

Event Date 

780108 

810718 

870620

920205 

920327 

940209 

980305 

951004 

940425 

970321 

980203 

930610 

981211 

910920 

960723 

910924

27296012 

27891017 

28295013 

28587018

Z 

40

950927 c 

870504 c 

Fort Calhoun 28587025 870923 a

Classification' 

a 

a 

a 

b, c 

b, c 

c 

C 

b, c 

c 

C 

b, c 

c 

c 

c 

c 

Personnel 

Error, b

Control Room Chilled Water System 

Instrument Air 

Containment Spray High Pressure Safety 

Injection 

Low Pressure Safety Injection 

Diesel Generator Cooling System

System 

Shutdown Cooling (RHR) 

Shutdown Cooling (RHR) 

Main Feedwater 

Instrument Air 

Main Steam Isolation Valve 

High Pressure Safety Injection 

Engineered Safeguards Systems (HPSI, 

LPSI, CS, CCW) 

HP Air System 

Containment 

Main Steam Isolation Valves 

Auxiliary Feedwater 

EDGs 

RCP Seal Return 

Control Rod Drive 

Pressurizer Pressure Operated Relief 

Valves 

RHR 

Main Steam Relief Valve
Table 5. (continued)

Plant 

Fort Calhoun 

Fort Calhoun 

Fort Calhoun 

Fort Calhoun 

Fort Calhoun 

Fort Calhoun 

Indian Point 3 

Indian Point 3 

Indian Point 3 

Indian Point 3 

Indian Point 3 

Indian Point 3 

Indian Point 3 

Indian Point 3 

Indian Point 3 

Indian Point 3 

Indian Point 3

61 

70 

a"

LER Numbera Event Date

28587033 

28588002 

28588004 

28588009 

28588028 

28590025 

28688009 

28693013 

28693035 

28693036 

28693045 

28693050 

28693053 

28696002 

28696004 

28696008 

28699002

870706

880125 

880311 

880406 

881020 

900929 

881025 

930414 

930916 

930915

Event Date

931023 c

931115 

931202 

960120 

960215 

960320 

990122

b, c 

b, c 

a 

c 

c 

c

Classification' 

a, c 

c 

c 

c 

a 

c 

b, c 

a 

c 

c

,..

System 

Instrument Air 

Fire Protection System 

High Pressure Safety Injection 

Instrument Air 

High Pressure Safety Injection 

Component Cooling Water 

Main Steam Isolation Valves 

Component Cooling Water 

Raw Water 

Containment Spray System 

Reactor Coolant System 

Liquid Waste 

Central Control Room (CCR) Ventilation 

System 

WCCPPPS, IA and SA 

Central Control Room (CCR) Ventilation 

System 

Central Control Room (CCR) Ventilation 

System 

Service Water System 

Emergency Diesel Generators 

Emergency Diesel Generators 

Containment Isolation Diaphragm Valves 

Isolation Valve Seal Water System 

Containment
Calvert Cliffs 

Calvert Cliffs 

Calvert Cliffs

1 

1

Calvert Cliffs I 

Hatch I 

Sequoyah I

Table 5. (continued) 

z 

Plant 

S Oconee 3 2 

C0 

~ Three Mile Island 1 2 

Pilgrim 2 

Pilgrim 2 

Pilgrim 2 

Pilgrim 2 

Cooper 2 

Crystal River 3 3 

Maine Yankee 3 

2 3 

Cook 1 3 

D. C. Cook 3

31789018 

32192003 

32792018

LER Number' 

8791007 

8986007 

9389002 

9389004 

.9397025 (Event No. 33360 refers.) 

.9397026 (Event No. 33360 refers.) 

9894013 

0297015 

0996003 

1197002(Event No. 31783 refers.) 

1597023(See Event No. 32939) 

v1597026 

1788009 

31789005 

31789018

Sequoyah 1 32797012

Event Date 

910703 

860326 

890110 

890127 

971123 

971206 

940712 

970612 

960213 

970213 

970916 

970925 

880824 

890314 

891106 

891106 

920102 

921026

Classification' 

a, b 

c 

c 

c 

c 

c 

c 

c 

C 

b, c 

c 

c 

c 

c 

c 

c 

b, c 

a

System 

Main Feedwater 

Condensate System 

Instrument Air 

Two Hour Backup Air Supply System, 

EFW, Atmosph. Dump Valves 

Containment Isolation System 

Containment Isolation System 

MSIVs, Containment 

Main Feedwater 

RCS, Core Spray 

Containment 

Main Feedwater 

Auxiliary Building Ventilation 

ESF Ventilation 

RHR HX Outlet Valves, Steam Generator 

PORVs 

Main Feedwater 

Instrument Air 

6 Different Systems 

Instrument Air 

Instrument Air 

RHR Core Spray 

Main Feedwater 

Essential Air 

Nonessential Air System 

Control and Service Air System 970801
Table 5. (continued) 

Plant

Duane Arnold 

Fitzpatrick 

Beaver Valley I 

Millstone 2 

Millstone 2 

Millstone 2 

Fermi 2 

Fermi 2 

Davis Besse 

Davis Besse 

Hope Creek 

Hope Creek 

Hope Creek 

San Onofre 2 

San Onofre 2 

(and 3) 

LaSalle 1 

LaSalle I

LER Numbera Event Date

33191005 

33389004 

33490007 

33689011 

33697011 

33698019 

34194004 

34197007 

34687015 

34688007 

35486063 

35494017 

35497020 

36196011 

36199003 

37385008 

37385011

910622 

890309

Classificationb

C 

C

900330 a,c

890906 b,c, 

Personnel 

Error

970402 c

980826 

940822 

970325 

871207 

880304 

860828 

941110 

970828 

961216 

990210 

850202 

850202

c 

c 

b 

b, c 

c 

b, c 

c 

c 

c 

c 

b 

b

System 

Main Steam Isolation Valves 

Instrument Air 

Electrical Cooling System 

Main Feedwater 

Instrument Air 

Turbine Building Closed Cooling Water 

Instrument Air

Containment, CVCS, RC Sample System, 

Liquid Radwaste System 

Auxiliary Feedwater 

Control Rod Drive Hydraulic System 

Reactor Coolant System 

Reactor Building Ventilation System 

Main Steam 

Main Feedwater 

Instrument Air 

Decay Heat Removal)Instrument Air 

10 Different Systems 

Instrument Air 

Diesel Denerator Room Cooling 

Safety Auxiliary Cooling 

PCIS 

Component Cooling Water 

Reactor Building Ventilation System 

Reactor Building Ventilation System

-4

z 

m 

0*

CD 

VA

Classific ion'
z 

trl

Table 5. (continued 

Plant 

LaSalle 1 

LaSalle I 

LaSalle 1

LaSalle 2 

LaSalle 2 

Summer 

Summer 

Shearon Harris 

-a Shearon Harris 

Catawba 1 

Millstone 3 

Millstone 3 

Millstone 3 

Millstone 3 

Millstone 3 

Millstone 3 

Perry 

Perry 

Comanche Peak 

Comanche Peak 

River Bend

1 

1

)

LER Number'

37389007 

37396011 

37397046 and 37398007(Event No. 33434 also 

refers.) 

37489018 

37495005 

39587019 

39598009 

40098001(Event No. 33513 refers.) 

40098001 (continued) 

41397002 

42393002 

42396013 

42396028 

42396031 

42396036 

42396040 

44087009 

44090021 

44595005 

44598001 

45889022

Event Date 

890210 

960928 

971216 and 

980319 

891216 

950218 

870730 

981006 

980109 

981009 

970506 

930207 

960515 

960916 

960906 

960926 

961024 

870227 

900709 

950831 

980110 

890502

Classificationb 

b 

C 

b, c

C 

b, c 

C 

C 

c 

C 

c 

C 

c C 

c 

b, c 

c 

c 

b, c 

b, c 

c 

c 

c

System 

Control Room HVAC System 

Containment Isolation System 

Reactor Core Isolation Cooling 

Reactor Building and Auxiliary Building

Division I Battery Room Heat Removal 

System 

Main Steam Isolation Valves 

Instrument Air 

Control Room Ventilation 

Emergency Feedwater 

Steam Generator 

Steam Generator 

Feedwater Containment Isolation Valve 

Control Room Pressurization System 

Residual Heat Removal System 

Charging System 

13 Different Systems 

High and Low Pressure Safety Injection 

Reactor Plant Component Cooling Water 

Emergency Diesel Generators 

Main Steam 

Pressurizer Power Operated Relief Valves 

Feedwater 

Instrument Air

H
Table 5. (continued) 

Plant LER Number' Event Date Classification' System

River Bend 

Clinton 

Clinton 

Palo Verde I 

Palo Verde I 

Palo Verde 1

45889024 890406 b,c

46190004 

46198009 

52889005 

52894009 

52895007

900703 

980203 

890412 

941118 

950512

Control Building HVAC 

Auxiliary Building HVAC 

Fuel Building HVAC 

Misc. Active Safety-related Systems 

EMDs

b, c 

b, c

c 

c 

c

Containment 

Main Steam 

Containment 

Containment

a. The NRC Licensee Event Report (LER) Number consists of the three-digit NRC Docket Number for the plant at which the event occurred, the 

last two digits of the year in which the LER was generated, and a three-digit sequential number of the LER. This is consistent with the NRC's 

Sequence Code Search System (SCSS) database designation. 

b. Events in the table are classified as follows: 

a. Failures of AOVs or events resulting from contaminated (not clean and dry) air; 

b. Failures of AOVs or events resulting from solenoid failures; and/or 

c. Failures of AOVs or events resulting from design deficiencies, material failure, or insufficient margin.

z 

61

cr

CA

(A
Table 6. Air-operated valves considered by the licensees in the plants visited to be risk 

C" W significant. 

Safety Category 

(Sr = Safety Related, Risk Ranking or 

Plant Or Site AOV Description and/or No. of Nsr = Non-Safety- Significance 

(See Note 1) Designation AOVs Related) (See Notes) Remarks

LA 

4ý1

Palo Verde 4 Atmospheric Dump Valves per 

unit. SGA-HV0178, 0179, 0184, 

and 0185 

Palo Verde 4 Feedwater Isolation Valves per 

unit. SGA-UVO130, 0135, 0172 and 

0175 

Palo Verde 2 Steam Generator Isolation Valves 

SGA-0500P and Q 

Palo Verde 4 Valves SGN-UV170, 171, 180 

and 181 

Palo Verde 2 Valves SGN-FV 1113 and 14 

Fermi 2 8 MSIVs B2103F022A, B, C, and 

D, F026 A, B, C, and D 

Fermi 2 4 Scram Discharge Vol. Vent and 

Drain Valves C1 1OOFO10, 011, 180, 

and 181

Passive SR 

Active SR

Active SR

Active SR 

Active SR 

Passive SR

SR

SR

High

High for common

cause failures 

High for common

cause failures 

High for common

cause failures 

High for common

cause failures 

High for common

cause failures 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided

Charging System Valve 

CHE-HV0532

Palo Verde 3 See Note 2. Only single valve 

listed as having "high risk 

significance." 

See Note 2. See LER 52889005 

for a description of the 

common-cause failure of these 

AOVs. 

See Note 2. These AOVs were 

analyzed by the licensee 

because of common cause 

failures attributed to lack of 

design margin. 

See Note 2. 

See Note 2. 

See Note 2. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel.

12

-1 

0%

12

6

12

6 

8

4
Table 6. (continued). 

Safety Category 

(Sr = Safety Related, Risk Ranking or 

Plant Or Site AOV Description and/or No. of Nsr = Non-Safety- Significance 

(See Note 1) Designation AOVs Related) (See Notes) Remarks

Fermi 2 

Fermi 2 

Fermi 2 

Fermi 2 

Fermi 2 

Fermi 2

370 

(Not 

added to 

total) 

1

1

SR

SR

SR

370 Control Rod Drive Scram Inlet 

and Outlet Valves C1 103D 

1 Drywell Floor Drain Area Sump 

Pumps C001A and B to Floor Drain 

Collection Tank CIV GI 100F003 

1 Drywell Floor Drain Area Sump 

Pumps C006A and B to Floor Drain 

Collection Tank CIV G1 100F003 

1 Pressure Control Valve N20F-400 

(condensate polishing demineralizer 

differential pressure control valve) 

1 Level Control Valve N20F-406 

(condensate condenser low-level 

make-up level control valve 

emergency supply) 

1 Level Control Valve N21F-403 

(level control valve: reactor feed 

pump start-up control V12-2512)

NSR

1 NSR

1 NSR

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel.

1

z 

6N H
Table 6. (continued). 

Safety Category 

(Sr = Safety Related, Risk Ranking or 

Plant Or Site AOV Description and/or No. of Nsr = Non-Safety- Significance 

(See Note 1) Designation AOVs Related) (See Notes) Remarks

tA 

-P

1

1

Fermi 2 1 Temperature Control Valve 

P43F402 (general service water 

from turbine building CCW heat 

exchanger) 

1 Temperature Control Valve 

P44F400A (temperature control 

valve EECW heat exchanger 

P4400B001 service outlet) 

1 Temperature Control Valve 

P44F400B 

2 Division 2 Pump Discharge 

Primary Containment Valves EESW 

P45F400A and B 

4 EDG Air Coolant System 3-Way 

Temperature Control Valves 

(R3000F023A, B, C, and D) 

1 Standby Gas Treatment Supply 

Chamber Purge Isolation 

ValveT4600F400

1

NSR

SR

SR 

SR

SR

NSR

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel.Note: During 

the visit we were told that this 

valve had a "surprisingly high" 

calculated risk significance. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel.

1 

1

1

Fermi 2

H 

CD

Fermi 2

00

Fermi 2 

Fermi 2 

Fermi 2
Table 6. (continued). 

Safety Category 

(Sr = Safety Related, Risk Ranking or 

Plant Or Site AOV Description and/or No. of Nsr = Non-Safety- Significance 

(See Note 1) Designation AOVs Related) (See Notes) Remarks

Fermi 2 

Fermi 2 

Fermi 2 

Fermi 2 

Fermi 2 

Palisades

1

1

1

1 Reactor Building HVAC Exhaust 

System Isolation Valve T4600F407 

1 Secondary Containment to 

Division 2 Standby Gas Treatment 

Isolation Valve T4600F408 

1 Standby Gas Treatment 

Secondary Containment to Division 

1 Standby Gas Treatment Isolation 

Valve T4600F409 

1 Standby Gas Treatment to Torus 

Air Purge Valve T4600F412 

2 Division 1 Torus Hard Pipe Vent 

Secondary Control Isolation Valves 

T4600F421 and 422 

1 LPSI Shutdown Cooling Heat 

Exchanger Bypass Valve CV-3006 

1 Condensate Inlet Containment 

Isolation Valve CV-2010

2

1

NSR

NSR

NSR

NSR

NSR

SR

1 SR

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

Listed as risk 

significant but no 

ranking provided 

High safety 

significance 

High safety 

significance

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 3. From tabulation of 

"AOVs That Are Risk 

Significant" as determined by 

an Expert Panel. 

See Note 4. From a tabulation of 

active AOVs modeled in the 

PSA. No numerical values or 

ranking were provided. 

See Note 4. From a tabulation of 

active AOVs modeled in the 

PSA. No numerical values or 

ranking were provided.

1

z

Palisades

H
Table 6. (continued). 

Safety Category 

(Sr = Safety Related, Risk Ranking or 

Plant Or Site AOV Description and/or No. of Nsr = Non-Safety- Significance 

(See Note 1) Designation AOVs Related) (See Notes) Remarks

ON SR 

SR

SR

Palisades 1 Normal Steam toP-8B from Steam 1 

Generator A Valve CV-0522B 

1 Shutdown Cooling to LPSI 1 

Isolation Valve CV-3025 

2 Containment Sump Isolation to 2 

East Engineered Safeguards Room 

Valves CV-3029 and 3030 

4 Steam Generator E-50B Steam 4 

Dump Control Valves CV-0779, 80, 

81 and 82 

1 Shutdown Cooling Inlet to I 

Shutdown Heat Exchangers CV

3055 

Loss of instrument air accounted for Not 

about 1/3 of CDF. See Note 5 specified 

Six of 7 ADS valves fail to function 7 per unit 

on demand total 

Two Atmospheric Dump Valves for 2 

A and B Steam Generator MS-V

004A and B (6" Fisher globe air 

piston)

See Note 5

SR

SR (Category 1)

High safety 

significance 

High safety 

significance 

High safety 

significance 

High safety 

significance 

High safety 

significance 

See Note 5

RAW = 74.8 F-V = 

0.0167 

F-V = MF-V range 

listed as 0.00136 to 

1.14 RAW = L

See Note 4. From a tabulation of 

active AOVs modeled in the 

PSA. No numerical values or 

ranking were provided. 

See Note 4. From a tabulation of 

active AOVs modeled in the 

PSA. No numerical values or 

ranking were provided. 

See Note 4. From a tabulation of 

active AOVs modeled in the 

PSA. No numerical values or 

ranking were provided. 

See Note 4. From a tabulation of 

active AOVs modeled in the 

PSA. No numerical values or 

ranking were provided. 

See Note 4. From a tabulation of 

active AOVs modeled in the 

PSA. No numerical values or 

ranking were provided. 

No individual rankings were 

provided. See Note 5. 

No individual rankings were 

provided. See Note 5. 

See Note 6.

SR

SR

Palisades 

Palisades

-]

00 

0 Palisades

Palisades

LaSalle 1 

and 2 

LaSalle 1 

and 2 

Three Mile 

Island 1
Table 6. (continued). 

Safety Category 

(Sr = Safety Related, Risk Ranking or 

Plant Or Site AOV Description and/or No. of Nsr = Non-Safety- Significance 

(See Note 1) Designation AOVs Related) (See Notes) Remarks

Three Mile 

Island 1 

Three Mile 

Island 1 

Indian Point 3

00

Indian Point 3

Indian Point 3 

Indian Point 3

Containment Isolation IC Isolation 

coolant return valve IC-V-0003 (6" 

Tufline plug air piston) 

Containment Isolation IC Isolation 

Coolant Supply IC-V-0004 (6" 

Tufline plug air piston) 

8 Auxiliary Feedwater Regulating 

Valves BFD-FCV-405A, B, C, and 

D and BFD-FCV-406A, B, C, 

and D (2" Copes-Vulcan D100 

diaphragm AOVs) 

1 Main Steam to Auxiliary Boiler 

Feedwater Pump MS-PCV-i 139 

(2.5" WKM globe valve)

2 Condensate Storage Tank to 

Condensers Level Control Valve 

CT-LCV-1158-1 and -2 (12" 

Golden Anderson Butterfly)

1 

1

8

I

1

1 Condensate Polisher Facility Inlet 1 

Stop CD-AOV-518 (Cameron Iron 

Works Ball Valve)

SR (Category 1) 

SR (Category 1)

SR

SR

SR

NSR

RAW = HF-V = 

MF-V range listed 

as 0.00116 to 24.2 

RAW = HF-V = 

MF-V range listed 

as 0.00116 to 24.2 

"High PRA"

"High PRA"

"High PRA"

"High PRA"

See Note 6. 

See Note 6.

See Note 7.(RAW = 202.000 

associated with BFD-FCV-406 

AOVs for CCF of AFW motor

driven pumps.) (RAW = 2.431 

for CCF of all 4 PM32 FCVs to 

open.)(RAW = 2.047 for any 

one AOV to close on demand.) 

See Note 7.(RAW = 5.956 for 

Steam Control Valve, PCV

1139 to not open.) (RAW = 

9.649 associated with AFW TD 

Pump #32 fails to continue to 

run.) 

See Note 7.(RAW = 1.141 for 

either valve to not close.) 

See Note 7.(No RAW value 

provided.)

-3
Table 6. (continued). 

Safety Category 

(Sr = Safety Related, Risk Ranking or 

Plant Or Site AOV Description and/or No. of Nsr = Non-Safety- Significance 

(See Note 1) Designation AOVs Related) (See Notes) Remarks

N 

0i 

03 Heater Drain Tank to Condenser 

Bypass HD-LCV-7001, 2, and 3 

(4" Masoneilan globe valve) 

4 Steam Generator Atmospheric 

Dump MS-PCV-1134, 5, 6, and 7 

(6" Copes-Vulcan D-100 globe 

valve) 

2 Pressurizer PORVs RC-PCV

455C and 456 (3" Copes-Vulcan F 

series globe valve)

Indian Point 3 MSIV 1-31, 1-32, 1-33, and 1-34

Indian Point 3 

Indian Point 3

EDG Flow Control Valves FCV

1176 and 1176A 

Steam Generator Blowdown AOVs, 

BD-PCV-1214, 1214A, 1215, 

1215A, 1216, 1216A, 1217, and 

1217A

3

4

NSR 

SR 

2 SR

4 SR

2 Not provided

8 SR

"High PRA" 

"High PRA" 

"High PRA"

Each individual 

valve is listed by the 

licensee as "No" in 

the column headed 

"High PRA" 

Not provided 

Each individual 

valve is listed by the 

licensee as "No" in 

the column headed 

"High PRA"

Indian Point 3 

Indian Point 3 

Indian Point 3

See Note 7.(No RAW value 

provided.) 

See Note 7.(RAW = 1.613 for 

any of the 4 valves to fail to 

close.) 

See Note 7.(RAW = 8.217 for 

PCV-455C to fail to close.) 

(RAW = 7.476 for PCV-456 to 

fail to close.) (RAW = 5.666 for 

PCV-455C does not 

open.)(RAW = 6.306 for PCV

456 does not open.) (RAW = 

6.229 for CCF of PORVs to 

open.) 

See Note 7.(RAW = 50.73 

associated with CCF of two or 

more MSIVs.) (RAW = 1.614 

for failure of each MSIV to 

close on demand.) 

See Note 7.(RAW = 46.97 

associated with CCF of EDG 

flow control valves.) 

See Note 7.(RAW = 3.458 

associated with CCF of 2 SG 

Blowdown valves.)

H

00 

00
Table 6. (continued). 

Safety Category 

(Sr = Safety Related, Risk Ranking or 

Plant Or Site AOV Description and/or No. of Nsr = Non-Safety- Significance 

(See Note 1) Designation AOVs Related) (See Notes) Remarks

Indian Point 3 

Turkey Point 

Turkey Point 

Turkey Point 

Turkey Point 

Turkey Point 

Turkey Point 

Turkey Point 

Turkey Point 

Turkey Point

Solenoid valves VS-SOV-1297, 

1298, 1303, 1304, 1306, and 1307 

3 AOVs FCV-3-6278A, B, and C, 

Steam Generator Blowdown 

Control Valves 

3 AOVs CV-3-6275A, B, and C 

Steam Generator Blowdown 

Control Valves 

3 AOVs CV-3-2816, 17, and 18 

Control Valve for water to steam 

generator from auxiliary feedwater 

pump 

3 AOVs CV-3-2831, 32, and 33 

Control Valve for water to steam 

generator from auxiliary feedwater 

pump 

AOV LCV-3-113B 

AOV LCV-3-114A 

AOV LCV-3-115B 

3 AOVs CV-3-2903, 4, and 5 

AOV CV-4-1605

6 Not provided

6

6

6

6

1 

1

1 

3

3 

182

SR

SR

SR

SR

SR 

SR 

SR 

SR 

SR

Not provided 

Risk significant 

Risk significant 

Risk significant 

Risk significant

Risk significant 

Risk significant 

Risk significant 

Risk significant 

Risk significant

See Note 7. (RAW = 1.035 for 

any SOV's failure to function.) 

See Note 8. The Expert Panel 

considered these valves to be 

not risk significant because 

manual valves are available as 

backup. 

See Note 8. The Expert Panel 

considered these valves to be 

not risk significant because 

manual valves are available as 

backup. 

See Note 8. 

See Note 8. 

See Note 8. 

See Note 8. 

See Note 8. 

See Note 8. 

See Note 8.

(Excluding 370 Control Rod Drive Scram Inlet and Outlet Valves)

00

CD

Total No. of AOVs
Table 6. (continued). 

z 

ýO Notes: 

I . Plants are listed in the order of the site visits for the AOV study. There are three units at Palo Verde, two at LaSalle, and two at Turkey 

Point. (Note: At Turkey Point some systems, e.g. AFW, are shared.) 

2. PRA input to the Palo Verde AOV program was provided in 1994 and described 27 AOVs (per unit) as having some impact on core 

damage. The AOVs were tabulated as "High Risk Significant," "High Risk Significant for Common Mode Failures," or "Low Risk 

Significant." No risk numbers or rankings within the categories were provided. The "Low Risk Significant" AOVs (10 per unit) are not 

included in the table above. We have since been told that the information may be obsolete but that no later information is available. 

3. Fermi 2 provided a Table of "AOVs That Are Risk Significant." The table listed (according to the summary provided with it) 370 control 

rod drive AOVs, 22 AOVs "that perform safety-related functions," and 11 AOVs "that perform a non-safety related risk-significant 

function." All of these AOVs were designated as "Category 1 (high safety significance)." Category I was defined as follows: 

- "AOVs falling in this category are those that play an active role within systems that fall within the scope of the Maintenance Rule and 

that were considered to be risk significant by the Fermi Plant expert panel. Valves in this category are candidates for activities such as 

review of the design basis and setpoint verification confirmed by periodic diagnostic testing." 

00 

- "Category 2 (less safety significance but having potentially significant economic consequences)" was defined as follows: 

- "Valves in this category may be either safety-related or are not safety-related. These valves either play an active safety-related 

function of low risk significance or are used in applications that may affect plant availability, capacity factor, or heat rate. AOVs in 

this category would be subject to design reviews and/or testing on an as-needed basis at the discretion of the plant staff." 

- "Category 3 (less safety significant and having little or no plant economic impact)" was defined as follows: 

- "AOVs in this category may or may not be safety-related, have been determined to be of less safety significance playing no active 

safety-related role and do not have a significant potential for affecting plant operation. Design basis review or diagnostic testing 

would not be performed for valves in this category." 

4. Three safety significance ratings were described in the Palisades PSA review of AOVs as follows: 

- Non-safety significance rated valves were not modeled in the PSA or were not modeled as active components.

- Low safety significance rated valves were those modeled in the PSA but not having high importance measures.
Table 6. (continued). 

- High safety significance rated valves were modeled in the PSA and had high importance measures. High importance measures were 

defined as Fussell-Vesely (F-V) >5E-3 or Birnbaum>5.15E-5. These importance measures were considered by the licensee to 

correspond to Risk-Achievement Worths (RAW) >2.0 and Risk-Reduction Worths (RRW) >1.005 respectively. 

5. At LaSalle, no tabulation was provided during the site visit about rankings for AOVs on the basis of risk significance. Two tables of interest 

here were provided (5a and 5b) in the LaSalle PRA Summary Document. These tables show Component Failure Mode Importances to CDF 

for RAW > 2 and Fussell-Vesely, respectively, for many different components, including AOVs. The components that could be positively 

identified from the descriptions in these tables as AOVs were included in the table above, along with the values of the importance measures 

listed. 

NOTE: A list of the top 100 core damage cutsets was included in the LaSalle PRA Summary Document. Transients with loss of instrument 

air were the largest initiating event category at LaSalle station and contributed 32% of the core-damage frequency. This risk number is 

noted as the first entry in the table above for LaSalle because it is considered an exceptionally high value and therefore pertinent to the study 

of AOVs and their risk significance. 

6. Two documents were provided during the site visit to Three Mile Island (TMI) I that provided data and information on risk-significant 

AOVs. These were: 00 

1. "Draft AOV Program Description, Topical Report 118," (not dated) categorized AOVs as follows: 

- Category I included AOVs that are required to perform a safety function are called upon during a design basis event. 

Category I AOVs are "...further categorized by an expert panel based on the relationship between the valve's risk ranking 

(level of safety significance) and amount of available operating margin. Category IA are High Safety Significant valves or 

valves that have Low Safety Significance with low operating margin. Category lB valves are those with Low Safety 

Significance and medium to high levels of operating margin." 

- Category 2 AOVs were classified as AOVs "...selected by an expert panel which provide significant assistance to the operation 

of safety systems as delineated in TMI Emergency Operating Procedures. Category 2 valves include valves selected by the 

Expert Panel that may have a significant effect on the safe operation of the plant. These valves might include valves that are 

Z the last line of defense for release of radiation to the environment or may provide emergency boration to the Reactor Coolant 

System. Additionally, AOVs whose failure may result in a plant trip, power reduction or have economic significance on plant 

operation are considered for inclusion in this category." 

- Category 3 AOVs were defined as "...valves that have no active safety role and do not have a significant potential for affecting 

plant operation." 

tJ'
z Table 6. (continued). 

- A list of Category 1 AOVs was included in the program plan (Appendix A thereto) and is summarized in the table above. 

2. A paper entitled "TMI-1 PRA Input to Maintenance Rule Risk Significant System List," developed by C. Adams, Risk Analysis Section 

(not dated), included Fussel-Vesely and Risk Achievement Worth (see Section 15.61 for further explanation of these terms) and RAW data 

ON and rankings for valves including AOVs. (It appears that a portion of the paper was not provided.) Pertinent data was included in the table 

.•1 above. A table of on-line maintenance risk rankings indicated that RAW values of 1 to 3 were considered Low (L), RAW values of 3 to 15 

were considered Medium (M), and RAW values of 15 to 30 were considered High (H). RAW values above 30 were not to be allowed 

without a senior management evaluation and a risk management evaluation. RAW (Risk Achievement Worth was defined as the factor 

increase in core damage frequency resulting when the structure, system, or component (SSC) is determined to be continuously failed. 

Eight AOVs were categorized as L in the reference and these were not included in the table above. Between 64 and 68 (64 by count and 68 

in a summary table) AOVs and SOVs were also listed in the reference as NM (not modeled), N (not ranked) or T (truncated), and these were 

also not included in the table above. 

7. Three lists that included information on the risk significance of AOVs were obtained at the site visit to Indian Point 3. Two of the lists 

(dated 10-Mar-I1998) included sorts of AOVs by system-component and by manufacturer-component and each identified "High PRA" 

AOVs. No ranking of the AOVs was provided in these two lists, nor was there an indication of what the cutoff for "High PRA" might have 

o7 been. The "High PRA" AOVs are included in the table above. 

The third list, attached to NYPA Memo from J. Circle to G. Smith dated November 5, 1996, included RRW, RAW, and unavailability 

estimates for a number of events, including events involving AOV failures. If it was possible to tie the events to a "High PRA" valve, the 

RAW values were listed in the Remarks column of the above table for the particular valve. In addition, several AOVs and SOVs that did not 

appear in the two lists described previously were found to be of interest regarding common-cause failures. These valves were also noted in 

the above table. 

(One item in the third list from IP-3 indicated that the CCF of IA compressors had an associated RAW value of 2.103, but no entry was 

included in the above table because the event did not refer to an AOV.) 

8. An internal memorandum was provided by the Turkey Point engineers that listed 19 risk significant AOVs. These are included in the above 

table, along with whatever data could be gathered for each AOV. In conversations with the risk analysts at Turkey Point on May 17, 1999, 

the risk ranking numbers in the internal memorandum could not be confirmed, and were omitted.
Table 7. Some recent (approximately last 5 years) events and conditions involving AOVs and/or air-operated components where the design basis 

was not met and/or not known. 

NOTE: Items in this table marked in the left margin as shown here were not included in tables of events included in the draft (for comment) report

dae A!{JLkpn! 2-0, 1 "7•. 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Clinton 

46198009 

980203

00 

--.

Comanche Peak 1 

44595005 

950831

ON 

ON

Engineering personnel determined during their review of a 10CFR21 notification by Engine Systems, 

Incorporated, that the air start solenoid pilot valves for the emergency diesel generators (EDG) would 

not operate as required by the design basis. The solenoid valves, which are used to valve air to the air 

operated main valves that supply starting air to the start motors on all three EDGs, would not operate 

reliably at the low end of the design basis air start system pressures and DC voltages. 

The apparent cause of the inadequate design of the solenoid valve was a failure of the design engineer 

to consider the full range of design basis design pressures and DC voltages when changing the spring 

size of the solenoid pilot valve. Corrective action for this event included: changing the spring size in the 

solenoid valve for the Division III EDG air start System; replacing the Division I and II EDG air start 

system solenoid valves, or otherwise modifying the system to meet the design basis requirements; and 

revising the annunciator procedures for the EDG air start system receiver low pressure alarm to require 

that the EDGs be declared inoperable when supply air pressure drops below 200 psig. 

This is considered to be a common-cause failure condition. 

Engineering personnel identified non-conservatism in the calculation that determined (1) leakage rates 

for accumulator check valves associated with various air operated valves and the nitrogen accumulators 

for the pressurizer Power Operated Relief Valves (PORVs), and (2) the pressure switch alarm set points 

for these accumulators. Engineering personnel performed evaluations which revealed that, with the 

exception of the PORVs, the valves associated with these accumulators were operable. The PORV 

accumulator low pressure alarm set points would still ensure operability during Modes 1, 2 and 3. 

However, for Modes 4, 5 and 6 the set points would not ensure operability for all conditions. 

TU Electric believed that the cause of the event was non-conservative design in the original calculation 

used to determine the low pressure alarm set point for the PORV accumulators and the accumulator 

check valve leakage rates.The PORV nitrogen accumulator low pressure alarm set points were to be 

raised to 90 psig. A review was to be performed for all safety-related accumulators used for valve 

actuation to verify that the appropriate set points are being used. 

This is considered to be a common-cause failure condition.

There are 335 SOVs, distributed among about 

17 nuclear plant owners, listed 

in the Part 21 notice. Point 

Beach LER 26698008 also 

refers to Engine Systems 

Part 21 Notification #1998120, 

dated January 26, 1998.

There are two PORVs per pressurizer, thus 4 for the site.

CD
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Cook I 

31597023 

970916 

(The event covered 

by this LER was 

originally reported 

as Event No. 32939)

ý0_ Cook I 0 31597026 

970925

0 

ON 

ON

This event involved dampers 

only (numbers unspecified). No 

AOVs were involved.

A design change to the bypass dampers for the Engineered Safeguards Features (ESF) ventilation 

system installed between December 1996, and August 1997, introduced the possibility of a single 

failure which could result in the loss of both trains of the ESF ventilation system. The loss of the 

85 psig air header without concurrent loss of the 20 psig air header would result in the ESF ventilation 

trains being unable to meet their design function. 

The ESF vent system charcoal inlet and bypass dampers originally both used a 20 psig air header. The 

charcoal bypass dampers were normally open and were intended to fail closed. The charcoal inlet 

dampers were normally closed and intended to fail open. The licensee installed new bypass dampers 

which required higher pressure to operate and were therefore transferred to the 85 psig air header. If 

85 psig air was lost, the bypass dampers would reposition to the closed position. The inlet dampers 

would remain closed and this would result in dead-heading of the filter trains and subsequent loss of 

cooling to ECCS equipment. 

The root cause of the event was the failure of the design change process to identify the potential adverse 

impact on the ESF ventilation system created by the modification of the control air supply to the bypass 

dampers. This is considered to be a common-cause failure condition. 

Due to a lack of overpressure protection on the 85, 50, or 20 psig control air headers, if a non-safety

related air regulator failed open it would result in an overpressurization of a control air header. This 

would result in the potential for common-cause failure of both trains of safety related equipment. The 

lack of overpressure protection on the control air headers due to a regulator failing open had not been 

identified as a mechanism that could overpressurize the low pressure headers. As a result, single failure 

of a non-safety related component affecting both trains of safety-related equipment was not identified. 

There would have been no significant effects for the 85 and 50 psig headers; however, 

overpressurization of the 20 psig header could have resulted in the degradation of the RHR system and 

the partial opening of the Unit 2 Steam Generator (SG) Power Operated Relief Valves (PORVs) for the 

duration of the overpressure event. Due to a single failure being identified that could have potentially 

resulted in the degradation of both trains of RHR, this event could have been significant. 

The cause of the lack of overpressure protection on the Control Air System was the fact that a regulator 

(non-safety-related) failing open was not identified as a mechanism that could overpressurize the low 

pressure headers. As a result, single failure of a non-safety related component affecting both trains of a 

safety related system was not identified.

Unknown. The LER and its 

attached Event Summary refer 

to "multiple" isolation and 

relief valves. The text of the 

LER refers to at least 6 valves 

by number.
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Crystal River 3 

30297015 

970612

00 \0o

Haddam Neck 

21396012 

960611

z 0 

tIl 

0' 

4

During a review of the differential pressure calculation for the Letdown Line [CB] Inboard 

Containment Isolation Valves, FPC discovered that the evaluation of the maximum differential pressure 

(d/P) that these valves could be subject to was in error. These valves are rated to close against a 

maximum d/P of 1800 psi, but could be subjected to a d/P in excess of 2000 psi in the event of a 

letdown line rupture downstream of outboard containment isolation valve (MUV-49), concurrent with a 

failure of MUV-49 to close and operator action in accordance with Emergency Operating Procedure 

(EOP) 3. Outboard containment isolation valve MUV-49 would not be capable of closing if subjected 

to 2000 psi d/P. Isolation at Penetration 333 requires either MUV-49 or MUV-40/41/ 505 to close. The 

cause of this event was the use of inappropriate assumptions in the calculation for the determination of 

maximum valve d/P. FPC planned to install a new inboard containment isolation valve prior to restart 

of CR-3. In addition, the air operator on MUV-49 was to be modified to allow closure against the 

predicted d/P. The closure of these valves is required to mitigate the effects of a Makeup System 

Letdown Line Failure Accident and in response to a Reactor Building Isolation signal. Consequently, 

the valves were outside of their design basis and their failure to close in the described scenario during 

previous operating periods, could have created an non-isolatable loss of reactor coolant to the Auxiliary 

Building. 

This is considered to be a common-cause failure condition. 

An engineering analysis determined that the main feedwater regulating valves (FRV) would not fully 

isolate feedwater flow as required for a main steam line break accident in containment. It was 

determined that the differential pressure across the valves would overcome the valve spring's closing 

forces. In a design basis steam line break analysis, the feedwater motor operated valves (MOV) are 

required to isolate; however, in the event of a single failure of the MOV the FRV is credited with 

isolation. The failure to isolate feedwater for a steam line break in containment could result in 

exceeding maximum containment design conditions. The cause of this condition was an erroneous 

assumption in the accident analysis that the FRVs would isolate against a high differential pressure. The 

differential pressure across the valve would overcome the valve spring's closing force. 

This is considered to be a common-cause failure condition. See LER 21396018. 

(Event No. 30619 dated 6/11/96 applies.)

The summary of the event in 

the LER indicates 8 valves per 

unit were involved. It was 

assumed that these were AOVs.

LER text indicated that 4 AOVs 

were involved.

C,-
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Haddam Neck 

21396018 

960822

SIndian Point 3 

C 28696004 

960215

zý

LER text indicated that 4 AOVs 

were involved.

An engineering analysis revealed that the feedwater regulating bypass valves would not fully isolate 

feedwater flow as required for a main steam line break inside containment. This condition was 

discovered during a follow-up to a similar problem with the main feedwater regulating valves 

(LER 21396012). The failure to isolate feedwater for a steam line break inside containment could result 

in exceeding maximum containment design conditions. This event did not involve any actual equipment 

failures. 

The cause of this condition was an erroneous assumption that the feedwater bypass valves would close 

and isolate against the differential pressure experienced between the steam generator feed pump and a 

faulted steam generator. Additionally, credit was taken for isolation of the bypass valves from the 

control room ten minutes after an accident, however the control air system which operates the valves is 

not a credited system. The bypass valves (FW-HICV-1301-1,2,3,4) are 1-1/2 inch, air to close, spring to 

open, manufactured by Masoneilan. They are normally closed during full power operation. On an 

auxiliary feedwater actuation signal they go full open until manual operator action is taken at the main 

control board to throttle flow. 

This is considered to be a common-cause failure condition. See LER 21396012. 

Two air-operated vapor containment isolation diaphragm valves in series were found to be inoperable, 

which violated Technical Specification 3.6.A. I. The original isolation valve design would close at 

system pressure when there was a differential pressure or accident design pressure, but not against 

system pressure with no pressure differential. The design and specification of the valves during initial 

design and construction was not adequate to ensure the valves met their design requirement for 

containment isolation. 

RC-AOV-519 and RC-AOV-552 (the valves are designed to positively seal with a differential pressure) 

would close against a differential pressure of 150 psi (the Primary Water System design pressure) but 

would not have closed with a line pressure greater than about 120 psig when there is no differential 

pressure. The original specification for these valves was for a maximum differential pressure of 200 psi, 

with no reference to a minimum pressure differential or constant line pressure requirement. The valve 

design had not been modified since the original plant design and construction. The condition could have 

existed when closing isolation valve RC-AOV-560, downstream and in series with RC-AOV-519 and 

552, in a sequence that maintained a line pressure greater than 12 psi. There would then be no 

differential pressure during stroke testing. Stroke testing in these conditions created another problem if 

the limit switches are adjusted after the valves are shut but do not fully close. The valves were assumed 

to be fully closed during the adjustment so subsequent failures to close would be masked and could 

allow a control room indication that the valves were fully closed when they were not. There was no 

direct external indication on the valves to show if they were fully closed. 

This is considered to be a common-cause failure condition.

2 AOVs.
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Indian Point 3 

28699002 

990122

A design condition that had the potential to place the plant outside its design basis was confirmed to 

exist. If a single failure of containment isolation valve VS-PCV- 1190 to close upon demand occurred, a 

potential containment release path would exist in the Weld Channel and Containment Penetration 

Pressurization System (WCCPPS) {BD) supply/containment atmosphere exhaust line. The WCCPPS 

supplies pressurized air to the spaces between the three in-series containment isolation valves 

(VS-PCV-1 190, -1191 and -1192) in the VC Pressure Relief System. This air is supplied when the 

isolation valves are closed; closure is determined by the closed position limit switches of the isolation 

valves being "made." Removal (i.e., exhaust) of WCCPPS occurs when the isolation valve control 

switches are actuated to initiate operation of the VC Pressure Relief system. Supply of WCCPPS air is 

directed to each of the spaces between the two isolation valve when the associated three-way solenoid 

valve {PSV} (PS-SOV-1280) is de-energized; a I-inch line connects each solenoid valve to its 

corresponding space between isolation valves. Exhaust of the WCCPPS air occurs through the same 

1-inch line back through the solenoid valve's exhaust port to the Primary Auxiliary Building (PAB) 

INF} atmosphere. This alignment and flow path occur when the solenoid valve is energized. As part of 

the initiation of a VC Pressure Relief Operation, PS-SOV- 1280 is energized to allow the WCCPPS air 

in the inter-space between VS-PCV- 1190 and -1191 to exhaust into the Piping Penetration area.

PS-SOV-1280 will remain in this position until valves VS-PCV-1 190 and 1191 are closed. If an 

"- automatic containment isolation signal or a manual isolation signal is created, all three isolation valves 

in the VC Pressure Relief System receive a closure signal. Should inboard isolation valve 

VS-PCV- 1190 (located inside the containment) fail to close under such a condition, an unintended 

containment atmosphere flow path would exist through the 1-inch line between the VS-PCV-1 190/1191 

inter-space and the exhaust port of PS-SOV-1280. PS-SOV-1280 is located in the Piping Penetration 

area of the PAB. This combination of circumstances, although unlikely, would represent a failure to 

meet the single failure criteria as applied to containment isolation. 

The most likely cause of the event is an oversight during the initial design phase of the plant when 

considering single failure. The designers of the VC Pressure Relief System apparently failed to 

recognize that a single failure of the inboard containment isolation valve (VS-PCV-1 190) to close 

during containment pressure relief could prevent the isolation of the vent path from containment into 

the PAB during a postulated event. 

z

One AOV.
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

LaSalle 137397046 

and 37398007 

971216 and 980319 

(Event No. 33 34 

also referred)

z 

O4 

C0 

61I

H 

0

0

LaSalle 1 36 AOVs (18 per unit).

37396011 

960928

While developing an Air-operated Valves (AOV) preventative maintenance program, inconsistent 

testing data were obtained for valves with WKM 70-13-1 pneumatic actuators. The inconsistent results 

appeared to be related to incorrect effective diaphragm areas (EDA) for the AOV actuators. 36 WKM 

AOVs were addressed in the LER (18 per unit). 13 AOVs in each unit are part of the primary 

containment isolation system (PCIS) and 5 are in the reactor core isolation cooling system (RCIC). 

Two problems associated with the EDA of the actuators of the WKM valves were identified. The first 

was related to the actual versus the manufacturer's published EDA of the actuator. If the actual EDA is 

less than what the manufacturer publishes, then the closing (or opening) forces installed in the valve 

(via spring/spring adjustment) will be less than required. The second problem was stretching of the 

diaphragm during valve travel resulting in a reduced EDA. 

This is considered to be a common-cause failure condition. 

(LER 37398007) An investigation of the allowable closure time assumptions for the turbine building 

high-energy line break check dampers (#1/2VT79YA/B/C) led the licensee to reconsider the original 

builders assumptions and calculations for the reactor building exhaust isolation dampers (I VR05YA 

and B). 

(LER 37397046) In the event of a Main Steam High Energy Line Break, the Turbine Building 

Ventilation isolation dampers, 1(2)VTO79YA, B & C, would not close fast enough to prevent the 

pressure from exceeding the pressure retaining capability of the walls, floors, and ceilings that separates 

the VT exhaust tunnel from the safety related High Pressure Core Spray (HPCS) electrical switchgear 

room. The apparent causes included an invalid calculation assumption dating from the original design. 

Dampers IVR05YA and B had malfunctioned and were rebuilt in 1985 (see LER 37385008 and 

LER 37385011). The licensee indicated that they revised the calculations and assumed a 0.4 second 

instrument time delay, and a 0.075 second solenoid valve response time. The plant was licensed in 

1982.

No AOVs. Four air-operated 

dampers (2 per unit).
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event 

Millstone 2 The closing force for multiple dual function (two separate pressure isolation functions) valves had been Eleven of 23 AOVs. (Event 

33697011 improperly set, resulting in the valves being incapable of closing to a leak tight condition against No. 32070 reported 19 of 23 

970402 normal operating system pressure (NOSP). Eleven of the 23 valves tested were not capable of providing AOVs.) 

an adequate closing force. This deficiency could have resulted in the potential for a release of 

radioactive materials to the Auxiliary Building greater than analyzed in the facility Final Safety 

Analysis Report (FSAR). The closing forces were incorrectly set during the period between October 

1986 and March 1997. 

The cause of this event was an insufficient program to ensure that facility procedures clearly addressed 

all related design basis functions. The affected valves were to have been adjusted to ensure they 

properly close against containment design pressure and NOSP, and the appropriate procedures were to 

have been revised to ensure that proper valve control parameters are specified and verified after any 

maintenance activities are performed that could affect dual function valve closing forces. 

This condition affected 11 valves in three systems. This is considered to be a common-cause failure 

condition. 

Event No. 32070 referred to this event. 

Millstone 2 In 1981, the AFW regulating valves were changed from MOVs to AOVs to provide electrical Two AOVs. 

3369801 independence from a failure which could preclude the valves from opening and supplying AFW to the 

9980826 steam generators. Both normally closed valves were designed to open by an Auto AFW Initiation 

(AAFWI) signal or fail open upon loss of instrument air. In 1992, non-safety grade backup air was 

installed to provide the ability to close the valves in the event of a beyond design basis main feedwater 

system line break in the turbine building (TB), i.e., a high energy line break (HELB), coincident with a 

feedwater check valve failure. This change was made to improve the unit's core melt frequency, 

however, it was not recognized at that time that more limiting HELBs could affect the TB. The (then) 

current main steam line break (MSLB) analysis assumed operator action to isolate the faulted S/G ten 

(10) minutes after the event. However, closure of the AFW regulating valve and isolation of the S/G 

may not be assured. On September 16, 1998, a NRC review also identified that following a HELB in 

the TB, a higher-than-previously analyzed ambient temperature may also have challenged the ability of 

the AFW regulating valve backup air equipment to remotely close and isolate flow. Subsequently, the 

Z licensee identified that other components of the AFW regulating valve also may have had been 

- challenged. 

From 1981 through 1998, when the plant operated in Modes 3 or higher, the capability to manually 

close the AFW regulating valves to isolate flow out the faulted S/G as analyzed, could not be assured 

and a potential for operating outside of the plant's design basis existed. H 

41.
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Millstone 3 

42396013960515

01ý 

Uh

CD

An engineering evaluation determined that a design deficiency in the Residual Heat Removal System Two AOVs.

(RHS) was a condition that was outside the design basis of the plant. A loss of control air supplied from 

the non-safety-related instrument air system could cause the RHS control valves to fail open. If this 

condition occurred during the initial phase of a plant cool down, the Reactor Plant Component Cooling 

Water System (CCP) temperatures could exceed 125°F. This is the design temperature used in the 

system stress analysis. If RHS heat exchanger operation was initiated at 350'F RCS temperature, as 

assumed, then the RHS heat exchanger CCP outlet temperature could be as high as 250'F if the valves 

failed open. Under the resultant conditions the CCP piping would not meet the ASME Section III, 

Appendix F stress criteria. 

The original plant design did not consider that if the RHS flow control valves failed open on a loss of 

air, it could create unacceptably high RHS heat exchanger CCP discharge temperatures. 

This is considered to be a common-cause failure condition. 

An engineering evaluation identified a failure scenario in which a loss of Instrument Air (IAS) to 

temperature control valves in the Charging Pump Cooling (CCE) system serving the charging pump 

lube oil coolers, coincident with 330'F Service Water (SWP) temperature could result in overcooling of 

both trains of the charging pump lube oil system and challenge charging pump operability. Failure of 

the air-operated CCE valves to the full open position due to a loss of the non-safety related IAS system 

would adversely affect both trains of the charging pumps by allowing excessive cooling of the CCE 

system which cools the lube oil system. This condition alone could have prevented the fulfillment of 

the safety function of the system. The cause of the charging pump inoperability was inadequate original 

design. This condition would result from overcooling of the lube oil system from a failure of the non

safety related Instrument Air system coincident with a worst case minimum SWP temperature and 

maximum flow and heat exchanger cleanliness. Under these conditions, the air-operated CCE valves 

would fail open and excessive cooling of the lube oil system would occur. This particular combination 

of conditions was not considered in the initial design. A short term corrective action was committed to 

in the LER to install a temporary modification to limit the failure position of the three way CCE 

temperature control valve to ensure sufficient bypass flow around the SW heat exchanger to maintain 

CCE temperature above 45'F. Other long-term actions were also discussed. 

This is considered to be a common-cause failure condition.

The LER indicates "multiple" 

AOVs.

Millstone 3 

42396028 

960916
Table 7. (continued).

Plant, LER No.,

Millstone 3 

42396036 

960926

Millstone 3 

42396040 

961024

z4 

(Cý 

0"

flp•rintinn

The High Pressure Safety Injection (SIH) and Low Pressure Safety Injection (SIL) systems would have 

been subject to degraded performance due to possible mispositioning of normally closed safety related 

air operated valves. Mispositioning of these 38 valves can be postulated to occur under post-accident 

harsh environmental conditions due to failure of non-qualified power and control circuits. As a result 

the potential diversion of SIH and/or SIL flow under accident conditions could have been more than the 

margin allowed within the Loss of Coolant Accident analysis. 

Seventeen additional safety-related air operated and solenoid operated valves were subsequently 

identified where failures of non-qualified control circuits could degrade performance of a safety system 

function. These additional valves are located in the following systems: Reactor Plant Component 

Cooling Water (CCP), Containment Vacuum, Reactor Plant Sampling (SSR), Post Accident Sampling 

(SSP), and Main Steam to the auxiliary feedwater steam turbine. 

The cause of the reported conditions was a design mistake. The initial plant design did not adequately 

consider the potential mispositioning of these valves under harsh environmental conditions or active 

failure. 

This is considered to be a common-cause failure condition. 

An engineering evaluation determined that a failure scenario for the Reactor Plant Component Cooling 

Water (CCP) system had the potential for a loss of system safety function. The failure scenario involves 

a loss of the non-category 1 Instrument Air System, which would allow CCP valves to reposition to a 

maximum cooling configuration. Coupled with a low heat load and minimum Service Water (SWP) 

inlet temperature, the CCP system could reach temperatures lower than values for which they are 

analyzed, thereby rendering the CCP system, and systems that it serves, potentially inoperable. This 

was caused by improper initial design of the CCP system. 

The failure mode described in the LER was a design oversight by the plant's architect engineer that 

occurred during the original design process. The design basis analysis for the CCP system focused on 

high CCP heat load conditions. The designers did not analyze for extremely low CCP heat loads 

concurrent with very low SWP temperatures. 

This is considered to be a common-cause failure condition.

Number of AOVs 

Involved in the Event

38 AOVs/SOVs.

At least 6 AOVs were involved 

in this event although the LER 

does not specify an exact 

number. The LER summary 

lists the number of valves as 

"unknown."

CD

and Event Date Descri r tion
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

zý

-3 

CD

The LER indicated that 

"multiple" AOVs were 

involved in this event but no 

specific valves or numbers of 

valves were provided.

Palisades 

25598006 

980305

Event No. 33843 for Palisades in Table 4 was originally recorded for this event. The LER was 

subsequently issued and is discussed here. The original design bases, and subsequent design reviews, for 

the HP Air System address the Large Break Loss of Coolant Accident (LBLOCA), yet did not address, in 

detail, the SBLOCA. As a result, the impact of a loss of the HP Air compressors on the ability to supply 

air needed to align valves for sump recirculation during a SBLOCA, and the need to incorporate manual 

operator actions in EOPs to assure HP air reliability were not considered. 

As a result of the design basis reconstitution for the HP Air System, the licensee identified the need for 

procedural guidance to direct manual operator actions to restore the HP Air System during a SBLOCA. 

The HP Air System supports ECCS realignment to the recirculation mode by providing air required to 

operate control valves needed to switch the suction of the ECCS pumps from the Safety Injection 

Refueling Water Tank (SIRWT) to the containment sump. During a LOCA with LOOP, the HP Air 

System compressors are load shed from safety-related electrical buses and must be manually re-powered 

at their respective motor control centers (MCCs). When the air compressors are without power, the 

receiver tanks are not being charged. During this time, the HP Air System pressure experiences a gradual 

decay due to air bleed-off from pressure regulators and air leakage through seals in piston-driven control 

valve actuators. 

The LER focuses on the procedures needed to ensure restoration of HP air; however, the effects of gradual 

loss of air on the failed position on valves that are supplied by HP air had not been investigated. 

Three downcomer feedwater isolation valves (DCFWIVs) at PVNGS failed to open, following closure 

after a main steam isolation signal (MSIS) during a Unit 1 reactor trip on November 26, 1995. LER 

52895012 covered the reactor trip but not the DCFWIV failures. The source of the information concerning 

the DCFWIV failures was a comprehensive report (the copy received during the plant visit was untitled) 

provided by the licensee during the site visit for this study. 

There are four DCFWIVs in each unit, two in series to each of the two steam generators. The DCFWIVs 

are 8-inch flex-wedge gate AOVs. The actuators are Miller Fluid Power single-acting pneumatic cylinders 

with internal cylinder springs and external spring-and-stanchion setups. The actuators are designed to use 

pressurized nitrogen to open the valves and spring force to close the valves. The pneumatic control circuits 

include a 3-way SOV and a 3-way air switch valve. The valves are intended to fail closed on loss of the air 

signal to the 3-way air switch valve or loss of nitrogen to the operator. The valves are intended to fail open 

if electric power to the SOVs is lost. The DCFWIVs have three key safety functions: heat removal, trip 

initiation, and containment isolation. These valves are also Appendix R safe-shutdown components. 

(continued next page)

Palo Verde I Four AOVs per unit, 12 total.

None 

951126
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Palo Verde I 

None 

951126 

(continued)

The most probable root causes for the multiple valve failures to open were: 

- A lack of prudent actuator design margin. The low actuator margin resulted from using a non

conservative valve factor (0.3) in the original sizing. Recent tests on motor-operated gate valves in 

response to NRC Generic Letter 89-10 indicate that 0.3 is not a conservative valve factor for flex

wedge gate valves; 

- Not allowing for the potential effects of thermal binding in the original sizing of the actuator (Pressure 

locking and thermal binding of power-operated gate valves, including AOVs, isdiscussed in NRC 

Generic Letter 95-07. The DCFWIVs were excluded from the licensee's GL 95-07 evaluation because 

they are normally open valves.); and 

- Not allowing for the potential effects of degradation of the nitrogen supply in the calculation of 

actuator margin. 

Results from static diagnostic testing did not reveal why three of the DCFWIVs failed to open and the 

remaining AOV opened. A dynamic test program to establish the operability of these AOVs was 

conducted and valve factors of 0.64 and 0.57 were established. 

From the perspective of this study, the significance of these AOV failures was that the actuators were 

undersized and had been known to be undersized for some time. This, combined with thermal binding in 

the gate valves and marginal nitrogen pressure all contributed to the common-cause failure of the 

DCFWIVs. 

Another interesting point is that the licensee did not (and did not have to) report the DCFWIV failures to 

the NRC. The key safety functions of the DCFWIVs are to close upon actuation of a main-steam isolation 

signal (MSIS) and to open upon demand to establish a flow path from one of the auxiliary feedwater 

pumps to the steam generators when that pump is chosen to make up steam generator inventory during 

periods of hot standby and plant cooldown. These AOVs successfully closed on demand from the MSIS, 

thus satisfying their safety function. However, the AOVs failed to open following cooldown of the unit, 

and as such, the failures were not classified as safety-related failures. Also, the failures were considered by 

the licensee as not being Maintenance Rule functional failures. The failure mode that occurred was not 

considered by PVNGS to be safety significant or subject to review under the requirements of the 

Maintenance Rule. Three Fisher air-operated letdown/containment isolation valves were found to be 

undersized and the bench settings were also set too low. These valves were provided by the system vendor 

(ABB-CE) and the design basis evaluation was found to be incorrect. Modifications included spring 

replacement, reducing the stroke length of the actuator, and modifying the limit switches. Similar 

conditions were found at Units 2 and 3. 

(continued next page) 

0-E

14
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Palo Verde 1 

52895007 

950512

00

Point Beach 1 

26697014 

970321

z 

tIl 

JI 4.

0'a

Three AOVs per unit (9 total).

Three AOVs per unit (9 total).

Palo Verde I 

None 

951126 

(continued)

The system vendor (ABB-CE) procured valves that had undersized air actuators and bench sets which 

were too low to provide desired valve seating force for the differential pressure which would be present 

across the valves during a letdown line break. They could not determine whether or not the procurement 

of the valves was approved by the Architect Engineer or the owner. The root cause of the design 

deficiency was attributed to the absence of a detailed design basis evaluation for air operated valves as 

part of an AOV program. 

This is considered to be a common-cause failure condition. 

Three Fisher air-operated letdown/containment isolation valves were found to be undersized and the 

bench settings were also set too low. These valves were provided by the system vendor (ABB-CE) and 

the design basis evaluation was found to be incorrect. Modifications included spring replacement, 

reducing the stroke length of the actuator, and modifying the limit switches. Similar conditions were 

found at Units 2 and 3. 

The system vendor (ABB-CE) procured valves that had undersized air actuators and bench sets which 

were too low to provide desired valve seating force for the differential pressure which would be present 

across the valves during a letdown line break. They could not determine whether or not the procurement 

of the valves was approved by the Architect Engineer or the owner. The root cause of the design 

deficiency was attributed to the absence of a detailed design basis evaluation for air-operated valves as 

part of an AOV program. 

This is considered to be a common-cause failure condition. 

Engineers discovered a condition that alone could have prevented the Auxiliary Feedwater (AFW) 

System from automatically performing its safety-related function during design basis accidents 

involving a loss of instrument air and reduced steam generator pressures. A loss of instrument air 

during the accident would cause both motor-driven AFW pump (MDAFWP) flow control valves to fail 

open. The postulated event scenario is the result of a latent characteristic of the original AFW system 

design. The original design did not provide ample assurance that the MDAFWPs would automatically 

function during all design basis events. The expected fix was to furnish a reliable pneumatic supply to 

the control valves. Event No. 31995 refers. 

This is considered to be a common-cause failure condition.

Two AOVs.
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Salem 1 

27296012 

960723

On Three AOVs per unit (6 total). July 23, 1996, a review determined that the keys/keyways on the actuators for the Residual Heat 

Removal (RHR) flow control valves (6 AOVs per unit) were subject to failure. The valves were made 

by Fisher Controls International, Model Type 656/7600. When using the simplified Fisher Catalog 14 

methodology, the calculated maximum stem torque exceeds the vendor specified allowable torque. 

Preliminary calculations of the shaft torque and resulting average shear stresses in the valve stem key 

and keyway were also performed. The calculations showed that for normal operating conditions, the 

average shear stress in the key may exceed the material yield stress. The calculations also showed that 

although the average shear stress in the valve stem was estimated to be less than the minimum material 

yield stress, the fatigue life of the shaft keyway appeared to be limited to a low number of cycles. Key 

failures in the past are attributed to an overload during normal operation of these valves. 

A review of the original design revealed that these valves were installed with little or no design margin 

and the keys were likely to fail due to low cycle fatigue with stress levels exceeding yield strength. 

Corrective action was to replace the valves and review Fisher Model 7600 valves for similar concerns. 

This is considered to be a common-cause-failure condition. 

On December 10, 1996, the licensee, in developing a test program for air operated valves, applied test 

equipment to containment isolation valve 2HV0513. It was likely that the actuator settings for valve 

2HV0513 would not generate sufficient closing force to overcome internal pressure and packing drag 

under design basis conditions. The licensee concluded that valve 2HV0513 had been inoperable when 

Unit 2 was in Mode I.The licensee concluded that either of two separate errors could have caused valve 

2HV0513 to have insufficient closing force: (1) vendor setpoint methodology error, or (2) a deficiency 

in the reassembly procedure. 

The licensee committed to reset and retest valve 2HV0513 prior to returning Unit 2 to Mode 4. They 

also were to upgrade its maintenance procedures to conform to the current vendor manual. They 

performed an analysis on all air-operated containment isolation valves and confirmed that all other 

Units 2 and 3 air-operated containment isolation valves have sufficient actuator closing force to remain 

operable. The licensee also was to verify actuator settings for Unit 2 air-operated containment isolation 

valves which may not have had their actuator settings established in accordance with the vendor 

recommendations. Similar actions were to have been completed for Unit 3 during its next refueling 

outage.

One AOV each in units 2 and 3 

(2 total). San Onofre 2 

36196011 

961216

60 

ON -D
Table 7. (continued).

Plant, LER No., 

and Event Date 

San Onofre 2 

(and 3) 

36199003 

990210

feserintion

z 

0' 

0*

Number of AOVs 

Involved in the Event

On February 7, 1999, while Unit 2 was shut down for its Cycle 10 refueling outage, the Component Four AOVs per unit (8 total).

Cooling Water Non-Critical Loop (CCW-NCL) isolation valves failed to satisfy close stroke time 

requirements during a pneumatic supply test. Subsequent investigations determined that the 

corresponding valves in Unit 3 might not satisfy stroke time requirements. On February 10, 1999, the 

non-critical loop valves at Unit 3 were declared inoperable and the unit entered a 72-hour action 

statement. The cause of this condition was that instrument air failure modes other than a loss of 

pressure were not adequately considered in the original design of the CCW-NCL valves. Design 

modifications were implemented on both units to correct a previously unrecognized condition 

associated with a blockage of the non-seismic pneumatic supply. 

The San Onofre Nuclear Generating Station (SONGS) Component Cooling Water System (CCW) 

consists of two independent critical loops (trains) and one non-critical loop (NCL). All three loops are 

interconnected, such that the NCL can be aligned to either one of the critical loops. A third swing pump 

can be aligned to either CCW train. The safety function of CCW is to transfer the combined heat load 

from safety-related systems and components to the Salt Water Cooling System (SWC) during normal 

and accident conditions, including seismic events. The CCW-NCL isolation valves (HV-6212 and 6218 

for Train A, and HV-6213 and HV-6219 for Train B) are designed to close upon receipt of a 

containment isolation signal or a low-low level signal in the associated CCW train surge tank. These 

pneumatically actuated valves (28-inch Fisher Type 9241) are each required to close upon loss of 

pneumatic supply and/or loss of control power to its solenoid valve, and are equipped with twin air 

receivers to ensure valve closure. The air receivers are aligned to the valve actuator by a pneumatic trip 

valve that actuates on low supply pressure. The pneumatic system incorporates a 4-way solenoid 

supply/exhaust valve, two air receiver units, and a pressure actuated trip valve. Normal pneumatic 

pressure is provided by the instrument air system. De-energizing the solenoid valve vents the top 

chamber of the piston actuator and supplies air to the bottom chamber through the normal supply port 

of the trip valve. The differential pressure across the actuator piston causes it to move, thereby closing 

the valve. The valve actuator is designed to fail in the closed position upon loss of normal supply 

pressure or loss of control power. The air receivers provide for valve closure upon a loss of air/nitrogen 

pressure. If the supply pressure drops to less than 75% of the pressure in the air receivers, the 

differential pressure created across the trip valve piston moves to block the normal supply port and 

open a flow path from the air receivers to the actuator. If the pneumatic supply pressure remains above 

this set point, the trip valve allows normal automatic operation with the solenoid valve. In 1995, the 

non-safety-related nitrogen lines were connected locally to the non-safety-related air supply to increase 

reliability of cooling to the reactor coolant pump seals. This provided a backup supply to preclude 

spurious CCW-NCL valve closure. This modification included installation of check valves in both the 

air and nitrogen systems to isolate one system from another. (continued next page)

,. 

0"

0

Descri tion
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event 

San Onofre 2 In October 1997, the licensee recognized that CCW-NCL supply and return valves were required to 

(and 3) close within a certain sequence and time band to preclude a potential water hammer during postulated 

36199003 accident conditions. In the fall of 1998, the licensee recognized that existing test records did not 

990210 document the valve closure times using the accumulators only. Testing to obtain these valve closure 

(continued) times was completed during the Unit 2 Cycle 10 refueling outage in early 1999. 

During the Cycle 10 refueling outage, the closure function of the Unit 2 Train A CCW-NCL isolation 

valves was tested by various modes of closure. Results using the plant nitrogen and the instrument air 

system were satisfactory. Closure via the air receivers was tested by isolating the air and nitrogen to the 

valve actuator and then de-energizing the supply solenoid. During this test, 2HV6212 did not move for 

several minutes and 2HV6218 closed in approximately 47 seconds. These stroke times exceeded the 

valve's operability criteria. Follow-up investigation determined that with the pneumatic supply isolated, 

but with the supply line still pressurized, the trip valve in the actuator pneumatic system would not 

reliably actuate. With the pneumatic supply isolated, de-energizing the solenoid valve causes the supply 

volume, trapped upstream of the solenoid valve, to expand into the volume downstream of the solenoid 

valve. The resulting pressure depends on the volume trapped upstream of the solenoid relative to the 

volume downstream. The valve will stroke if the trapped upstream volume is either small compared to 

the downstream volume or very large compared to the downstream volume. If the trapped upstream 

-- volume is small compared to the downstream volume, the resulting pressure will be low enough to fully 

actuate the trip valve. If the trapped volume falls between these limits, the pressure will drop to, or 

close to, the trip valve set point. At this pressure, the trip valve piston may only partially move, 

blocking both the normal and accumulator air supply to the NCL valve actuator. As such, the valve may 

not be able to perform its safety function under all design basis conditions. 

Based on the results of the investigation described above, the licensee concluded that all 8 CCW-NCL 

valves could potentially be affected. 

4ro
Table 7. (continued).

z 

Cr

Plant, LER No., 

and Event Date 

Shearon Harris 

40098001 

980109 

Summer 

39598009 

981006

Description 

On January 9, 1998, a condition was identified during operation that results in the plant being potentially 

outside it's design basis. Specifically, a potential failure mechanism exists where a leak in the non-safety 

Instrument Air System could result in the inoperability of the Steam Generator Pre-heater Bypass Isolation 

Valves. These valves are safety-related containment isolation valves that are required by plant procedures 

to automatically shut in 10 seconds or less upon receipt of a Main Feedwater Isolation Signal. These 

valves are positioned by a pneumatic piston-operated actuator which is supplied by the non-safety-related 

Instrument Air System. They are designed to automatically close if control air supply is lost. An air leak 

was postulated in the Instrument Air system that could possibly reduce the air inlet pressure to just low 

enough to affect proper operation of the actuator's 3-way and 4-way pilot valves and not be detected by 

Operations personnel. If this occurred, the pilot valves would shuttle, causing the accumulator pressure to 

bleed off, which would prevent the valves from closing as required. This potential scenario constituted 

operation outside the design basis of the plant and was reported to the NRC via the emergency notification 

system on January 9, 1998, at 1450 hours. The cause of this condition was inadequate design control 

during development of a plant modification implemented in August 1984 in response to NRC Information 

Notice 82-25. The investigation for this event also revealed several other missed opportunities to identify 

this condition during subsequent plant modifications and/or related evaluations. 

It was discovered that there was an issue concerning the ability of a non-safety-related component to affect 

the operation of the Turbine Driven Emergency Feedwater Pump (TDEFWP). This non-safety- related 

device (ISY-02034) is a current to pneumatic (I/P) transducer used in the speed control circuitry for the 

TDEFWP. The licensee conducted an engineering evaluation of the failure of the TDEFWP to achieve 

rated speed and concluded that there were failure mechanisms that could have prevented the TDEFWP 

from performing its design function under postulated accident conditions outside containment. The I/P had 

previously been thought to only fall (on loss of air or power) to full speed. During a steam line break 

coincident with the loss of offsite power, and a loss of "B" train DC power, there may not have been 

sufficient EFW flow to the non-faulted steam generators to mitigate the accident. The unanalyzed 

condition Is the potential failure mechanisms/modes for the I/P converter which allow it to fail in a 

position which corresponds to less than rated speed. 

The I/P could fail in such a way that the TDEFWP speed never reaches the required speed and flow. This 

was determined during the engineering evaluation for condition report 98-0823 after the TDEFW did not 

achieve rated speed due to internal contamination of the I/P converter. 

The cause of this event was the belief that the I/P could only fail due to loss of air or power. As a result of 

this belief, no other failure modes were considered credible. This was an engineering oversight dating 

back prior to the receipt of the plant operating license. 

Other I/Ps in the plant were evaluated by the licensee. The licensee discovered that the non-qualified sub

components could potentially fail nonconservatively in a high radiation field that will occur after an 

accident during the recirculation or cooldown phase.

Number of AOVs 

Involved in the Event 

Three AOVs. 

The LER lists "multiple" flow 

control valves and an I/P 

controller.

N0
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Three Mile Island 1 

Vermont Yankee 

27198025 (Rev. 2) 

981211

Potentially insufficient margins in two AOVs were described by the TMI engineers during the visit for 

this AOV study. Calculations for five Crane-Aloyco, 2.5 to 6 inch, 1500 and 150 pound class, 

flex-wedge and split wedge gate AOVs with Miller DA-63-B and A-63-B cylinder actuators were 

reviewed by one of the TMI engineers. The architect/engineer (AE) requested that the valve 

manufacturer (Crane-Aloyco) perform thrust calculations on the five AOVs, as part of a limit switch 

upgrade modification, in order to verify that limit switch installation would not affect valve operability. 

The resultant thrust calculations, using "present day" methodology, indicated that two of the five valves 

had negative closing margins for the specified differential pressure (d/P). The AE then requested that 

the manufacturer re-perform the calculations using the methodology by which the valves were 

originally sized. The revised Crane-Aloyco calculations, based on the original valve sizing 

methodology, indicated positive margin in both the opening and closing directions, but using a valve 

factor of zero. Upon review of the Crane-Aloyco calculations, TMI convened a review group. This 

group verified that the two containment isolation valves, which had negative margins in the 

manufacturers' first calculations, were, in fact, operable and would be able to perform their designed 

safety function because the required operating conditions were less severe than the original design basis 

conditions. This conclusion was based on TMI's calculations using a 0.75 friction factor and d/P of 

1600 psi (that required for containment isolation). 

On 12/11/98, with the plant at 100% power, it was determined that one Scram Discharge Volume 

(SDV) drain valve (CRD-LCV-33B) did not meet the stroke time requirements of the In-Service Test 

(IST) Program. Both the North and the South SDVs have two drain valves in series, CRD-LCV-33A/C 

on the North and CRD-LCV-33B/D on the South. The drain valve, CRD-LCV-33B, was subsequently 

declared inoperable. 

The SDVs are used to limit the loss of and contain the reactor vessel water from all control rod drives 

during a scram. These volumes are provided in the scram discharge header. During normal plant 

operation, the volumes are empty with all the drain and vent valves open. Upon receipt of a scram 

signal, the vent and drain valves close. While scrammed, the control rod drive seal leakage continues to 

flow to the discharge volumes until the discharge volume equals reactor pressure. Following a scram, 

when the scram signals are cleared and the Reactor Protection System (RPS) logic is manually reset, the 

vent and drain valves are opened and the SDV drained. 

According to the LER, the root causes of this event were inadequate actuator sizing calculations 

(performed by the vendor) under the design conditions specified in the procurement specification, and 

inadequate manufacturing Quality Assurance controls to ensure specification requirements were 

maintained. Contributing causes included conflicting information in the design package that was not 

identified by Vermont Yankee, and closing forces of the actuator at either end of the valve stroke were 

apparently not as designed. (continued next page)

0,

Five AOVs. 

Four AOVs.

Ea 

CD~

z 

70~ 

(.1
Table 7. (continued). 

Plant, LER No., Number of AOVs 

and Event Date Description Involved in the Event

Vermont Yankee 

27198025 

981211 

(continued)

Z 

C71 

0

NOTES: 

1. The NRC Licensee Event Report (LER) Number consists of the three-digit NRC Docket Number for the plant at which the event occurred, the last two digits of the year in 

which the LER was generated, and a three-digit sequential number of the LER. This is consistent with the NRC's Sequence Code Search System (SCSS) database 

designation. The LER system is described in 10 CFR 50.73. 0o

(continued from previous page) 

The actuators and valves were installed under Engineering Design Change EDCR 97410 in April of 

1998 and were sized in accordance with the vendor's recommendations. These drain valves are required 

to go shut on a scram signal to isolate the SDV and act as a primary containment isolation valve when 

the scram valves are open. This issue was addressed by the installation of larger actuators, on 

December 21, 1998, via the Minor Modification process, which were properly sized to operate the 

valves under any design conditions at Vermont Yankee. 

The licensee stated that the LER constituted a Part 21 notification in accordance with IOCFR21.2(c) 

and NUREG 1022, Revision 1. Further, the licensee informed NRC by Event 35150 dated 

December 14, 1998, that this event was considered to be a common-cause failure.

,..
Appendix A 

NRC Generic Communications Related 

to Air-Operated Valves
Appendix A

NRC Generic Communications Related 

to Air-Operated Valves 

1. NRC GENERIC COMMUNICATIONS RELATED TO AIR-OPERATED VALVES 

2. Bulletin 71-001 - PERFORMANCE OF THE MAIN STEAM ISOLATION VALVES 

3. Bulletin 73-002 - MALFUNCTION OF CONTAINMENT PURGE SUPPLY VALVE SWITCH 

4. Bulletin No. 75-03 -INCORRECT LOWER DISC SPRING AND CLEARANCE DIMENSION 

IN SERIES 8300 AND 8302 ASCO SOLENOID VALVES 

5. Bulletin 76-006 - DIAPHRAGM FAILURES IN AIR OPERATED AUXILIARY ACTUATORS 

FOR SAFETY/RELIEF VALVES 

6. Bulletin 78-004 - ENVIRONMENTAL QUALIFICATION OF CERTAIN STEM MOUNTED 

LIMIT SWITCHES INSIDE REACTOR CONTAINMENT 

7. Bulletin 78-014 - DETERIORATION OF BUNA-N COMPONENTS IN ASCO SOLENOIDS 

8. Bulletin 79-001A - ENVIRONMENTAL QUALIFICATION OF CLASS 1E EQUIPMENT 

(DEFICIENCIES IN THE ENVIRONMENTAL QUALIFICATION OF ASCO SOLENOID 

VALVES 

9. Bulletin 79-OOIB - ENVIRONMENTAL QUALIFICATION OF CLASS IE EQUIPMENT 

10. Bulletin 79-OOIB Sup. 2 - ENVIRONMENTAL QUALIFICATION OF CLASS 1E EQUIPMENT 

11. Bulletin 80-001 - OPERABILITY OF ADS VALVE PNEUMATIC SUPPLY 

12. Bulletin 80-014 - DEGRADATION OF BWR SCRAM DISCHARGE VOLUME CAPABILITY 

13. Bulletin 80-016 - POTENTIAL MISAPPLICATION OF ROSEMOUNT INC. MODELS 1151 

AND 1152 PRESSURE TRANSMITTERS WITH EITHER "A" OR "D" OUTPUT CODES 

14. Bulletin 80-017 - FAILURE OF 76 OF 185 CONTROL RODS TO FULLY INSERT DURING A 

SCRAM AT A BWR 

15. Bulletin 80-017 Sup. 2 - FAILURES REVEALED BY TESTING SUBSEQUENT TO FAILURE 

OF CONTROL RODS TO INSERT DURING A SCRAM AT A BWR 

16. Bulletin 80-017 Sup. 3 - FAILURE OF CONTROL RODS TO INSERT DURING A SCRAM AT 

ABWR 

17. Bulletin 80-017 Sup. 4 - FAILURE OF CONTROL RODS TO INSERT DURING A SCRAM AT 

A BWR

A-1 NUREG/CR-6654
Appendix A

18. Bulletin 80-023 - FAILURES OF SOLENOID VALVES MANUFACTURED BY VALCOR 

ENGINEERING CORPORATION 

19. Bulletin 80-025 - OPERATING PROBLEMS WITH TARGET ROCK SAFETY-RELIEF 

VALVES AT BWRs 

20. Bulletin 86-01 - MINIMUM FLOW LOGIC PROBLEMS THAT COULD DISABLE RHR 

PUMPS 

21. Bulletin 86-03 - POTENTIAL FAILURE OF MULTIPLE ECCS PUMPS DUE TO SINGLE 

FAILURE OF AIR-OPERATED VALVE IN MINIMUM FLOW LINE 

22. Bulletin 88-04 - POTENTIAL SAFETY-RELATED PUMP LOSS 

23. Circular 79-018 - PROPER INSTALLATION OF TARGET ROCK SAFETY-RELIEF 

24. Circular 79-022 - STROKE TIMES FOR POWER OPERATED RELIEF VALVES 

25. Circular 80-008 - BWR TECHNICAL SPECIFICATION INCONSISTENCY - RPS RESPONSE 

TIME 

26. Circular 80-015 - LOSS OF REACTOR COOLANT PUMP COOLING AND NATURAL 

CIRCULATION COOLDOWN 

27. Circular 81-014 - MAIN STEAM ISOLATION VALVE FAILURES TO CLOSE 

28. Circular 81-015 - UNNECESSARY RADIATION EXPOSURES TO THE PUBLIC AND 

WORKERS DURING EVENTS INVOLVING THICKNESS AND LEVEL 

29. Generic Letter 79-024 - INADVERTENT REACTOR SCRAM AND SAFETY INJECTION 

DURING MONTHLY SURVEILLANCE TESTS OF THE SAFEGUARDS 

30. Generic Letter 79-045 - TRANSMITTAL OF REPORTS REGARDING FOREIGN REACTOR 

OPERATING EXPERIENCES 

31. Generic Letter 79-046 - CONTAINMENT PURGING AND VENTING DURING NORMAL 

OPERATION - GUIDELINES FOR VALVE OPERABILITY 

32. Generic Letter 79-054 - CONTAINMENT PURGING AND VENTING DURING NORMAL 

OPERATION 

33. Generic Letter 80-004 - OPERABILITY OF ADS VALVE PNEUMATIC SUPPLY 

34. Generic Letter 80-012 - ANALYSIS OF A PWR MAIN STEAM LINE BREAK WITH 

CONTINUED FEEDWATER ADDITION 

35. Generic Letter 80-029 - MODIFICATIONS TO BOILING WATER REACTOR CONTROL ROD 

DRIVE SYSTEMS 

36. Generic Letter 80-097 - iE Bulletin NO. 80023: FAILURES OF SOLENOID VALVES 

MANUFACTURED BY VALCOR ENGINEERING CORPORATION

NUREG/CR-6654 A-2
Appendix A

37. Generic Letter 81-009 - BWR SCRAM DISCHARGE SYSTEM 

38. Generic Letter 81-014 - SEISMIC QUALIFICATION OF AUXILIARY FEEDWATER 

SYSTEMS 

39. Generic Letter 81-021 - NATURAL CIRCULATION COOLDOWN 

40. Generic Letter 87-012 - LOSS OF RESIDUAL HEAT REMOVAL (RHR) WHILE THE 

REACTOR COOLANT SYSTEM (RCS) IS PARTIALLY FILLED 

41. Generic Letter 88-014 - INSTRUMENT AIR SUPPLY SYSTEM PROBLEMS AFFECTING 

SAFETY- RELATED EQUIPMENT 

42. Generic Letter 88-017 - LOSS OF DECAY HEAT REMOVAL 

43. Generic Letter 90-006 - RESOLUTION OF GENERIC ISSUE 70, "POWER-OPERATED 

RELIEF VALVE AND BLOCK VALVE RELIABILITY," AND GENERIC ISSUE 94, 

"ADDITIONAL LTOP FOR LWRs" 

44. Generic Letter 91-015 - OPERATING EXPERIENCE FEEDBACK REPORT, SOLENOID

OPERATED VALVE PROBLEMS AT U.S. REACTORS 

45. Generic Letter 95-07 - PRESSURE LOCKING AND THERMAL BINDING OF SAFETY

RELATED POWER-OPERATED GATE VALVES 

46. Generic Letter 96-05 - PERIODIC VERIFICATION OF DESIGN-BASIS CAPABILITY OF 

SAFETY-RELATED MOTOR-OPERATED VALVES 

47. Notice 79-04 - DEGRADATION OF ENGINEERED SAFETY FEATURES 

48. Notice 79-08 - INTERCONNECTION OF CONTAMINATED SYSTEMS WITH SERVICE AIR 

SYSTEMS USED AS THE SOURCE OF BREATHING AIR 

49. Notice 79-27 - STEAM GENERATOR TUBE RUPTURES AT TWO POWER PLANTS 

50. Notice 80-11 -GENERIC PROBLEMS WITH ASCO VALVES IN NUCLEAR APPLICATIONS 

INCLUDING FIRE PROTECTION SYSTEMS SSINS No. 6870 

51. Notice 80-16 - SHAFT SEAL PACKING CAUSES BINDING IN MAIN STEAM SWING DISC 

CHECK AND ISOLATION VALVES 

52. Notice 80-39 - MALFUNCTION OF SOLENOID VALVES MANUFACTURED BY VALCOR 

ENGINEERING CORPORATION 

53. Notice 80-40 - EXCESSIVE NITROGEN SUPPLY PRESSURE ACTUATES SAFETY-RELIEF 

VALVE OPERATION TO CAUSE REACTOR DEPRESSURIZATION 

54. Notice 81-12 - GUIDANCE ON ORDER ISSUED JANUARY 9,1981, REGARDING 

AUTOMATIC CONTROL ROD INSERTION ON LOW CONTROL AIR PRESSURE 

55. Notice 81-14 - POTENTIAL OVERSTRESS OF SHAFTS ON FISHER SERIES 9200 

BUTTERFLY VALVES WITH EXPANDABLE T RINGS

A-3 NUREG/CR-6654
Appendix A

56. Notice 81-15 - DEGRADATION OF AUTOMATIC ECCS ACTUATION CAPABILITY BY 

ISOLATION OF INSTRUMENT LINES 

57. Notice 81-27 - FLAMMABLE GAS MIXTURES IN THE WASTE GAS DECAY TANKS IN 

PWR PLANTS 

58. Notice 81-29 - EQUIPMENT QUALIFICATION TESTING EXPERIENCE 

59. Notice 81-38 -POTENTIALLY SIGNIFICANT EQUIPMENT FAILURES RESULTING FROM 

CONTAMINATION OF AIR-OPERATED SYSTEMS 

60. Notice 82-17 - OVER PRESSURIZATION OF REACTOR COOLANT SYSTEM 

61. Notice 82-19 - LOSS OF HIGH HEAD SAFETY INJECTION EMERGENCY BORATION AND 

REACTOR COOLANT MAKEUP CAPABILITY 

62. Notice 82-25 - FAILURES OF HILLER ACTUATORS UPON GRADUAL LOSS OF AIR 

PRESSURE 

63. Notice 82-45 - PWR LOW TEMPERATURE OVERPRESSURE PROTECTION 

64. Notice 82-52 - EQUIPMENT ENVIRONMENTAL QUALIFICATION TESTING EXPERIENCE 

- UPDATING OF TEST SUMMARIES PREVIOUSLY PUBLISHED IN INFORMATION 

NOTICE 81-29 

65. Notice 83-57 - POTENTIAL MISASSEMBLY PROBLEM WITH AUTOMATIC SWITCH 

COMPANY (ASCO) SOLENOID NP 8316 

66. Notice 83-70 - VIBRATION-INDUCED VALVE FAILURES 

67. Notice 83-70 Supplement 1 - VIBRATION INDUCED VALVE FAILURES 

68. Notice 84-04 - FAILURE OF ELASTOMER SEATED BUTTERFLY VALVES USED ONLY 

DURING COLD SHUTDOWN 

69. Notice 84-12 -FAILURE OF SOFT SEAT VALVE SEALS 

70. Notice 84-23 - RESULTS OF THE NRC-SPONSORED QUALIFICATION METHODOLOGY 

RESEARCH TEST ON ASCO SOLENOID VALVES 

71. Notice 84-31 - INCREASED STROKING TIME OF BETTIS ACTUATORS BECAUSE OF 

SWOLLEN ETHYLENE-PROPYLENE RUBBER SEALS AND SEAL SET 

72. Notice 84-48 - FAILURES OF ROCKWELL INTERNATIONAL GLOBE VALVES 

73. Notice 84-48 Supplement 1 - FAILURES OF ROCKWELL INTERNATIONAL GLOBE 

VALVES 

74. Notice 84-53 - INFORMATION CONCERNING THE USE OF LOCT1TE 242 AND OTHER 

ANAEROBIC ADHESIVE/SEALANTS

NUREG/CR-6654 A-4
Appendix A

75. Notice 84-68 - POTENTIAL DEFICIENCY IN IMPROPERLY RATED FIELD WIRING TO 

SOLENOID VALVES 

76. Notice 84-74 - ISOLATION OF REACTOR COOL LOW-PRESSURE SYSTEMS OUTS 

77. Notice 84-81 - INADVERTENT REDUCTION IN PRIMARY COOLANT INVENTORY IN 

BOILING WATER REACTORS DURING SHUTDOWN AND STARTUP 

78. Notice 85-06 - CONTAMINATION OF BREATHING AIR SYSTEMS 

79. Notice 85-08 - INDUSTRY EXPERIENCE ON CERTAIN MATERIALS USED IN SAFETY

RELATED EQUIPMENT 

80. Notice 85-17 - POSSIBLE STICKING OF ASCO SOLENOID VALVES 

81. Notice 85-17 Supplement 1 - POSSIBLE STICKING OF ASCO SOLENOID VALVES 

82. Notice 85-21 - MAIN STEAM ISOLATION VALVE CLOSURE LOGIC 

83. Notice 85-26 - VACUUM RELIEF SYSTEM FOR BOILING WATER REACTOR MARK I AND 

MARK II CONTAINMENTS 

84. Notice 85-27 - NOTIFICATIONS TO THE NRC OPERATIONS CENTER AND REPORTING 

EVENTS IN LICENSEE EVENT REPORTS 

85. Notice 85-35 -FAILURE OF AIR CHECK VALVES TO SEAT 

86. Notice 85-35 Supplement 1 - FAILURE OF AIR CHECK VALVES TO SEAT 

87. Notice 85-47 -POTENTIAL EFFECT OF LINE-INDUCED VIBRATION ON CERTAIN 

TARGET ROCK SOLENOID-OPERATED VALVES 

88. Notice 85-59 - VALVE STEM CORROSION FAILURES 

89. Notice 85-67 - VALVE-SHAFT-TO-ACTUATOR KEY MAY FALL OUT OF PLACE WHEN 

MOUNTED BELOW HORIZONTAL AXIS 

90. Notice 85-72 - UNCONTROLLED LEAKAGE OF REACTOR COOLANT OUTSIDE 

CONTAINMENT 

91. Notice 85-75 - IMPROPERLY INSTALLED INSTRUMENTATION, INADEQUATE QUALITY 

CONTROL AND INADEQUATE POSTMODIFICATION TESTING 

92. Notice 85-84 - INADEQUATE INSERVICE TESTING OF MAIN STEAM ISOLATION 

VALVES 

93. Notice 85-95 - LEAK OF REACTOR WATER TO REACTOR BUILDING CAUSED BY 

SCRAM SOLENOID VALVE PROBLEM 

94. Notice 85-100 - ROSEMOUNT DIFFERENTIAL PRESSURE TRANSMITTER ZERO POINT 

SHIFT

A-5 NUREG/CR-6654
Appendix A

95. Notice 86-09 - FAILURE OF CHECK AND STOP CHECK VALVES SUBJECTED TO LOW 

FLOW CONDITIONS 

96. Notice 86-16 - FAILURES TO IDENTIFY CONTAINMENT LEAKAGE DUE TO 

INADEQUATE LOCAL TESTING OF BWR VACUUM RELIEF SYSTEM VALVES 

97. Notice 86-50 - INADEQUATE TESTING TO DETECT FAILURES OF SAFETY-RELATED 

PNEUMATIC COMPONENTS OR SYSTEMS 

98. Notice 86-51 -EXCESSIVE PNEUMATIC LEAKAGE IN THE AUTOMATIC 

DEPRESSURIZATION SYSTEM 

99. 86-57 - OPERATING PROBLEMS - VALVES AT NUCLEAR POWER PLANTS 

100. Notice 86-72 - FAILURE 17-7 PH STAINLESS STEEL SPRINGS IN VALCOR VALVES DUE 

TO HYDROGEN EMBRITTLEMENT 

101. Notice 86-78 - SCRAM SOLENOID PILOT VALVE (SSPV) REBUILT KIT PROBLEMS 

102. Notice 86-82 - FAILURES OF SCRAM DISCHARGE VOLUME VENT AND DRAIN VALVES 

103. Notice 86-82 Rev. 1 - FAILURES OF SCRAM DISCHARGE VOLUME VENT AND DRAIN 

VALVES 

104. Notice 86-109 - DIAPHRAGM FAILURE IN SCRAM OUTLET VALVE CAUSING ROD 

INSERTION 

105. Notice 87-02 - INADEQUATE SEISMIC QUALIFICATION OF DIAPHRAGM VALVES BY 

MATHEMATICAL MODELING AND ANALYSIS 

106. Notice 87-17 - RESPONSE TIME OF SCRAM INSTRUMENT VOLUME LEVEL DETECTORS 

107. Notice 87-28 - AIR SYSTEMS PROBLEMS AT U.S. LIGHT WATER REACTORS 

108. Notice 87-28 Sup. 1 - AIR SYSTEMS PROBLEMS AT U.S. LIGHT WATER REACTORS 

109. Notice 87-38 - INADEQUATE OR INADVERTENT BLOCKING OF VALVE MOVEMENT 

110. Notice 87-48 - INFORMATION CONCERNING THE USE OF ANAEROBIC 

ADHESIVEISEALANTS 

111. Notice 88-24 - FAILURES OF AIR-OPERATED VALVES AFFECTING SAFETY-RELATED 

SYSTEMS 

112. Notice 88-43 - SOLENOID VALVE PROBLEMS 

113. Notice 88-51 -FAILURES OF MAIN STEAM ISOLATION VALVES 

114. Notice 88-86 - OPERATING WITH MULTIPLE GROUNDS IN DIRECT CURRENT 

DISTRIBUTION SYSTEMS

NUREG/CR-6654 A-6
Appendix A

115. Notice 88-86, Supplement 1 - OPERATING WITH MULTIPLE GROUNDS IN DIRECT 

CURRENT DISTRIBUTION SYSTEMS 

116. Notice 88-94 - POTENTIALLY UNDERSIZED VALVE ACTUATORS 

117. Notice 88-97 - POTENTIALLY SUBSTANDARD VALVE REPLACEMENT PARTS 

118. Notice 88-97 Sup. 1 - POTENTIALLY SUBSTANDARD VALVE REPLACEMENT PARTS 

119. Notice 89-07 - FAILURES OF SMALL-DIAMETER TUBING IN CONTROL AIR, FUEL OIL, 

AND LUBE OIL SYSTEMS WHICH RENDER EMERGENCY DIESEL 

120. Notice 89-26 - INSTRUMENT AIR SUPPLY TO SAFETY-RELATED EQUIPMENT 

121. Notice 89-28 - WEIGHT AND CENTER OF GRAVITY DISCREPANCIES FOR COPES

VULCAN AIR-OPERATED VALVES 

122. Notice 89-30 - HIGH TEMPERATURE ENVIRONMENTS AT NUCLEAR POWER PLANTS 

123. Notice 89-38 - ATMOSPHERIC DUMP VALVE FAILURES AT PALO VERDE UNITS 1, 2, 

AND 3 

124. Notice 89-52 - POTENTIAL FIRE DAMPER OPERATIONAL PROBLEMS 

125. Notice 89-54 - POTENTIAL OVER PRESSURIZATION OF THE COMPONENT COOLING 

WATER SYSTEM 

126. Notice 89-66 - QUALIFICATION LIFE OF SOLENOID VALVES 

127. Notice 90-06 - POTENTIAL FOR LOSS OF SHUTDOWN COOLING WHILE AT LOW 

REACTOR COOLANT LEVELS 

128. Notice 90-11 - MAINTENANCE DEFICIENCY ASSOCIATED WITH SOLENOID-OPERATED 

VALVES 

129. Notice 90-17 - WEIGHT AND CENTER OF GRAVITY DISCREPANCIES FOR COPES

VULCAN VALVES 

130. Notice 90-18 - POTENTIAL PROBLEMS WITH CROSBY SAFETY RELIEF VALVES USED 

ON DIESEL GENERATOR AIR START RECEIVER TANKS 

131. Notice 90-64 - POTENTIAL FOR COMMON-MODE FAILURE OF HIGH PRESSURE SAFETY 

INJECTION PUMPS OR RELEASE OF REACTOR COOLANT OUTSIDE CONTAINMENT 

DURING A LOCA 

132. Notice 91-58 - DEPENDENCY OF OFFSET DISC BUTTERFLY VALVE'S OPERATION ON 

ORIENTATION WITH RESPECT TO FLOW 

133. Notice 91-83 - SOLENOID-OPERATED VALVE FAILURES RESULTED IN TURBINE 

OVERSPEED 

134. Notice 92-60 - VALVE STEM FAILURE CAUSED BY EMBRITTLEMENT

A-7 NUREG/CR-6654
Appendix A

135. Notice 92-64 - NOZZLE RING SETTINGS ON LOW PRESSURE WATER-RELIEF VALVES 

136. Notice 92-67 - DEFICIENCY IN DESIGN MODIFICATIONS TO ADDRESS FAILURES OF 

HILLER ACTUATORS UPON A GRADUAL LOSS OF AIR PRESSURE 

137. Notice 93-35 - INSIGHTS FROM COMMON-CAUSE FAILURE EVENTS 

138. Notice 94-06 - POTENTIAL FAILURE OF LONG-TERM EMERGENCY NITROGEN SUPPLY 

FOR THE AUTOMATIC DEPRESSURIZATION SYSTEM VALVES 

139. Notice 94-25 - FAILURE OF CONTAINMENT SPRAY HEADER VALVE TO OPEN DUE TO 

EXCESSIVE PRESSURE FROM INERTIAL EFFECTS OF WATER 

140. Notice 94-44 - MAIN STEAM ISOLATION VALVE FAILURE TO CLOSE ON DEMAND 

BECAUSE OF INADEQUATE MAINTENANCE AND TESTING 

141. Notice 94-55 - PROBLEMS WITH COPES-VULCAN PRESSURIZER POWER- OPERATED 

RELIEF VALVES 

142. Notice 94-61 - CORROSION OF WILLIAM POWELL GATE VALVE DISC HOLDERS 

143. Notice 94-61 Sup. 1 - CORROSION OF WILLIAM POWELL GATE VALVE DISC HOLDERS 

144. Notice 94-71 - DEGRADATION OF SCRAM SOLENOID PILOT VALVE PRESSURE AND 

EXHAUST DIAPHRAGMS 

145. Notice 95-25 - VALVE FAILURE DURING PATIENT TREATMENT WITH GAMMA 

STEREOTACTIC RADIOSURGERY UNIT 

146. Notice 95-34 - AIR ACTUATOR AND SUPPLY AIR REGULATOR PROBLEMS IN COPES

VULCAN PRESSURIZER POWER-OPERATED RELIEF VALVES 

147. Notice 95-47 - UNEXPECTED OPENING OF A SAFETY/RELIEF VALVE AND 

COMPLICATIONS INVOLVING SUPPRESSION POOL COOLING STRAINER 

148. Notice 95-47 Rev. 1 - UNEXPECTED OPENING OF A SAFETY/RELIEF VALVE AND 

COMPLICATIONS INVOLVING SUPPRESSION POOL COOLING STRAINER 

149. Notice 95-53 - FAILURES OF MAIN STEAM ISOLATION VALVES AS A RESULT OF 

STICKING SOLENOID PILOT VALVES 

150. Notice 96-02 - INOPERABILITY OF POWER-OPERATED RELIEF VALVES MASKED BY 

DOWNSTREAM INDICATIONS DURING TESTING 

151. Notice 96-07 - SLOW FIVE PERCENT SCRAM INSERTION TIMES CAUSED BY VITON 

DIAPHRAGMS IN SCRAM SOLENOID PILOT VALVES 

152. Notice 96-42 - UNEXPECTED OPENING OF MULTIPLE SAFETY RELIEF VALVES 

153. Notice 96-48 - MOTOR-OPERATED VALVE PERFORMANCE ISSUES (Includes a discussion 

of the EPRI MOV Performance Prediction Program, which, according to EPRI is applicable to 

valves with air-operators)

NUREG/CR-6654 A-8
Appendix A

154. Notice 96-68 - INCORRECT EFFECTIVE DIAPHRAGM AREA VALUES IN VENDOR 

MANUAL RESULT IN POTENTIAL FAILURE OF PNEUMATIC DIAPHRAGM 

155. Notice 97-07 -PROBLEMS IDENTIFIED DURING GENERIC LETTER 89-10 CLOSEOUT 

INSPECTIONS

A-9 NUREG/CR-6654
Appendix B 

Air-Operated Valve Event Significance Analysis
Appendix B 

Air-Operated Valve Event Significance Analysis 

PURPOSE AND BACKGROUND 

The purpose of this analysis was to use precursor data to determine the relative significance that 

operational events involving air-operated valves (AOVs) or AOV-related support systems contributed to 

the potential for core damage accident sequences. 

The precursor results documented in the NRC-sponsored Accident Sequence Precursor Program 

(ASP) reports (NUREG/CR-4674, "Precursors to Potential Severe Damage Accidents...") for the time 

period 1984 through 1995 were used to determine AOV event significance. The ASP program involves 

the systematic review and evaluation of operational events or conditions that have occurred at U.S. 

commercial light-water reactors, reported under the requirements of 10 CFR 50.72 and 50.73. 

The ASP Program uses simplified risk models to provide estimates of operating event significance 

in terms of the potential for inadequate core cooling and subsequent severe core damage, i.e., as an 

accident sequence precursor. Identification of an operational event as an accident sequence "precursor" 

does not of itself imply that a significant potential for core damage existed. It does mean that at least one 

of a series of protective features designed to prevent core damage was compromised. 

An estimate of the likelihood of core damage is then obtained using PRA techniques, conditional 

on the occurrence of the operational event and the estimated failure probabilities of the remaining 

mitigative features. This estimate of the likelihood of core damage given such an operational event or 

condition is defined as the conditional core damage probability (CCDP). 

The ASP Program provides a methodology to document and evaluate those operational events or 

conditions that were determined to be risk-significant. The primary result of the ASP Program is the 

identification of operational events and conditions with CCDP's greater than or equal to 1.0 x 10-6 (ME-6) 

that satisfy at least one of four precursor screening criteria: 

* A core damage initiator requiring safety system response, 

The failure of a complete system required to mitigate the consequences of a core damage 

initiator, 

"* Degradation of more than one system required for mitigation, or 

"* A trip or loss of feedwater with a degraded mitigating system. 

In general, for the nuclear power industry, an accident sequence precursor is a sequence of events 

that did not progress to core damage, but if additional failures had occurred, would have resulted in 

inadequate core cooling and subsequent core damage. Those events with conditional probabilities of 

subsequent severe core damage greater than or equal to 1E-6 are identified and documented as precursors.

B-1 NUREG/CR-6654
Appendix B

Events or conditions originally identified through screening as potential precursors are subjected to 

an initial engineering evaluation. Generally, this evaluation eliminates events from further consideration 

as precursors if, for example, they involve only: 

* A component failure with no loss of redundancy, 

"* A short-term loss of redundancy in only one system, 

"* A structural degradation, 

"• An event that occurred prior to initial facility startup, 

"* A design error discovered by re-analysis, or 

"* An event with no appreciable impact on safety systems. 

Events identified for further consideration and detailed analysis include: 

* Unexpected initiators, 

* All events in which an initiating event occurred and a safety-related component failed, 

* All support system failures, 

* Any event in which two or more failures occurred, or 

* Any event or operating condition that was not predicted or that proceeded differently from 

the plant design basis. 

If an AOV-related event was potentially a contributor to a risk significant conditional core damage 

sequence (i.e., CCDP greater than or equal to 1E-6), and met the precursor criteria, there was a high 

likelihood (but not a certainty) that it was identified as a selected precursor event that was analyzed as 

part of the ASP program for the year in which it occurred. Although the simplified nature of the ASP 

models and reporting criteria for LERs result in the omission or exclusion of many plant AOVs (and thus, 

many AOV failures will not have been evaluated), some AOV failures were included in the ASP analyses. 

Evaluation of these AOV-related events provides a useful starting point to understand the risk 

significance of AOV failures. 

RESULTS OF THE ANALYSIS 

Table 1 provides a summary of the findings. Events with CCDPs greater than or equal to 1E-4 

have traditionally been considered to be the most important in the ASP Program. Only one event 

exceeded this range in 1995. Events with CCDPs less than 1 E-6 are considered as events of low 

significance and are not considered in the ASP methodology. There were 288 precursors that had CCDPs 

greater than or equal to 1 E-6 between 1984 and 1995, and 26 of these were AOV-related. There were 

89 precursors that had CCDPs greater than or equal to 1E-4 between 1984 and 1995, and 12 of these were 

AOV-related.

NUREG/CR-6654 B-2
Appendix B

The majority of AOV-related events during this time period involved loss of main feedwater, 

requiring either automatic or manual trip of the reactor including: 

"* Several events involving loss of feedwater have occurred due to problems with the 

instrument air supply to air-operated main feedwater regulating and control valves. 

"* Loss of feedwater control occurred due to electrical failures in air-operated feedwater 

regulating valve control circuitry, in some cases, due to moisture intrusion. 

A significant number of potential common-cause mechanisms (12 events out of the 26 AOV

related events had some common-cause implications) were observed for this set of AOV-related failures 

including: 

"* Water or other contamination in instrument air supply lines to air-operated components (4 

events), 

"* Design and installation errors that caused air-operated component failures or operation in a 

non-fail-safe configuration (4 events), and 

"* Moisture in electrical components such as feedwater regulating valve pneumatic controllers 

(1 event). 

The AOV-related failures analyzed in this report impacted many plant safety and non-safety related 

systems including: 

"* Scram system (1 event), 

"* Main feedwater and auxiliary feedwater systems (8 events), 

"* Main condenser and MSIVs (2 events), 

"* Core Spray (1 event), 

"* Service water/ Component Cooling Water (2 events), 

* Diesel-Generators (2 events), 

"* ADS (3 events), 

"* Chemical Volume and Control system (1 event), 

"* PORVs (4 events), and 

0 ECCS Recirculation (1 event). 

Table 2 provides totals of the number of AOV-related precursor events by year that were analyzed 

in the ASP reports. The relative ranking of the AOV-related events based on CCDP, as compared to all 

precursor events analyzed for a given year, is also included in Table 2. Twenty-six events (about 9%) of 

a total of 288 events were AOV-related.

B-3 NUREG/CR-6654
Appendix B

Table 3 provides yearly totals of the number of AOV-related precursor events with CCDPs greater 

than or equal to 1E-4, as compared to all precursor events with CCDPs greater than or equal to 1E-4. 

Twelve events (about 13%) of the total of 89 events with CCDPs greater than or equal to 1E-4 were 

AOV-related. 

As noted in the precursor reports, a direct comparison of CCDP results between years should be 

approached with caution because CCDP estimates have become more refined over time due to analysis 

differences in the selection and modeling of events. 

Following are descriptions of five AOV-related events which were evaluated in the ASP Program 

and identified as precursors of high risk significance. 

Turkey Point 3 (LER No. 250/85-021) 

This event occurred at Turkey Point 3 in 1985, and had an estimated CCDP of 8.96E-4, which 

ranked it as the fourth highest event out of forty precursor events identified for that year. The event 

involved multiple AOV failures, primarily due to moisture in the instrument air system. 

While in hot standby and recovering from a prior reactor trip caused by a lightning strike, a Main 

Feedwater bypass valve failed to open. A low steam generator level resulted in an Auxiliary Feedwater 

auto-start signal, but AFW trains A and C failed on demand due to governor problems. Train A was 

restarted but during the subsequent recovery, a MFW bypass valve failed to close, resulting in a high SG 

level and trip of the only operating MFW pump. AFW auto-started but train B's flow control valve failed 

open. Later, during restart, two AFW train flow control valves failed to open during testing. The air

operated valves failed because of moisture in the instrument air lines. 

Oyster Creek I (LER No. 219185-012) 

This event occurred at Oyster Creek 1 in 1985, and had an estimated CCDP of 2.28E-4, which 

ranked it as the seventh highest event out of forty precursor events identified for that year. 

Subsequent to a reactor scram, one of the two Scram Discharge Volumes (SDV) did not fully 

isolate. The resulting flow of hot water from the reactor through the Scram Discharge Volume caused 

steam and paint fumes to discharge in the reactor building. This, in turn, activated the deluge fire system 

on one level of the reactor building. The plant experienced a non-isolatable leak of reactor fluid outside 

primary containment. 

One SDV drain valve that isolates the Scram Discharge Volume bottomed out before the valve was 

fully seated because the stroke adjustment was improperly set. The actuator spring on a second drain 

valve was improperly sized and opened slightly when pressure from the first valve was applied to its seat. 

With both valves partly open, the leak was established. 

Fort Calhoun (LER No. 285/87-025) 

This event occurred at Fort Calhoun in 1987, and was estimated to have a CCDP of 6.2E-4, which 

ranked it as the second highest event out of thirty-three precursor events identified for that year. 

During conversion of the diesel generator (DG)-room fire system to a dry-pipe water system, water 

inadvertently entered the Instrument Air system, due to check valve problems. Upon discovery, the 

system was repaired and the instrument air system blown down to remove the water. Individual 

components with safety-related air accumulators were also blown down and drained. However, the

NUREG/CR-6654 B-4
Appendix B

emergency DG radiator exhaust damper accumulators, and the AOVs which control the radiator dampers 

were overlooked. 

During a subsequent test, the DG 2 automatically shut down due to high coolant temperature. The 

AOV controlling radiator exhaust dampers failed to fully open because of the prior water intrusion 

incident. The lack of full air flow through the radiator resulted in the high coolant temperature. The pilot 

orifice valve was found to be blocked by foreign material, from the interaction of O-ring lubricant and 

water. The backup air accumulator for the pilot valve was found to contain 50% water (2 quarts). Water 

was subsequently found in the accumulator for DG 1. 

If a loss of offsite power event had occurred while the DG radiator damper AOVs and their 

accumulators were inoperable, both DGs would have run to destruction because the high coolant 

temperature shutoff would have been bypassed during a real demand. 

Peach Bottom 3 (LER No. 278/91-017) 

This event occurred at Peach Bottom 3 in 1991, and was estimated to have a CCDP of 3.3E-4, 

which ranked it as the seventh highest event out of twenty-seven precursor events identified for that year. 

Improperly installed insulation on the automatic depressurization system (ADS)/safety relief valves 

(SRVs) resulted in damage to SRV control wiring. The ADS main steam SRVs comprise 5 of the 

11 MSRVs. An investigation determined that the MSRV thermal insulation was installed incorrectly 

during a previous refueling outage. The temperature increase from the installation error caused the 

expiration of the equipment qualification life of the components after approximately 3 days of operation. 

Haddam Neck (LER No. 213/94-005) 

This event occurred at Haddam Neck in 1994, and was estimated to have a CCDP of 1.4E-4, which 

ranked it as the second highest event out of the nine precursor events identified for that year. 

During testing, it was discovered that the air operators for the pressurizer power-operated relief 

valves (PORVs) were experiencing control air leaks and that the PORVs could not be operated properly 

from their safety-grade control air supply. Investigation revealed that repairs to fix a prior PORV failure 

(see 1993 event in Table 1) were made incorrectly during the previous refueling outage. The PORV 

diaphragms were not seated correctly and were coated with a lubricant rather than a required sealant. 

Substantial air leaks resulted, and the PORVs could not be opened more than 50%. 

The LER for the event indicates that two safety functions were potentially compromised by the 

PORV failures: feed-and-bleed cooling and high-pressure safety injection (HPSI) makeup during certain 

small-break LOCAs.

B-5 NUREG/CR-6654
Table B-1. Summary of AOV-related precursor analysis results. 

CCDP Contribution 

Events 

With CCDP 

Events with > le-6 & Plant Status at Time of Event; Year LER Number/Plant Name CCDP > le-4 <le-4 Ranking Method of Discovery Brief Event Description

61 

z 

9x

FERRY 1

1984 325/84-006/BRUNSWICK 1 2.60e-04 

1984 325/84-014/BRUNSWICK 1 1.20e-04 

1984 331/84-001/DUANE 1.20e-04 

ARNOLD 

1984 387/84- 1.40e-04 

010/SUSQUEHANNAH 1 

1885 219/85-012/OYSTER 2.28e-04 

CREEK 1

1985 220/85-021/NINE MILE 

POINT 1

6/33 At Power; Operational event 

15/33 At Power; Operational event 

16/33 At Power; Operational event 

13/33 At Power; Testing 

7/40 At Power; Operational event 

7.25e-06 25/40 At Power; Operational event

IY84 298/4-U32/iJKUWNS 6.60e-06 28/33 At Power: Operational event - • -- IV ................

0*

Low Pressure Core Spray testable check valve's 

air operator was installed backwards and resulted 

in inadvertent over pressure and leakage from 

system relief valve during MOV test while at 

power. 

Operators working on instrument air system 

caused MFW to trip off and resulted in a plant 

trip; would also have affected Unit 2 had it been 

operating. 

Following reactor trip and Group 1 isolation 

signal, a MSIV failed to close when its air

operated solenoid pilot valve failed. 

Loss of feedwater occurred when FW 

recirculation valve failed open due to a broken 

air supply line fitting. 

Manual scram was required when ADS SRV 

failed to close during testing when its air

operated solenoid pilot valve failed. 

One SDV drain valve that isolates the Scram 

Discharge Volume bottomed out before the valve 

was fully seated because the stroke adjustment 

was improperly set. The actuator spring on a 

second drain valve was improperly sized and 

opened slightly when pressure from the first 

valve was applied to its seat. With both valves 

slightly open, the leak was established. 

A problem with Instrument Air caused 

malfunction of a feedwater flow control 

valve resulting in a reactor trip.
Table B-1. (continued). 

CCDP Contribution 

Events 

With CCDP 

Events with > 1 e-6 & Plant Status at Time of Event; 

Year LER Number/Plant Name CCDP >_ le-4 <le-4 Ranking Method of Discovery Brief Event Description

1985 250/85-021/TURKEY 

POINT 3 

1986 247/86-017/INDIAN POINT 

2 

1987 285/87-025/FORT 

CALHOUN 

1987 440/87-009/PERRY 1 

1988 280/88-01 1/SURRY 1 

1988 346/88-007/ DAVIS BESSE 

1

8.96e-04 

1.00e-04 

6.20e-04 

2.30e-04

1988 369/88-007/ McGUIRE 1 

1989 317/89-005/CALVERT 

CLIFFS 1 

1989 338/89-005/ NORTH ANNA 1.90e-04 

1

4/40 Routine Shutdown; 

Operational event 

6/18 At Power; Operational event 

2/33 At Power; Testing 

8/33 At Power; Testing 

1.50e-05 17/32 Routine Shutdown; 

Operational event 

1.60e-06 31/32 At Power; Maintenance 

1.00e-06 32/32 At Power; Operational event 

1.40e-06 27/30 Routine Shutdown; Testing 

5/30 At Power; Operational event

Aux. Feedwater flow control valves and MFW 

)ypass valves malfunctioned due to moisture in 

Instrument Air lines. 

Safety Injection Actuation & Reactor Trip was 

-aused by condenser steam dump valves opening 

n a faulty steam dump controller signal. 

DG2 shutdown due to high coolant temperature 

-aused by a stuck air flow pilot valve for the 

*-adiator exhaust dampers. Water was found in the 

ir accumulator of DG 1 & 2 pilot valves. 

Control air solenoid valves failed resulting in 

inoperability of two DGs. 

Two PORVs failed to open due to improper 

torque on actuator diaphragm bolts. 

Leakage of air from service water valve 

accumulators after prolonged loss of instrument 

air could have resulted in loss of SW. 

A FW regulating valve failed closed due to 

controller failure resulting in need for reactor 

trip. 

Instrument air boundary check valve leak could 

result in saltwater pump runout during 

containment sump recirculation after a LOOP. 

Reactor tripped due to MFW regulating valve 

closure caused by loss of instrument air supply.

z 

61
Table B-1. (continued). 

CCDP Contribution 

Events 

With CCDP 

Events with > le-6 & Plant Status at Time of Event; Year LER Number/Plant Name CCDP > le-4 <le-4 Ranking Method of Discovery Brief Event Description 61 (-I

1989 458/89-022/ RIVER 

BEND 1 

1990 206/90-006/SAN 

ONOFRE 1 

1990 285/90-025/FORT 

CALHOUN 

1990 293/90-013/PILGRIM 

1991 287/91-007/OCONEE 3 

1991 272/91-030/SALEM I

1989 530/89-001PALO VERDE 3 4.90e-05 9/30 At Power; Operational event 

1.30e-05 18/30 Refueling; Testing 

6.00e-05 9/28 At Power; Testing 

1.70e-06 24/28 At Power; Operational Event 

8.40e-05 8/28 At Power; Operational event 

1.80e-05 19/27 At Power; Operational event 

4.40e-06 24/27 Routine Shutdown; Testing

ADVs at PV3 did not actuate from either the 

control room or the remote shutdown panel 

following a transient and reactor trip. One ADV 

was manually opened and another was damaged 

in an unsuccessful attempt to open it. 

Investigation of the ADVs at PV 1, 2, and 3 

revealed common-cause design problems with all 

the ADVs, as well as moisture contamination in 

the air systems and inadequate accumulator 

capacity. 

Instrument air solenoid isolation valves were 

installed backward and would impact long-term 

ADS operations. 

Chemical volume and control system pneumatic 

valve installed to fail open instead of closed on 

loss of IA. This could have resulted in gas 

binding of charging pumps. 

Due to cross ties between CCW and Raw Water 

system by normally closed, air-operated valves, 

loss of IA would cause valves to open and drain 

CCW. 

Manual scram after moisture caused an electrical 

short led to loss of the feedwater regulating 

valves. 

Reactor trip occurred after LOFW from clogged 

instrument air flow in master valve controller for 

condensate demineralizer system. 

Both PORVs failed due to leakage from their air

operated actuators.

tz
Table B-1. (continued). 

CCDP Contribution 

Events 

With CCDP 

Events with > 1 e-6 & Plant Status at Time of Event; 

Year LER Number/Plant Name CCDP > le-4 <le-4 Ranking Method of Discovery Brief Event Description 

1991 278/91-017/PEACH 3.30e-04 7/27 Refueling; Maintenance Improperly installed insulation on ADS SRVs 

BOTI[OM 3 resulted in damage to the SRV control wiring. 

1993 213/93-007/HADDAM 6.50e-05 5/16 At Power; Maintenance Leak in diaphragm assembly of PORV would 

NECK have prevented feed & bleed cooling following 

loss of containment air compressors. 

1994 213/94-005/HADDAM 1.40e-04 2/9 At Power; Testing During testing it was determined that both 

NECK PORVs were leaking from their air-operated 

actuators.

6N
Appendix B

Table B-2. Totals of ASP analyzed LER events and AOV-related events with CCDP > 10-6 per year. 

Total No. of AOV

Total No. of All Related Events 

Events With With CCDP > 10-6 AOV Event Ranking 

Year CCDP > 10-6 Year Within Year 

1984 33 5 6,13,15,16,28 

1985 40 3 4,7,25 

1986 18 1 6 

1987 18 2 2,8 

1988 32 3 17,31,32 

1989 30 4 5,9,18,27 

1990 28 3 8,9,24 

1991 27 3 7, 19,24 

1992 27 0 

1993 16 1 5 

1994 9 1 2 

1995 10 0 

Totals 288 26

NUREG/CR-6654 B-10
Appendix B

Table 3. Totals of ASP analyzed LER events and AOV-related events with CCDP > 10.4 per year.

Year 

1984 

1985 

1986 

1987 

1988 

1989 

1990 

1991 

1992 

1993 

1994 

1995 

Totals

NUREG/CR-6654

Total No. of All Events 

With CCDP > 10. 

17 

10 

5 

10 

7 

7 

6 

13 

7 

4 

2 

1 

89

0 

1 

0 

12

Total No. of Aov-Related 

Events With CCDP > 10. 

4 

2 

1 

2 

0 

1 

0 

1 

0

B-11I
Appendix C 

Trip Reports
Appendix C

TRIP No. 1 REPORT 

STUDY OF AIR-OPERATED VALVES 

PALO VERDE, OCTOBER 27 AND 28, 1997 

We had two days of meetings and interviews with the engineers at Palo Verde who are concerned 

with AOVs. We also were shown portions of the air system and valves served by it. 

Palo Verde has a dedicated AOV service group comprised of engineers and technicians. The 

engineers provided us with two notebooks, prepared for our visit, that described their program and some 

of the pertinent problems and studies regarding AOVs. These are being reviewed as part of this AOV 

study. The contents of the notebooks included a description of their AOV program, examples of AOV 

sizing calculations, and several root cause analyses of failures of AOVs to perform as intended. 

The cooperation, courtesy, and knowledgeable responses from the members of the Palo Verde staff 

were noted and appreciated by those of us who are involved in this study of AOVs. 

Each of the plants at Palo Verde has an independent service air system that also provides a source 

to the instrument air system. The systems are described in a "Compressed Gas System Evaluation and 

Analysis Report," (13-MS-A20, Revision 2, dated June 15, 1989, Revision 3, undated). This report is 

used as a working document for analyzing and maintaining the air system. The air systems each consist 

of an instrument air subsystem and nitrogen subsystem. The nitrogen system is used when the air header 

pressure drops below 85 psig. The air/nitrogen systems are not safety-related and safety-related items 

supplied by them are augmented by backup accumulators. 

The AOVs at Palo Verde are categorized as I, II, or III. Category I includes active, safety-related 

AOVs. Category II includes safety-related AOVs not in category I. Category III includes all other 

AOVs. There are 41 category I AOVs in each unit (123 total for the site) and 131 category II AOVs (393 

for the site). We were told that there are a total of approximately 8400 AOVs on the site. For 

comparison, there are 831 motor-operated valves on the site (277 per unit), of which 336 are safety

related and in the Generic Letter 89-10 program (112 per unit). 

The AOV engineers provided detailed root-cause analysis reports and information on four of the 

major problems that they had encountered involving AOVs at Palo Verde. These were: 

Atmospheric Dump Valve (ADV) Failures (see LER 52889005), 

Letdown Containment Isolation Valve Leakage, 

Downcomer Feedwater Isolation Valve Failures, and 

Failure of a Vacuum Breaker Solenoid and Subsequent Investigations Involving SOVs. 

The ADVs were subjected to evaluation, as well as diagnostic and dynamic testing after the failures 

occurred. Excessive piston ring leakage, combined with inadequate pilot valve relieving capacity, created 

high forces in the valve bonnet (also called the balance chamber) that could not be overcome by the 

actuator. Other problems were also found that compromised the operability of the ADVs. These 

included:

C-1 - NUREG/CR-6654
Appendix C

- Valve oscillations caused by lower than required nitrogen pressure (the regulators exhibited 

excessive leakage); 

- Positioners that were not adjusted and/or maintained properly; 

- Springs that were left on the valve operators should have been removed prior to startup; 

- An actuator piston that was fitted with a non-qualified Buna-N rather than Viton 0-ring; 

- Air and nitrogen quality that was suspect (particulate contamination); and 

- Several non-qualified pressure gages that were left installed on the positioners. 

PVNGS completed an investigation, in 1995, of the recurring seat leakage over several years, of 

three letdown containment isolation valves. LER 52895007 (a previous LER 52894009 also applies) and 

PVNGS Condition Report 95Q028 of 5/11/95 describe the results of the investigation. The most probable 

causes of the seat leakage problems were: 

undersized pneumatic actuators resulting from not accounting for the high frictional loads of 

graphite style packing during original sizing of the actuators; and 

not maintaining the specified bench set on the spring-and-diaphragm actuators. 

Three downcomer feedwater isolation valves (DCFWIVs) at PVNGS failed to open, following 

closure after a main steam isolation signal (MSIS) during a Unit 1 reactor trip on 11/26/95. The most 

probable root causes for the multiple valve failures to open were: 

- a lack of prudent actuator design margin. The low actuator margin resulted from using a non

conservative valve factor (0.3) in the original actuator/valve manufacturer (Anchor Darling) 

sizing. Recent tests on motor-operated gate valves indicate that 0.3 is not a conservative valve 

factor for flex-wedge gate valves. 

- not allowing for the potential effects of thermal binding in the original sizing of the actuator.' 

- not allowing for the potential effects of degradation of the nitrogen supply in the calculation of 

actuator margin in the original valve/actuator manufacturer (Anchor Darling) sizing. 

On November 26, 1995, a loss of condenser vacuum condition occurred at Unit 1 of PVNGS and 

caused a main turbine trip. A reactor trip followed shortly thereafter. The event was initiated when a 

condenser vacuum breaker inadvertently opened. Two similar SOVs on Unit 2 were found to have air 

leaks. The root cause of the SOV failure was attributed to aging. 

Palo Verde conducted extensive research, root cause analyses, static testing, diagnostic testing, and 

dynamic testing in their analysis of the events that are described briefly above. 

'Note: Pressure locking and thermal binding of power-operated gate valves, including 

AOVs, is discussed in NRC Generic Letter 95-07. GL 95-07 was not applied by PVNGS to gate 

valves whose safety-related function is to close, such as the DCFWIVs.

NUREG/CR-6654 C-2
Appendix C

Among the observations made during the visit to PVNGS for this study, it was noted that several 

safety-related AOVs had low margins, according to the calculations that were furnished for discussion. 

Specifically: 

Globe valves: The licensee's engineers were using the port area times the differential pressure 

(DP) times 1.0 in their thrust estimate calculations. Instrument Society of America guidance is to 

use the port area times d/P times 1.0. Motor-operated valve testing for other types of globe 

valves would indicate a valve factor of 0.9 to 1.1. Depending on the particular valve design, the 

licensee thrust estimates may be as much as 10% low. 

It is important to note that the air-operated globe valves used at Palo Verde have not been 

dynamically tested in a Generic Letter 89-10 type of program to verify valve factors. Also the 

licensee has only found physical valve/actuator performance issues with actuators (for globe 

valves) which were not sized in accordance with the original equipment manufacture's published 

recommendations. 

Gate Valves: The licensee's engineers were using the seat area (as provided by the valve 

manufacturer) times d/P times a valve factor of 0.6 in their thrust estimate calculations. A valve 

factor of 0.6 is reasonable for cold water systems. INEEL and EPRI test results both support the 

use of a mean seat area [0.5 X (inside diameter + outside diameter)]. 

NOTE: Palo Verde suggested the above wording in their comments on the draft report and 

provided the following explanation as justification for the gate valve discussion: The Palo Verde 

engineers had originally used the port area for the thrust estimates but had revised the calculation 

to use the seat area prior to the site visit. The calculation shown to the authors of this study 

included an error in that some text explaining the calculation indicated the port area was used 

rather than the seat area. This text was missed as part of the calculation revision to use seat area 

and has since been revised. The quantitative part of the calculation reviewed by the authors 

actually used the seat area even though the text indicated the port area had been used. 

The quality of the air supply was monitored intermittently for moisture content. 

The following tabular summary was prepared to describe the information gathered during the 

subject visit. The NRC, with assistance from the INEEL, is studying the performance of air-operated 

valves (AOVs) in commercial nuclear power plants. The information was collected in accordance with 

the Program Plan dated 10/22/97 (INEEL Letter to H. Ornstein, NRC, from J. Bryce, 10/23/97, Job Code 

E8238, Task Order 15 - JHB-167-97).

C-3 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO PALO VERDE NUCLEAR GENERATING STATION

ITEM No. INFORMATION RESPONSE OR INFORMATION 

1 Dates. October 27 and 28, 1997. 

2 Names of Interviewers. Hal Ornstein, NRC/AEOD, 301/415-7574 

Owen Rothberg, INEEL/Rockville, 301/816-7773 

John Watkins, INEEL/Idaho Falls, 208/526-0567 

3 Plant Name & Docket No. Palo Verde 1, 2, and 3, Dockets 50-528, 529, and 530 respectively. 

4 Person(s) Interviewed, Martin Grissom, Licensing, 602/393-5744 

Title(s), Phone Number(s), Mike Renfroe, Design Engineering, 602/393-1914 

E-Mail address, short 

description of Sonja Waters, Design Engineering, 602/393-1935 

organization(s) and duties. Steve Quan, 1ST, 602/393-6215 

Mike Hooshmand, Valve Services Engineering, 602/393-1090 

Dave Stricker, Valve Services Engineering, 602/393-1938, 

FAX 602/393-1854 

Jim Minnicks, Valve Services Department, 602/393-1070 

Benny Malekzadeh, Valve Services Engineering, 602/393-1026 

Katie Clifton, Valve Services Engineering, 602/393-1085 

Lonnie Bullington, PRA, 602/393-6523 

Tim Mitchell, HVAC Maintenance Engineering, 602/393-3541 

John Glover, IA Systems Engineering, 602/393/6254 

Steve Coppock, Technical Assistant Vice President, Engineering 

William Ide, Vice-President Engineering 

5 If necessary, and if person(s) A notebook was provided that describes the AOV Program at PVNGS and 

interviewed can do so, provides detailed valve lists, sample calculations, failure analyses reports 

obtain any missing for 1995, and other data. 

information not provided A notebook was provided that describes significant events and issues 

prior to the site visit, as involving AOVs at Palo Verde. 

described in the outline for 

Task 4 above. Note what A copy of the Palo Verde "Compressed Gas System Evaluation and 

information was provided. Analysis Report" was provided. This document was originally prepared to 

evaluate and analyze a March 1989 event involving loss of power to the 

compressed gas system. We were told that this has become something of a 

"living" reference document for the technical staff involved with 

compressed gas systems.

NUREG/CR-6654 C-4
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO PALO VERDE NUCLEAR GENERATING STATION

ITEM No. INFORMATION RESPONSE OR INFORMATION 

6 Describe plant events In March 1989, a reactor trip resulted in loss of IA compressors and 

involving AOVs and provide degradation of air pressure. Backup nitrogen did not restore pressure. 

reference information, if Investigation revealed a number of problems including leaks and 

possible. unanalyzed demands on the IA system. See reports furnished by PVNGS. 

Recent: The ADVs did not function as designed and a separate report noted a 

number of design and QA problems with the valves. These problems are 

Recurring: considered significant and resulted in a 10 CFR Part 21 report by the 

Significant: licensee. 

In October 1995 a report describing excessive seat leakage of three 

letdown isolation valves, in events over the previous three years, was 

prepared. The root cause problem appeared to be undersized actuators 

attributable to packing loads from graphite packing not originally 

accounted for. A copy was furnished during the trip. 

An investigation of the failure of three downcomer feedwater isolation 

valves to open in November 1995 was provided. The root cause appeared 

to be undersized actuators attributable to a not conservatively chosen valve 

factor. 

In November 1995, a loss of condenser vacuum led to a reactor trip at Unit 

1. This event was caused by a condenser vacuum breaker failing open, 

which was caused by leakage of the SOVs. The SOV failures were 

attributed to aging. 

7 Describe AOV or air system See Component Failure Analysis Report (CFAR) of 9/21/95 for an 

actual or detected potential analysis of failures. In addition, it was determined that several AOVs have 

failures at the plant? Provide margins that are below standard (DCFWIVs). PVNGS is examining 

reference information, if options to increase the margin. 

possible. However, globe and gate valve thrust prediction methods that had not 

been validated and that had been shown to be nonconservative per MOV 

testing were, at the time of the visit, being used. 

8 Describe actions taken after See the Compressed Gas System Evaluation and Analysis Report that was 

events or failures involving generated after the March 1989 event and is maintained as a working 

AOVs or the air system. document by the plant staff concerned with compressed gas systems. A 

Provide reference number of recommendations mostly having to do with keeping the air 

information, if possible. system clean were included. 

In-plant dynamic diagnostic testing of an AOV was performed. 

9 Were there any actual or The March 1989 event revealed a number of common cause failure 

potential common mode or mechanisms including debris in the IA and nitrogen backup system, 

common cause failures in unanalyzed demands on the IA system, and faulty check valves in the 

the air system or AOVs at accumulators. 

the plant? Describe and Globe and gate valve thrust prediction methods that had not been validated 

provide reference and that may have been nonconservative per MOV testing were, at the 

information, if possible. time of the visit, being used.

C-5 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO PALO VERDE NUCLEAR GENERATING STATION

ITEM No. INFORMATION RESPONSE OR INFORMATION

Packing adjustments were poorly controlled and their effect on operability 

was not always known.

Performance assumptions (such as the thrust or diaphragm area versus 

stroke) of actuators is being questioned by some in the industry. 

The diagnostic test equipment uses the effective diaphragm area as an 

assumption. The same assumption is used in the actuator verification 

calculation. This unverified assumption could lead to a common-cause 

nonconservative seating load.

10 Describe root cause analysis Formal root cause analyses are conducted by PVNGS and examples were 

procedures for the plant. furnished. The engineering staff has applied standard engineering 

Provide reference methods to determine root causes of valve, IA, and nitrogen systems 

information, if possible. deficiencies and these appear to be well researched and comprehensive. 

11 Describe root-cause analyses Refer to the documentation and calculations provided by PVNGS. The 

performed for air system or root causes of AOV failures appear to have been competently investigated. 

AOV failures at the plant. A dynamic test with diagnostic equipment was performed on one valve as 

Provide reference part of this effort. SOV failures were not tracked, in some cases, because 

information, if possible. PVNGS replaced failed SOVs as piece-parts. 

12 Describe maintenance Air quality with respect to water vapor and hydrocarbons is monitored 

procedures for the air semi-annually (recently changed from quarterly and considering annually). 

system. Provide reference Particulate contamination is controlled by scheduled filter replacement and 

information, if possible. monitoring three micron filters in the air system. Dew point monitoring is 

available at the dryer outlets via a tell-tale color change; however, the 

operators are not required to record color changes on their rounds sheets. 

13 Describe maintenance A dedicated team of engineers and technicians is assigned to maintenance 

procedures for AOVs. of AOVs. The AOVs are monitored in accordance with their ranking as 

Provide reference Category 1 (active safety-related), Category 2 (inactive safety related), or 

information, if possible. Category 3 (non-safety-related). In addition, PVNGS has categorized 

AOVs as either active safety-related, passive safety-related, or important

Safety-related: to-safety, in response to the Maintenance Rule, 10 CFR 50.65. 

Important non-safety

related: 

Non-safety-related: 

14 Describe IST procedures for The air system is monitored semi-annually for particulates, water 

the air system. Provide contamination, and hydrocarbons. The accumulators for the four ADVs 

reference information, if are monitored under plant surveillance test 73SP-9SG05. The remaining 

possible. accumulators are not monitored. 

15 Describe IST procedures for ASME Section XI, stroke-timing testing is conducted for safety-related 

AOVs. Provide reference AOVs. Most of these tests are conducted without process fluid loads 

information, if possible. (pressure and/or flow). The ADVs have been stroked with fluid flow or 

Safety-related: pressure in the lines although it is not known if this is a standard 

S surveillance procedure.

NUREG/CR-6654 C-6
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO PALO VERDE NUCLEAR GENERATING STATION 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

Important non-safety- No regular IST schedule was observed for non-safety-related AOVs. 

related: 

Non-safety-related: 

16 Describe diagnostic systems, The Fisher FlowScanner AOV diagnostic system was used at Palo Verde 

if any, used for AOVs. at the time of the visit. The engineers indicated that the system was 

Provide reference somewhat obsolete and PVNGS is investigating purchase of a more 

information, if possible. modem AOV diagnostic system. Several vendors have presented systems Description of system: but the plant engineers had not made a decision at the time of the visit. A detailed description of the diagnostic systems used for AOVs was not 

Specifications: obtained, although the diagnostic system in use at the time of the visit was 

Data collected and discussed with the plant engineers and the training mock-up was viewed. 

frequency of collection: A copy of the procedure used to bench test AOVs was provided. The 

Fisher FlowScanner system is used routinely to determine various 

Vendor assistance provided, operating characteristics of safety-related and non-safety-related AOVs at 

if any: PVNGS. 

17 Describe design (and Design bases estimates and calculations of margin were provided by 

analysis) procedures for PVNGS for active safety-related AOVs. The demand calculations are 

AOVs. Describe how based on estimates of the loads to be encountered. In several cases, the 

design basis is established margins appear to be quite narrow (low or negative on Calculation Sheet 

and maintained for AOVs. 13-MC-ZZ-219, page 4 of 18) for the downcomer feedwater isolation 

Provide reference valves. The PVNGS engineering staff was working on the problem. 

information, if possible. Globe and gate valve thrust prediction methods that had not been validated 

and that may have been nonconservative per MOV testing were, at the 

time of the visit, being used. 

18 Describe analyses and/or PVNGS engineers conducted a number of analyses to estimate the margins 

testing for verification of available for AOVs at design basis conditions. They have also done some 

operability during postulated bench testing and diagnostic evaluations that resulted in hardware 

transient or accident modifications. They are working to improve the margins on AOVs that 

conditions. Provide they consider important. 

reference information, if 

possible. 

19 Describe training for Training on mock-ups is in place. No flow loop is available on-site. 

installation, maintenance, Diagnostic testing is done using the Fisher FlowScanner system. Valves 

and testing of AOVs. are tested in accordance with IST requirements and sometimes when 

Provide reference anomalous behavior is noted. 

information, if possible. 

20 Describe databases used to PVNGS maintains records of valve behavior in several databases but does 

track maintenance, failures, not maintain a maintenance/performance history database on each valve. 

and events regarding AOVs. The engineers are aware of the performance and performance 

Provide reference characteristics of the hardware in the plant. 

information if possible.

C-7 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO PALO VERDE NUCLEAR GENERATING STATION

ITEM No. INFORMATION RESPONSE OR INFORMATION 

On site: 

Company wide: 

Industry: 

21 Describe the impact of the The Maintenance Rule required ranking of systems within the plant in 

Maintenance Rule, 10 CFR terms of risk significance. Several AOVs that had not been considered 

50.65 on AOV and air significant in the past are now regarded as being important and the AOV 

system maintenance and group is more focused on the performance of these valves than previously. 

testing. Provide reference 

information if possible. 

22 Is PRA data used for According to the "AOV Maintenance Program Description Outline" 

predictive maintenance or provided during the visit, PRA rankings are to be used to identify 

replacement of AOVs? If important AOVs. The schedule for accomplishing this was not defined as 

so, how? of the time of the visit. 

23 Are AOVs serviced on site, AOVs are generally serviced on site. SOVs are generally replaced as 

serviced offsite, or replaced piece-parts when they fail. On occasion, for important events or if 

as piece-parts if found to assistance is needed to solve a problem, valves are sent to the manufacturer 

require service? for repair or analysis. 

24 Identify and describe the The failure incident report showed a number of problems and concerns 

most common recurring regarding the air system and AOVs, particularly with regard to ADVs. We 

maintenance problem(s) and were concerned that there was no continuous monitoring of moisture in the 

failures regarding AOVs and air system, although there are indications that such monitoring could 

the air system. What did you improve equipment performance. 

see? Provide reference 

information if possible. 

25 Interviewer comments The station air system and components in the turbine building were 

regarding actual valves viewed. Several turbine control and feedwater control valves were also 

viewed during the visit, in viewed. The shops and training facilities were visited. 

the plant, undergoing 

maintenance or replacement, 

or in the material, if 

applicable to this interview. 

26 Has the plant made changes Yes. See engineering evaluation reports provided on atmospheric dump 

to valves or systems that valves, letdown isolation valves, downcomer feedwater isolation valves, 

include AOVs, or replaced and loss of condenser vacuum. 

AOVs with different models 

of AOVs or different valves 

that are not AOVs? If so, 

describe the changes and the 

circumstances. 

I What prompted the change? 1

NUREG/CR-6654 C-8
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO PALO VERDE NUCLEAR GENERATING STATION 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

Was the change made for 

this plant only? 

27 Does the plant follow The AOV engineers at Palo Verde are familiar with the EPRIINMAC 

EPRI/NMAC guidelines for guidelines and other industry literature. According to their remarks, the 

maintaining AOVs and the EPRI/NMAC guidelines provided several insights for their program. 

air system(s)? If not, 

describe differences and 

reasons for the differences. 

Provide reference 

information, if possible. 

28 What is the plant doing or PVNGS is actively involved in improvements to their air system and 

planning to do in response to AOVs. They have dedicated a specific group of engineers and 

the recent Industry technicians. They are aware of the industry work on the subject and 

correspondence on AOVs. indicated that they are open to suggestions for improvement. 

Provide reference 

information, if possible. 

29 Do you have any The PVNGS personnel consider that they are aggressively pursuing 

suggestions for improving performance and maintenance of AOVs in the plant. 

the performance of AOVs, AOVs on auxiliary equipment such as the diesel generators and in the 

particularly in the areas of HVAC system are not under the direct control of the AOV group. They surveillance, testing, or mareintance? 

maintenance? tare covered separately by the EDG and HVAC engineers respectively. 

30 Provide a list of 10 CFR A number of LERs and plant internal evaluations were discussed during 

50.59 and 10 CFR 50.72 the visit. A notebook was provided by the plant engineers describing 

reports on AOVs and AOV analyses of the atmospheric dump valves, failures of letdown isolation 

support systems (air or inert valves, downcomer feedwater isolation valves, and condenser vacuum 

gas supply, etc.) that have breaker event. 

been issued for this plant. 

31 What thrust or torque The plant engineers indicated that PVNGS does not have a specific margins are expected for numerical requirement for a margin. A margin in the range of 25% was 

AOVs? Are different informally discussed by the engineers as possibly being a desirable high margins used for safety- end goal. (Note: This question was asked during the interviews to get related, important non- some idea of how knowledgeable engineers viewed expectations of valve 

safety-related, or non-safety- margins and was not an attempt to establish some arbitrary standard for 

related AOVs? any margin.) In some cases, the valves fall far short of this margin, 

although we were assured that the valves meet all regulatory requirements. 

Refer to Anchor Darling Gate Valve calculations (13-MC-ZZ-219, page 4). Revision 1 of the calculations stated that the selected valves had been 

re-analyzed using new valve data provided by Anchor Darling. The 

calculations refer to a "port" area in some places and a "seat/port" area in 

others. If the port area was used, the resulting thrust requirements may be 

under predicted.

C-9 NUREG/CR-6654
Appendix C

NUREG/CR-6654

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO PALO VERDE NUCLEAR GENERATING STATION 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

Globe and gate valve thrust prediction methods that had not been validated 

and that may have been nonconservative per MOV testing were, at the 

time of the visit, being used. 

32 What maintenance or The accumulators for the ADVs (4 per plant) are monitored under plant 

surveillance is done to AOV surveillance test 73SP-9SG05. The remaining accumulators are not 

accumulators to ensure monitored. Several accumulator check valves have been repaired or 

air/nitrogen quality and replaced because of excessive seat leakage. 

pressure? Were seismic 

considerations and size 

_ _ verified? 

33 Describe problems with None noted during visit. 

pressure regulators, if any. 

34 Describe problems with None noted during visit. 

feedwater regulating valves, 

if any. 

35 What, if any, is your The plant engineers are actively involved in the AOV Users Group. Dave 

involvement with the AOV Stricker is the present Chairman and former Vice-Chairman of the 

Users Group? Describe. AOVUG. Sonja Waters is the current Vice-Chairman and former 

I Engineering and Design Committee Chairman of the AOVUG.

C-10
Appendix C

TRIP No. 2 REPORT 

STUDY OF AIR-OPERATED VALVES 

FERMI 2, NOVEMBER 3 AND 4,1997 

We had two days of meetings and interviews with the engineers at Fermi 2 who are concerned with 

AOVs. We also were shown portions of the air system and valves served by it. 

The AOVs at Fermi 2 are under the cognizance of the Plant Support Engineering (PSE) group. 

This group is supported by engineering services provided by Tenera and Sargent & Lundy. Detroit 

Edison is also a member of a group of utilities that pool their resources to devise integrated solutions to 

problems and concerns. This group of utilities is know as the Utilities Service Alliance, USA, made up of 

Fermi 2, WNP 2, Ft. Calhoun, Clinton, Cooper, Palisades, and Wolf Creek. AOVs are being studied at 

Fermi as one of the USA's group efforts. 

The engineers provided us with a notebook, prepared for our visit, that described their program and 

some of the pertinent problems and studies regarding AOVs. These are being reviewed as part of this 

AOV study. The contents of the notebook included a description of their AOV program, and several root 

cause analyses of failures of AOVs to perform as intended. 

The cooperation, courtesy, and knowledgeable responses from the members of the Fermi 2 staff 

were noted and appreciated by those of us who are involved in this study of AOVs. 

The compressed air systems at Fermi 2 are described in Section 9.3.1 of the Fermi 2 FSAR. The 

station air system and the control air system provide clean, dry, oil-free compressed air for plant 

operation. The control air system consists of an interruptible air system (IAS) and a non-interruptible air 

system (NIAS). The control air system provides air with a dewpoint of -40OF (at pressure). The station 

air system and IAS are non-safety-related, non-seismic, and designed to ASME Code Section VIII and 

ANSI B3 1.1.0. The NIAS is safety-related, seismic category 1, and designed to ASME Section 3, Class 3 

requirements. The NIAS is required for safe-shutdown and for control during long-term recovery. 

The station air and IAS equipment is located in the turbine building and the NIAS equipment is 

located in the auxiliary building. The station air system consists of three compressors, two 150 ft3 

capacity receivers, and an associated distribution network (piping, valves, and fittings). The station air 

compressors and their associated coolers are cooled by the turbine building closed cooling water system. 

A station air connection supplies the LAS. The IAS includes two 100% redundant dryers, each with its 

own pre-filter, after-filter, and instrumentation, IAS receiver, and an associated distribution network. The 

NIAS consists of two, 100% capacity 100 scfm compressors, two parallel trains of oil filters, air dryers, 

and after-filters, two parallel air receivers, and an associated distribution network. The control air 

compressors and after-coolers are cooled by the reactor building closed cooling water system or the 

emergency equipment cooling water system. 

The source of non-interruptible (NIAS) and interruptible (IAS) control air during normal plant 

operation is from the station air system through interconnections between the station air system and 

control air system. One interconnection supplies Division 1 and 2 of the NIAS and another 

interconnection supplies the IAS. The air is filtered and dried. The FSAR indicated that the air for the 

NIAS is filtered to remove particles greater than 0.5 microns nominal and 0.9 microns absolute, but 

filtering requirements for the air supplied by the IAS or station air was not included. There are no alarms 

for high dew point. 

The NIAS supplies control air to the:

C-11I NUREG/CR-6654
Appendix C 

- standby gas treatment system, 

- control center air conditioning system, 

- main steam isolation valve leakage control system, 

- primary containment atmosphere monitoring system, 

- emergency equipment cooling water system, 

- primary containment pneumatic supply system, 

- torus to reactor building vacuum relief system, and 

- railroad bay airlock door seals. 

In addition, Division 1 of the NIAS supplies control air to the: 

- primary containment isolation of drywell equipment and floor drain sump pump discharge lines, 

and 

- back-up supply for Division 1 (nitrogen) pneumatic supply to the primary containment. 

In addition, Division 2 of the NIAS supplies AOVs in the following systems: 

- high pressure coolant injection system, 

- reactor core isolation cooling system, 

- standby gas treatment primary containment isolation valves which support torus venting, and 

- torus vent secondary containment isolation valves. 

All other control air users are connected to the IAS. 

Any one of three station air compressors will be operating during normal operation. One of the 

other two station air compressors is on "auto" and the other is "off." Normal operating pressure is 100 

psig. If the station air header pressure drops below 95 psig, the compressor on "auto" is to automatically 

start and if the pressure drops to 90 psig, an alarm is initiated in the control room. The compressor in 

"off' can be manually started form the control room. If the station air header pressure decreases to 85 

psig the station air header isolates, air is supplied only to the IAS and NIAS, and an alarm is initiated in 

the control room. If station air supply pressure to either division of the NIAS decrease to 85 psig, its 

division's control air compressor automatically starts. If the pressure continues to decrease, the station air 

supply isolates from the NIAS and alarms at 75 psig, and each division of NIAS is supplied by its own 

control air compressor from then on.

NUREG/CR-6654 C-12
Appendix C

There is a normally locked-closed intertie between the NIAS divisions and each division may be 

supplied from the other. There is also a normally closed IAS inter-system tie to Division 2 of the NIAS 

which may be opened (for maintenance, etc.). In this case, a loss of offsite power or loss of header 

pressure would render NIAS Division 2 inoperable, so an intertie auto isolation valve is installed to 

automatically close in order to maintain Division 2 NIAS receiver integrity. 

The primary containment is normally supplied with nitrogen from a separate nitrogen inerting 

system (FSAR section 9.3.6). An intertie is provided to permit Division lof the NIAS to be used as an 

emergency backup to Division 1 of the containment pneumatic supply system. Bottled nitrogen can also 

be used. 

The control air compressors are automatically started upon loss of off-site power, with power 

supplied by the emergency diesel generators (EDGs). Each division of the NIAS has enough receiver 

capacity to supply control air for 10 minutes after loss of offsite power and before the EDGs come on 

line. Control air accumulators are also individually provided for safety-related equipment. 

Another separate system is dedicated to provide starting air to the EDGs. 

The AOVs at Fermi 2 are categorized as 1, 2 or 3. Category 1 includes "AOVs having relatively 

high safety significance." Category 2 includes "less safety significant AOVs that support safety related 

functions or have relatively high economic consequences." Category 3 includes "AOVs having limited 

safety or economic consequences." These definitions were taken from Figure 1, page 6, under tab 3 in the 

notebook of handout materials provided by the licensee. The definition for Category 3 AOVs in Figure 1, 

page 6, did not exactly match the definition found on page 3, tab 3. Further, the scope depicted in Figure 

1 of those AOVs covered by the Maintenance Rule, 10 CFR 50.65, appeared to be incomplete. These 

observations were passed on to the licensee's engineers for their evaluation. 

There are 29 Category 1 AOVs and 34 Category 2 AOVs out of approximately 1100 (our own 

rough estimate) AOVs in the plant. These valves are equipped with actuators from 37 different 

manufacturers. In addition, there are approximately 2482 solenoid operated valves in the plant, of which 

1442 are classified as class QA-1. By way of comparison, there are 147 motor-operated valves in the 

Fermi 2 MOV program. 

The air systems are not equipped with devices to monitor dew point and air quality is monitored at 

"approximately" six month intervals. 

Fermi 2 is heavily involved in compliance with the Maintenance Rule, 10 CFR 50.65, at this time. 

They are reviewing AOVs within the context of the Maintenance Rule to determine important (risk 

significant) AOVs. They are conducting design basis reviews to determine if the AOVs were designed 

with sufficient margin. They are going to follow up with static diagnostic testing to verify the condition 

of the AOVs. They do not appear to be particularly concerned about air system quality and believe it to 

be acceptable. However, there had been previous air system quality problems (moisture) which resulted 

in corrosion products in the air system and thus affected plant operation. Fermi 2 made instrument air 

system modifications, including the installation of new dryers, and this appears to have improved the 

situation. 

Fermi 2 is the lead BWR plant in an Electric Power Research Institute (EPRI) program to develop 

an overall Air-Operated Valve Program document. Detroit Edison is to conduct design basis system level 

(d/P, flow, temperature) and component level (required thrust or torque and actuator output capability or 

margin) evaluations for their Category 1 AOVs using the methods defined in the AOV Program 

document. This effort is a follow up to the EPRI Performance Prediction Program (PPM) that was

C-13 NUREG/CR-6654
Appendix C

devised for motor-operated valves and is planned to be used for AOVs, to the extent practical. EPRI and 

its contractors are working with the Detroit Edison staff to categorize and analyze the capabilities of 

selected, important AOVs. 

We noted from the statement of work for the EPRI project that the experience gained during the 

project is planned to be used in the preparation of a comprehensive AOV Evaluation Guide to be 

developed by EPRI (but which is beyond the scope of the project). 

The AOV engineers pointed out a particular AOV (PF45F400A or B, Division 2 Pump Discharge 

PCV, EESW) during a tour of the plant. They indicated that their recent risk analyses, performed in 

response to the Maintenance Rule, 10 CFR 50.65, revealed for the first time that these AOVs were 

particularly risk significant. Subsequent investigations by the licensee confirmed this conclusion. These 

valves receive an opening signal at pump start and it is important that the valves re-close to prevent 

diversion of flow. This observation was considered to be germane to the current study of AOVs and 

indicated the significance of verifying the design basis. 

Fermi does not have a dedicated diagnostic system for AOVs at this time. They are working with 

the Utilities Service Alliance to pool their resources and devise a common program. Their plans for 

obtaining a diagnostic system for AOVs did not appear to be firm at the time of our visit. 

The following tabular summary was prepared to describe the information gathered during the 

subject visit. The NRC, with assistance from the INEEL, is studying the performance of air-operated 

valves (AOVs) in commercial nuclear power plants. The information was collected in accordance with 

the Program Plan dated 10/22/97 (INEEL Letter to H. Ornstein, NRC, from J. Bryce, 10/23/97, Job Code 

E8238, Task Order 15 - JHB-167-97).

NUREG/CR-6654 C-14
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE ENRICO FERMI 2 NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

1 Date November 3 and 4, 1997. 

2 Name of Interviewer. Owen Rothberg, INEEL/LMITCO, 301/816-7773 

John Watkins, INEEL/LMITCO, 208/526-0567 

Hal Ornstein, NRC/AEOD, 301/415-7574 

Joe Colaccino, NRC/NRR, 301/415-2753 

3 Plant Name & Docket No. Enrico Fermi 2 Nuclear Power Plant, Docket No. 50-341.

4

_______ £

Person(s) Interviewed, 

Title(s), Phone Number(s), 

E-Mail address, short 

description of 

organization(s) and duties. 

NOTE: The telephone area 

code for Fermi was recently 

changed to 734.

Alan T. Goldsby, Valve Engineer, Detroit Edison, 313/586-1777 

Dan Thomas, DECO Maintenance Engineer, 313/586-5598 

J. O'Donnell, DECO Maintenance, 313/586-5209 

A. (Inadi) Nayaknadi, DECO Plant Engineer, 313/586-1195 

John Wald, DECO/ISIIPEP, 313/586-1619 

Joe Pendergast, DECO Licensing Engineer, 313/586-1682 

Linda Boguci, DECO Risk Assessment Engineer, 313/586-1317 

Jorge Ramirez, PSA Consultant, 313/586-1466 

John Tibai, Maintenance Rule Principal Engineer, 313/586-4289 

Edward J. Vinsilo, I&C Maintenance Supervisor for AOVs, 313/ 

586-4936 

Jaime L. Perez, Maintenance-Training, 313/586-4341 

Randy Kendrick, Maintenance Engineering, 313/586-5384 

Roger Tasell, DECO Plant Support Engineer, 313/586-1768 

Greg Lane, DECO Maintenance (SOVs), 313/586-1952 

John Tansek, DECO Chemistry, 313/586-5388 

Mark E. Soave, PSE-EQ, 313/586-1362 

Donald Cobb, OPS 

Steven Booker, Electrical Maintenance 

Kenneth Howard, Maintenance Support Engineer 

Ron Mathews, Maintenance 

Gabe Verespej, Electrical Maintenance 

David Roe, Electrical Maintenance 

John A. Hughes, NQA

C-15 NUREG/CR-6654

I
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE ENRICO FERMI 2 NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

Rodney Johnson, Licensing 

Neil O'Keefe, NRC Resident Inspector, 313/586-2798 

George Pickard, Air System Engineer (contacted 6/2/98) 

5 If necessary, and if person(s) A notebook was provided to us when we arrived, the contents of which 

interviewed can do so, are: 

obtain any missing 1 AOV Program Document 

information not provided 

prior to the site visit, as 2. Failure Data 

described in the outline for 3. PRA Application and Data Use 

Task 4 above. Note what 

information was provided. 4. Maintenance Data and Diagnostic Systems in Use 

5. Design Data 

6. Overall Number of AOVs 

7. Events Involving AOVs 

8. AOV Failures 

9. Actions Taken After AOV Events 

10. Root Cause Analyses 

11. Maintenance and IST Procedures for AOVs 

12. Diagnostic Systems Used For AOVs 

13. Design (and Analysis) Procedures for AOVs 

14. Training for Installation, Maintenance and Testing of AOVs. 

A memorandum titled "Failure Evaluation of ASCO Solenoid Valves for 

Fermi 2...", dated September 17, 1997, was provided. This memorandum 

refers to DER 97-1202, "Inadvertent Closure of T4901F466," a copy of 

which we were also given. 

6 Describe plant events Failure of a number of solenoid valves, as described in Fermi 2 DER 97

involving AOVs and 1202, involved use of thread-locking compounds which subsequently 

provide reference migrated to the working parts of the valves and caused them to stick in 

information, if possible. various positions. This was a recurring and significant problem that had 

Recent: occurred at several other plants and is a common-cause failure mechanism. 

Recurring: Spring preload settings on four AOVs were found deficient, as 

documented in LER 34194004. This may be a recurring deficiency and is 

Significant: a significant common-cause failure mechanism, depending on the valve 

function. 

7 Describe AOV or air system The air system has been, and continues to be, upgraded to minimize 

actual or detected potential moisture intrusion. A number of Fermi 2 Deviation Event Reports (DERs 

failures at the plant? 96-0730, 93-0045, and 88-1696) indicated that there were air quality 

Provide reference problems that led to moisture in the air lines and subsequent commonNUREG/CR-6654 C-16
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

QTT1 VTOUT TO THF. FNRTCO FERMI 2 NUCLEAR PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

information, if possible. cause failures of AOVs. 

Fermi 2 engineers do not monitor air quality for particulate or moisture 

contamination on a continuous basis. They do not have automatic 

moisture monitoring devices in the air system and, therefore, must rely on 

examination of drain samples. Moisture content for the interruptible 

portion of the air system is monitored at monthly intervals. The non

interruptible portion of the air system is monitored for moisture content at 

quarterly intervals. 

8 Describe actions taken after The use of thread locking compound was severely restricted by the 

events or failures involving licensee, training of maintenance personnel was initiated, and the potential 

AOVs or the air system. use of similar materials that might cause problems was investigated. Lab 

Provide reference analysis of adjacent valves was to be performed to estimate the extent of 

information, if possible. migration. 

The spring preload problem was resolved by re-engineering. 

9 Were there any actual or Solenoid valve problems involving use of thread locking compound 

potential common mode or constitute a common-cause failure occurrence which has been a concern in 

common cause failures in the industry for some time. This type of failure mechanism was 

the air system or AOVs at documented in NUREG-1275, Vol. 2, 12/87, and in Vol. 6, 2/91. See item 

the plant? Describe and 6, 8, and DER 97-1202. 

provide reference Several DERs (for example, 96-0730, 93-0045, and 88-1696) indicated 

information, if possible. that the air system had been contaminated with moisture and several AOV 

problems resulted. 

LER 34194004 indicated that several AOVs might not be capable of 

meeting their design basis demands because of inadequate control of the 

valve actuator settings, which resulted in insufficient preload settings in 

the actuator settings. 

10 Describe root cause analysis Root-cause analyses are noted in LERs. An example of a detailed root

procedures for the plant. cause analysis for the failure evaluation of solenoid valves was included 

Provide reference under Tab 9 of the notebook which was provided to us by the Fermi 2 

information, if possible. engineers. DER 97-1202 refers. 

11 Describe root-cause Root-cause analyses are noted in the materials provided and in several 

analyses performed for air LERs. See item 10. 

system or AOV failures at 

the plant. Provide reference 

information, if possible. 

12 Describe maintenance Maintenance was discussed with the plant representatives. Preventive 

procedures for the air maintenance consists of changing filters and monitoring air quality on 

system. Provide reference approximately six month intervals. Other maintenance is in response to 

information, if possible. failures or defects.

C-17 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE ENRICO FERMI 2 NUCLEAR PL.ANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

13 Describe maintenance The plant engineers are approaching AOVs based on their experiences 

procedures for AOVs. with MOVs. They have selected 41 AOVs for their initial design basis 

Provide reference review. 

information, if possible. The engineers were in the process of selecting a diagnostic system vendor 

Safety-related: for AOVs at the time of our visit. There was no diagnostic equipment non-safety- available on site. At the time of the visit, DECO was negotiating with a 

related: contractor who was to evaluate design basis capabilities of 41 important 

(category 1) AOVs. 

Non-safety-related: 

14 Describe IST procedures for No IST is done for the air system. 

the air system. 

Provide reference 

information, if 

possible. 

15 Describe IST procedures for IST is done on safety-related AOVs to meet the requirements of 1 OCFR. 

AOVs. Provide reference No dynamic testing of AOVs has been done. 

information, if possible. Other AOVs are stroke/time tested prior to return to service after 

Safety-related: maintenance. 

Important non-safety

related: 

Non-safety-related: 

16 Describe diagnostic systems, Fermi did not have a dedicated diagnostic system for AOVs at the time of 

if any, used for AOVs. our visit. They are working with a group of other utilities (the Utilities 

Provide reference Service Alliance, USA, made up of Fermi 2, WNP 2, Ft. Calhoun, Clinton, 

information, if possible. Cooper, and Wolf Creek) to pool their resources and devise a common 

Description of system: program. Their plans for obtaining a diagnostic system for AOVs did not 

appear to be firm at the time of our visit. 

Specifications: 

Data collected and 

frequency of collection: 

Vendor assistance provided, 

if any: 

17 Describe design (and The design of AOVs is being reviewed by Tenera and Sargent & Lundy 

analysis) procedures for under contract with Detroit Edison and with support from EPRI. EPRI is 

AOVs. Describe how providing technical as well as project management input toward 

design basis is established developing sizing criteria for AOVs. Tenera reviewed the safety-related 

and maintained for AOVs. and non-safety-related AOVs and categorized them into Category I, II, and 

Provide reference III based on PRA and safety significant functions. Sargent & Lundy will 

information, if possible. be performing the design basis calculations of Category I and II valves. 

Listings of valves by category were provided. The engineering staff 

learned recently about the high risk significance of a number of AOVs and

NUREG/CR-6654 C-18
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITF VISIT TO THE ENRICO FERMI 2 NUCLEAR PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

has modified their surveillance accordingly. 

18 Describe analyses and/or No dynamic testing was being done although the Fermi 2 program plan for 

testing for verification of AOVs indicates that dynamic testing of some AOVs was being 

operability during postulated considered. Fermi will depend on a verification by S&L/Tenera and some 

transient or accident static testing. The IST and maintenance procedures are outlined under 

conditions. Provide Tab 11 of the notebook that was provided to us by the Fermi 2 engineers. 

reference information, if 

possible. 

19 Describe training for The training program was described and the training facility was visited. 

installation, maintenance, Fermi has a test loop on site and several impressive mock-ups. Refer to 

and testing of AOVs. Tab 14 of the notebook that was provided to us by the Fermi 2 engineers. 

Provide reference 

information, if possible. 

20 Describe databases used to We were shown the databases and information that the plant uses to track 

track maintenance, failures, failures and events. There are a number of initiatives that the plant is 

and events regarding AOVs. following regarding design verification of AOVs. 

Provide reference 

information if possible. 

On site: 

Company wide: 

Industry: 

21 Describe the impact of the Many of the recent activities regarding review of AOV design and 

Maintenance Rule, 10 CFR maintenance are being motivated by compliance with the Maintenance 

50.65 on AOV and air Rule. Fermi 2 is using risk assessments as tools to determine those valves 

system maintenance and and systems that need to be tracked under the Maintenance Rule. Their 

testing. Provide reference design verification initiatives appear to be directly related to the 

information if possible. Maintenance Rule efforts. They appear to be enthusiastic about the 

application of these tools and believe that their efforts will improve 

performance and save money as well. 

The AOV engineers pointed out a particular AOV (PF45F400A or B, 

Division 2 Pump Discharge PCV, EESW) during a tour of the plant. They 

indicated that their recent risk analyses, performed in response to the 

Maintenance Rule, 10 CFR 50.65, revealed for the first time that these 

AOVs were particularly risk significant. Subsequent investigations by the 

licensee confirmed this conclusion. These valves receive an opening 

signal at pump start and it is important that the valves re-close to prevent 

diversion of flow. This observation was considered to be germane to the 

current study of AOVs and indicated the significance of verifying the 

design basis. 

22 Is PRA data used for Does not appear to be done yet. We were told that they plan to do so. 

predictive maintenance or Study of predictive maintenance program(s) is ongoing. 

replacement of AOVs? If

C-19 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE ENRICO FERMI 2 NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

so, how? 

23 Are AOVs serviced on site, AOVs are generally serviced on site. Solenoids are replaced as piece parts 

serviced offsite, or replaced and are shipped offsite for analysis, when appropriate. 

as piece-parts if found to 

require service? 

24 Identify and describe the Fermi has had a common-cause failure problem with certain solenoid 

most common recurring valves, related to use of thread locking compounds that migrate toward the 

maintenance problem(s) and working parts and foul them. Refer to DER 97-1202 (Tab 9 in the 

failures regarding AOVs and notebook provided to us by the Fermi 2 engineers). 

the air system. What did Previous problems were related to moisture in the air system, prior to 

you see? Provide reference recent and ongoing plant modifications. Several DERs (for example, 96

information if possible. 0730, 93-0045, and 88-1696) indicated that the air system was 

contaminated with moisture and several AOV problems resulted. 

25 Interviewer comments We viewed a diesel generator service water isolation valve, the diesel 

regarding actual valves generators and their directly attached valves, the training facility for 

viewed during the visit, in AOVs, a flow loop, and several mock-ups including a full-size MSIV. 

the plant, undergoing The AOV engineers pointed out a particular AOV (PF45F400A or B, 

maintenance or replacement, Division 2 Pump Discharge PCV, EESW) during a tour of the plant. They 

or in the plant stock system, indicated that their recent risk analyses, performed in response to the 

if applicable to this Maintenance Rule, 10 CFR 50.65, revealed for the first time that this AOV 

interview. 

was particularly risk significant because it may be called upon to open 

under accident loads. This AOV was to be evaluated as one of the 

Category 1 valves in the EPRI sponsored AOV program. This observation 

was considered to be germane to the current study of AOVs and indicated 

the significance of verifying the design basis. 

Several AOVs (diaphragm and spring types) were noted to be mounted 

with their actuators horizontal or at an angle with the vertical. We were 

assured that this was acceptable, although perhaps unusual. 

26 Has the plant made changes Solenoid valves were changed to different models and the addition of live

to valves or systems that load packing was made on some valves. 

include AOVs, or replaced Changes were prompted by recurring solenoid failures and packing 

AOVs with different models leakge. Rerr pro blemsrwith cont id air spd talves 

of AOVs or different valves leakage. Recurring problems with contaminated air supplied to valves 

that are not AOVs? If so, required changes to the air system and monitoring techniques. 

describe the changes and the The NRC, in Generic Letter 88-14, provided guidance to all licensees 

circumstances. regarding the necessity to verify the quality of air provided by the air 

What prompted the change? systems so that safety-related equipment would perform as expected. Fermi 2 upgraded their air system in response to GL 88-14. 

Was the change made for 

this plant only? 

27 Does the plant follow Yes. The plant follows the recommended actions that are applicable to its 

EPRINMAC guidelines for AOVs.

NUREG/CR-6654 C-20
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

•ITT1 V'TRTT Tfl TI--W 1UNRT[CO FF.RMI 2 NUJCLEAR PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

maintaining AOVs and the 

air system(s)? If not, 

describe differences and 

reasons for the differences. 

Provide reference 

information, if possible. 

28 What is the plant doing or Fermi had not made a special response to the industry correspondence as 

planning to do in response to of the time of the visit. 

the recent Industry 

correspondence on AOVs. 

Provide reference 

information, if possible. 

29 Do you have any The AOV engineers are aware of the importance of insights from their 

suggestions for improving Maintenance Rule ranking investigations and the need for confirming 

the performance of AOVs, AOV setups (post-maintenance testing). 

particularly in the areas of 

surveillance, testing, or 

maintenance? 

30 Provide a list of 10 CFR A number of inspection reports, DERs, LERs, and responses to NRC 

50.59 and 10 CFR 50.72 Bulletins or other NRC correspondence were provided. 

reports on AOVs and AOV 

support systems (air or inert 

gas supply, etc.) that have 

been issued for this plant. 

31 What thrust or torque This information was not well documented during our visit, but following 

margins are expected for the design basis verification currently in progress, they would like to have 

AOVs? Are different at least 25% margin. We were told that some AOVs (no details) are 

margins used for safety- suspected to have somewhat less, but still acceptable margins, and these 

related, important non- valves were being investigated. Currently, they adjust the packing load 

safety-related, or non-safety- based on the packing nut torque. No specific testing to determine packing 

t related AOVs? drag is performed at the present time; only a stroke/time test. 

32 What maintenance or No special inspection or surveillance is done. 

surveillance is done to AOV 

accumulators to ensure 

air/nitrogen quality and 

pressure? Were seismic 

considerations and size 

verified? 

33 Describe problems with DERs 88-1696 and 96-0730 specifically refer to contamination of the air 

pressure regulators, if any. supply. DER 95-0663 refers to repetitive failures of a pressure regulator. 

Pressure regulator problems are to be expected if clean, dry air is not 

r _consistently provided by the air system.

C-21 NUREG/CR-6654
Appendix C

NUREG/CR-6654

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE ENRICO FERMI 2 NUCLEAR PLANT 

ITEM No. INFORMATION I RESPONSE OR INFORMATION 

34 Describe problems with None were discussed during the visit. 

feedwater regulating valves, 

if any. 

35 What, if any, is your Randy Kendrick of Fermi is a member on the AOV Users Group (AUG) 

involvement with the AOV and is active and aware of their work. 

Users Group? Describe.

C-22
Appendix C 

TRIP No. 3 REPORT 

STUDY OF AIR-OPERATED VALVES 

PALISADES, NOVEMBER 18 AND 19,1997 

We had two days of meetings with the engineers and technicians at Palisades concerned with the 

air system and air-operated valves. We did not have an opportunity to view equipment in the plant; 

however, we did tour the diagnostic facilities that were used to evaluate power operated valves at 

Palisades. 

Palisades has a dedicated AOV Program Engineer under the Manager of System Engineering. At 

the time of the visit, Palisades was a member of a group of utilities that pool their resources to devise 

integrated solutions to problems and concerns. This group of utilities is known as the Utilities Service 

Alliance, USA, and was made up of Fermi 2, WNP2, Palisades, Ft. Calhoun, Clinton, Cooper, and Wolf 

Creek. The Palisades AOV Program was part of this group effort. Palisades is still participating with this 

group regarding their AOV program, but to a lesser extent than previously since the completion of their 

design basis reviews. 

The engineers at Palisades provided us with a notebook prepared for our visit that described their 

program and some of the pertinent problems regarding AOVs. The contents of the notebook also 

included a description of the Palisades organization, a document entitled "Air Operated Valve Program" 

(that specifically excludes HVAC dampers from its scope), and several lists of AOVs, as well as lists of 

AOV failures and deficiencies that have occurred. We were provided with several summary charts to 

indicate the pertinent Probabilistic Safety Assessment (PSA) results for Palisades. We were provided 

with a copy of an engineering analysis entitled "System Level Design Basis Review for Air Operated 

Valves (AOV) in the Engineered Safeguards System (ESS)." Copies of the hand-written logs associated 

with the 1978 and 1981 incidents involving failure of AOV CV-3025 in the shutdown cooling mode were 

also provided. 

We were provided with a summary of the results and goals associated with the Maintenance Rule 

review. The primary goal is to implement an action plan for the improvement of the high pressure air 

system. A closeout memo for the previously implemented plan for improvement of instrument air 

compressors, summarizing what had been accomplished, was also provided. In addition, we were 

provided with a copy of another action plan for improvement of the plant air system that was not 

associated with actions required by the Maintenance Rule. The reason for the distinction was not clear. 

In early December, the plant engineers forwarded a document entitled "Compressed Air System 

Safety System Design Confirmation Report" (SSDC), dated November 21, 1997. This report was 

prepared by the plant for the purpose of an internal review and was mentioned during our visit. 

The cooperation, courtesy, and knowledgeable responses from the members of the Palisades staff 

were noted and appreciated by those of us who are involved in this study of AOVs. 

Service and instrument air is provided by three compressors, each with a separate receiver. The 

receivers are connected to the compressed air header, which branches to an instrument air header and a 

service air header. The instrument air header is equipped with a single desiccant dryer and, pre-filters, 

and post.filters. There are two additional compressors that can be connected to the instrument air system. 

High pressure compressed air is provided by three high pressure compressors, each with its own 

refrigeration type dryer and air receiver.

C-23 NUREG/CR-6654
Appendix C

Nitrogen is supplied from bottles or in bulk as a backup. Two banks of 2000 psig nitrogen bottles 

provide limited backup for the auxiliary feedwater system valves. Four other nitrogen backup stations, 

each consisting of 2000 psig nitrogen bottles, are located in the auxiliary and turbine buildings to provide 

for operation of certain safety-related AOVs. A bulk nitrogen backup system of the instrument air system 

provides for the operation of the atmospheric dump valves (ADVs). 

The instrument air system and the high pressure air system have had significant design and 

operational weaknesses, which could have led to safety significant events (such as the losses of shutdown 

cooling in 1978 and 1981, due to failure of the single CV-3025 AOV) and common-cause failures of 

AOVs and other pneumatic equipment. Those weaknesses included: 

- Lack of redundant dryers for the instrument air system. The dryer was being bypassed when 

serviced and the system was left without drying capability during that time. 

- Use of refrigerant dryers on the high pressure air system. These dryers lack the capability to lower 

the dew point sufficiently to ensure that a supply of sufficiently dry air is provided. 

- Misplacement of filters. Several filters were noted to have been placed downstream of the pressure 

regulators that they are intended to serve. 

- Deterioration of piping and equipment served by the high pressure air system. Contaminated air 

regulators and corroded piping were reported. 

A bank of air bottles provides 1800 psig air to backup the high pressure air system for the operation 

of AOV CV-3018, to meet the fire protection requirements of Appendix R of 10 CFR. 

The Condensate Demineralizer Building compressed air needs are supplied by either of two air 

compressors, each with an integral intercooler and separate aftercooler and receiver. Service air is piped directly from the receivers, while instrument air is routed from the receivers to a dryer and then to the 

instruments. 

The operation and control of the air and nitrogen systems is described in the FSAR. 

The AOVs at Palisades are classified as Category 1, 2, or 3. Category 1 valves include safety

related AOVs with an active safety function, AOVs that are important to safety based on their PSA risk 

significance, or AOVs designated by an Expert Panel. Category 2 valves may be safety-related and of 

low risk significance or non-safety related and used in critical applications that could affect plant 

availability, capacity factor, heat rate, or maintenance costs. The remaining AOVs are included in 

Category 3. There are 111 Category 1 AOVs, and 42 Category 2 AOVs. There are approximately 714 

AOVs, total, in the Palisades plant. For comparison, there are 54 motor-operated valves (MOVs) in the 

plant, 30 of which are covered by NRC Generic Letter 89-10. 

Palisades is now involved in efforts in response to the requirements of the Maintenance Rule, 10 

CFR 50.65, and those efforts complement their AOV program. Specifically the Palisades staff is 

reviewing the importance of their AOVs from several different perspectives, including PRA and Expert 

Panel insights. Palisades has a formal AOV maintenance program and is in the early stages of 

implementing a plan for improvement of the performance of AOVs similar to that previously invoked and 

accomplished for motor-operated valves (MOVs), and based on experience with MOVs. Palisades is also 

part of an EPRI pilot program on AOVs similar to the one described to us in more detail at Fermi 2, and 

using EPRI's Performance Prediction Program devised for motor-operated valves. Palisades and EPRI 

are collaborating to develop design basis AOV calculations.

NUREG/CR-6654 C-24
Appendix C 

The goal of the Palisades AOV Program is to ensure that the program valves are capable of 

performing their design basis functions. Both static and dynamic testing have been performed to compare 

actual valve performance to assumptions made in the AOV calculations. 

Implementation of the Maintenance Rule made the plant engineers consider each AOV and rank 

the valves in terms of risk significance, in accordance with industry guidelines. The ranking process 

resulted in about 11 of 84 active AOVs in the PSA model being categorized as "high safety significance." 

Those 11 AOVs are listed in item 21 in the table that follows. 

After reviewing their program and the air system (including recent incidents), we made several 

observations about the air quality and the failures that they had experienced. Our comments on the 

Palisades AOVs and air system are summarized in the table that follows. 

The following tabular summary was prepared to describe the information gathered during the 

subject visit. The NRC, with assistance from the INEEL, is studying the performance of air-operated 

valves (AOVs) in commercial nuclear power plants. The information was collected in accordance with 

the Program Plan dated 10/22/97 (INEEL Letter to H. Ornstein, NRC, from J. Bryce, 10/23/97, Job Code 

E8238, Task Order 15 - JHB-167-97).

C-25 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE PALISADES NUCLEAR PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

1 Date. November 18 and 19, 1997. 

2 Name of Interviewer. Owen Rothberg, INEEL/LMITCO, 301/816-7773 

John Watkins, INEEL/LMITCO, 208/526-0567 

Hal Ornstein, NRC/AEOD, 301/415-7574 

Gerry Weidenhamer, NRC/RES, 301/415-6015 

3 Plant Name & Docket No. Palisades Nuclear Plant, 50-255.

Person(s) Interviewed, 

Title(s), Phone Number(s), 

E-Mail address, short 

description of 

organization(s) and duties.

Plant address is Palisades Nuclear Plant, 27780 Blue Star Memorial 

Highway, Covert, Michigan (MI), 49043. (See the phone listing in the 

materials provided for phones not shown). 

Philip D. Flenner, Senior Licensing Engineer, 616/764-2544 

Gary W. Foster, Component Engineer, 616/674-2684 

Robert A. Gambrill III, Component Engineering Supervisor, 

CE, Engineering Programs, 616/674-2497 

Thomas E. Bordine, Licensing Manager, CE 

Daniel Mauck, Crane-Movats, Component Engineer 

Chet Cynoski, CYNCOM, Consulting Engineer 

Leslie Bradshaw, Valve Engineer, CE 

Ronald Penna, Alpine Enterprises, AOV Technical Specialist 

Judy K. Ford, Manager of Engineering Programs, Engineering Programs 

Dept., CE, 616/764-2340 

Kerry A. Toner, Licensing Supervisor, CE 

Ken T. Speicher, Systems Engineer CAS, CE 

Robert A. White, Reliability Engineering Supervisor, CE, 616/ 

764-2860 

Paul F. Prescot, NRC Resident Inspector, 616/764-2741 

Eric Grindahl, Diesel Generator Engineer, CE 

R. A. Fenech, Sr. V.P. Generation, CE 

T. J. Palmisano, Site V.P., Palisades, CE 

Bill Beach, NRC, Region III 

Melvin Leach, NRC, Region III 

Gregory B. Szczotka, Manager NAPD, CE

NUREG/CR-6654

4

C-26
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE PALISADES NUCLEAR PLANT 

ITEM No. INFORMATION I RESPONSE OR INFORMATION

'I.

___________ L

If necessary, and if person(s) 

interviewed can do so, 

obtain any missing 

information not provided 

prior to the site visit, as 

described in the outline for 

Task 4 above. Note what 

information was provided.

I I

Describe plant events 

involving AOVs and 

provide reference 

information, if possible. 

Recent: 

Recurring: 

Significant:

Kurt Haas, Director of Engineering (Acting), CE 

Rob McCaleb, NPAD, CE 

Phillip Young, Project Engineer, DE&S 

Ken Squibbs, System Engineer Supervisor, CE 

Paul Fitton, System Engineer Manager, CE 

R. A. Vincent, Licensing Supervisor, CE 

Daniel G. Malone, Operations Supt., CE

We were given the following materials when we arrived and during the 

course of the discussions: 

1. A notebook containing organization charts, the Air-Operated 

Valve Program Plan, an analysis of the AOV Program scope, a list 

of Maintenance Rule (10 CFR 50.65) evaluations, and several 

valve lists. 

2. A design basis review of AOVs in the Engineered Safeguards 

System. 

3. Action plans for air systems improvements and instrument air 

compressor improvements.

LERs 25578003 and 25581030 were concerned with the significant 1978 

and 1981 failure-to-open of a particular shutdown cooling system AOV 

(CV-3025) when activated at shutdown. The valve is also required to 

open to mitigate the effects of a small-break LOCA. The single active 

failure of the valve on each occasion resulted in a rise in core temperature 

and apparent boiling or near-boiling conditions in the core. We do not 

know the exact amount of time that it would take to uncover the core, but 

it is believed to be only a few hours. Although the valve had not failed 

when called to open at shutdown during the last 16 years, we became 

concerned that the quality of the air provided by the instrument air system 

is such that the event could occur again. We expressed our concerns at the 

exit meeting. 

Another significant and recent event of interest was the common-cause 

contamination of nine high-pressure air regulators caused by rust in the air 

lines. This situation was originally reported to AEOD in April 1997 but 

was not covered by an LER. (Palisades Condition Report C-PAL-97

0404, dated 3/18/97 and INPO OE 8335, dated 4/22/97 refer to this event. 

Note that the authors did not have access to the INPO OE report.) The 

conclusion was that the high pressure air system, which serves many 

ECCS components, has not been operating in accordance with industry 

standards. As a result, the pneumatic equipment (including AOVs) 

serving the ECCS is highly susceptible to common-cause failures.

NUREG/CR-6654

5

6

C-27/"
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE PALISADES NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

7 Describe AOV or air-system In discussions with the plant staff in charge of the air system, it became 

actual or detected potential clear that the quality of the air delivered by the instrument air and high 

failures at the plant? pressure air systems at Palisades was suspect. In addition to the problems 

Provide reference noted in item 6 above, we were told that the instrument air system has 

information, if possible. only one dryer and it is bypassed when required by plant operations. 

According to Palisades, this is less than one day per year. We were also 

informed that the high-pressure air system's refrigerant dryers do not work 

properly or reliably, resulting in frequent instances of air system 

contamination (water and/or rust). In addition, several filters in the high 

pressure air system had been installed downstream of pressure regulators 

served by the air system rather than upstream. As a result, corrosion 

products have produced common-cause pressure regulator contamination. 

8 Describe actions taken after The plant has focused on the high-pressure air system as part of its 

events or failures involving Maintenance Rule reviews. The plant conducted a review of their air 

AOVs or the air system. systems and have added several goals under the Maintenance Rule, some 

Provide reference of which are still in the process of being implemented. The high-pressure 

information, if possible. air system was on the Maintenance Rule "a(l)" list due to concerns about 

long-term performance. Our additional comments prompted plant 

management to assure us of renewed and additional attention to the quality 

of air. 

9 Were there any actual or Yes. See item 6, above. According to the comments on the draft of this 

potential common mode or report from Palisades, actions to prevent reoccurrence have been taken. 

common cause failures in 

the air system or AOVs at 

the plant? Describe and 

provide reference 

information, if possible. 

10 Describe root cause analysis Not described. 

procedures for the plant. 

Provide reference 

information, if possible. 

11 Describe root-cause Not described. 

analyses performed for air 

system or AOV failures at 

the plant. Provide reference 

information, if possible. 

12 Describe maintenance Maintenance and operating practices for the air system were discussed at 

procedures for the air length. The air system is maintained by an engineer and station 

system. Provide reference maintenance personnel dedicated to this system. The station AOV 

information, if possible. program is managed by several valve engineers and technicians. Dampers 

for ventilation and diesel generator AOVs are maintained by others. See 

items 6 and 7, above.

NUREG/CR-6654 C-28
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE PALISADES NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

Air quality with respect to water and particulate content is monitored 

"periodically." 

13 Describe maintenance Palisades has a formal AOV maintenance program and is in the early 

procedures for AOVs. stages of implementing a plan for improvement of the performance of 

Provide reference AOVs, similar to that previously invoked and accomplished for motor

information, if possible. operated valves (MOVs) and based on experience with MOVs. (See the 

material in the notebook provided by the plant engineers.) Diagnostic 

Safety-related: testing equipment specifically adapted to AOVs is being evaluated. The 

Important non-safety- prioritization of AOV importance is being established from studies 

related: implemented to meet the requirements of the Maintenance Rule. It 

Non-safety-related: appears that the AOVs are getting the deserved attention. 

14 Describe IST procedures for IST is performed on the required ASME Class 1, 2, and 3 components, 

the air system. Provide which does not include much of the instrument air and high-pressure air 

reference information, if systems. Post-maintenance testing is done on components and portions of 

possible. the system affected by repairs or maintenance. 

15 Describe IST procedures for Periodic testing of AOVs is done on Category 1 AOVs and consists of 

AOVs. Provide reference ASME stroke/time testing (no load). 

information, if possible. Diagnostic testing and bench testing are done on AOVs, although 

Safety-related: apparently not periodically. 

Important non-safety- According to comments provided by Palisades on the draft, special 

related: dynamic testing of AOVs is performed as needed to verify that AOVs can 

Non-safety-related: perform their design function. No details of this testing were obtained 

during the site visit. 

16 Describe diagnostic systems, Palisades uses a MOVATS universal diagnostic system for AOVs and is 

if any, used for AOVs. trying to adapt Liberty Technologies Easy Torque/Thrust (ETT) sensor for 

Provide reference AOVs. They have become quite adept at diagnostic testing because of 

information, if possible. their MOV experience and believe that they can determine margins using 

Description of system: such tools. 

The new AOV diagnostic system is in the early stages of evolution and 

Specifications: Palisades is evaluating the software. See Palisades AOV Program, 

Data collected and frequency Procedure EM-28-03, which was included in the notebook provided to us, 

of collection: for further discussion. 

Vendor assistance provided, 

if any: 

17 Describe design (and We were provided with a copy of a document entitled "System Level 

analysis) procedures for Design Basis Review for Air Operated Valves (AOV) in the Engineered 

AOVs. Describe how Safeguards System (ESS)." Similar documents for other systems were 

design basis is established observed during the visit. Palisades is reviewing their design bases for 

and maintained for AOVs. AOVs to ensure that they are accurate and complete. As part of that 

Provide reference process, they are reviewing or revising the original calculations for the 

information, if possible. valves. In some cases plant calculations are being generated for the first

C-29 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE PALISADES NUCLEAR PLANT 

ITEM No. INFORMATION I RESPONSE OR INFORMATION 

time. Palisades is conducting system and component level design basis 

reviews of their AOVs, as described in their Program Plan. 

18 Describe analyses and/or Analyses are described in item 17, above. No specific test program is in 

testing for verification of place for verifying design-basis operability of AOVs, other than diagnostic 

operability during postulated testing and comparisons with the EPRI PPM methodology. 

transient or accident According to recent comments from Palisades, dynamic testing is 

conditions. Provide performed if the valves to not exhibit adequate margin. Special test 

reference information, if procedures are prepared for the dynamic tests to allow Operations to 

possible. manipulate the plant to maximize differential pressure and flow. Details 

and results of such tests were not obtained during the site visit. 

19 Describe training for Training for valve disassembly, diagnostic testing, and valve maintenance 

installation, maintenance, is provided. We viewed part of the Palisades facility for diagnostic testing 

and testing of AOVs. and the procedure was demonstrated. The engineers were quite 

Provide reference knowledgeable about the various diagnostic systems available. 

information, if possible. Palisades personnel made presentations regarding diagnostic testing at the 

AUG meetings, to the benefit of other plants. 

20 Describe databases used to Palisades maintains a number of computerized, on-line databases to track 

track maintenance, failures, failures and events. These can be sorted for particular valves and valve 

and events regarding AOVs. types. Engineers and technicians are familiar with the operation and 

Provide reference performance of equipment in the plant on a detailed level. 

information if possible. 

On site: 

Company wide: 

Industry: 

21 Describe the impact of the Implementation of the Maintenance Rule made the plant engineers 

Maintenance Rule, 10 CFR consider each AOV and rank the valves in terms of risk significance, in 

50.65 on AOV and air accordance with industry guidelines. The ranking process resulted in 

system maintenance and about 11 of 75 active AOVs in the PSA model being categorized as having 

testing. Provide reference "high safety significance." Those 11 AOVs are: 

information if possible. - CV-3006, LPSI Shutdown Cooling Exchanger Bypass 

- CV-2010, Condensate Inlet Containment Isolation 

- CV-0522B, Normal Steam to P-8B from Steam Generator "A" 

- CV-3025, Shutdown Cooling to LPSI Isolation 

- CV-3029, Containment Sump Isolation to East Engineered 

Safeguards Room 

- CV-3030, Containment Sump Isolation to East Engineered 

Safeguards Room 

- CV-0779, Steam Generator E-50B Steam Dump Control

NUREG/CR-6654 C-30
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE PALISADES NUCLEAR PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

- CV-0780, Steam Generator E-50B Steam Dump Control 

CV-0781, Steam Generator E-50B Steam Dump Control 

CV-0782, Steam Generator E-50B Steam Dump Control 

CV-3055, Shutdown Cooling Heat Exchanger Bypass 

22 Is PRA data used for According to recent comments from Palisades, PRA is receiving limited 

predictive maintenance or use as an input to determine predictive maintenance activities on AOVs. 

replacement of AOVs? If Corrective maintenance appeared to be the norm. Selected AOVs do 

so, how? have maintenance procedures that discuss overhauls and predictive 

maintenance checks. 

23 Are AOVs serviced on site, No standard procedure is used. Maintenance, repair, or replacement 

serviced offsite, or replaced methods depend on the circumstances. 

as piece-parts if found to 

require service? 

24 Identify and describe the There is only one dryer for the instrument air system. As a result, when 

most common recurring the dryer must be taken off line for any reason the compressor after-cooler 

maintenance problem(s) and is relied upon for air drying. A backup dryer is available but, because of 

failures regarding AOVs and the expense, had yet to be installed. Air quality was monitored 

the air system. What did intermittently. 

you see? Provide reference The high-pressure air system serves both safety-related ECCS equipment 

information if possible. and non-safety-related equipment. Refrigerant dryers for the non-safety

related high pressure air system tend to freeze up and so it is necessary to 

cycle the compressors to prevent that. An adjustment was made to the 

drain valve timer to ensure that moisture does not remain in the air dryer 

exchanger. 

Recently, several regulators were found contaminated with corrosion 

products which were caused by moisture in the air system. In addition, 

several filters are installed downstream of the air regulators in the high 

pressure air system rather than upstream of them. 

One valve CV-3025 (high risk significance AOV, see item 21), used for 

shutdown cooling, failed in 1978 and again in 1981 in single-failure 

incidents that led to boiling, or near-boiling, conditions in the reactor 

during shutdown. Although the valve had been modified (provided with a 

hand wheel), and has not failed since 1981, the quality of the air in the air 

system is suspect, and therefore, so is the potential performance of this 

valve. This AOV is now stroked quarterly. Previously, CV-3025 was 

stroked on a cold shutdown frequency. See item 6 in this table for 

additional information concerning the history of this AOV.

C-31 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE PALISADES NUCLEAR PLANT

ITEM No. INFORMATION [ RESPONSE OR INFORMATION 

24 Identify and describe the The Compressed Air System SSDC provides detailed information on the 

most common recurring licensee's assessment of the Palisades air systems. It was noted that the 

maintenance problem(s) and CV-3025 valve was not mentioned in Attachment B of that report, which 

failures regarding AOVs and describes air system functional requirements. 

the air system. Several SOV related events have occurred. One event of interest, 

documented in LER 25592007, occurred on February 5, 1992 while the 

plant was operating at 100% power. As a result of an ongoing Equipment 

Classification (Q-list) review program, it was determined that the main 

steam isolation valve (MSIV) actuator solenoid valves could be rendered 

inoperable by a main steam line break outside of containment. There were 

several contributing factors related to the cause of the MSIVs solenoids 

not meeting the EQ rule (10 CFR 50.49) requirements regarding electrical 

isolation: 

- The redundant set of solenoid valves were installed in a 

non-harsh environment to ensure that the main steam isolation 

valves would still close in the event of a main steam line break 

outside of containment. This modification, however, used the 

same power source as the original SOVs without ensuring 

appropriate isolation of non-qualified equipment on the power 

circuit. This resulted in the second set of solenoid valves not 

being completely redundant. 

Identify and describe the In 1981 the plant environmental qualification (EQ) project 

most common recurring evaluators believed the FSAR to be correct. They failed to realize 

maintenance problem(s) and that the FSAR was incorrect and that the non-harsh environment 

failures regarding AOVs and solenoid valves were truly not redundant. Based on the erroneous 

the air system. information, the MSIV solenoids were removed from the EQ list. 

Another event that described inadequate environmental qualification of 

SOVs which are piece-parts of AOVs is documented in LER 25592016. 

SOVs and position switches for the control valves which control the 

service water flow from the CAW Heat Exchangers were not 

environmentally qualified in accordance with 10 CFR 50.49. SOVs 

SV-0823A, SV-0823B, SV-0826A, SV-0826B, and position switches 

POS-0823 and POS-0826 provide control and indication for control valves 

CV-0823 and CV-0826 which control the service water flow from the 

CAW Water Heat Exchangers E-54A and E-54B, respectively. These 

components were not qualified for a high-energy line break outside 

containment and were not on the EQ list. Furthermore, they were not 

electrically isolated from environmentally qualified instruments in the 

same electrical scheme. 

The root cause for this event was attributed an inadequate engineering 

analysis. 

25 Interviewer comments We did not have time to view specific AOVs in the plant itself; however, 

regarding actual valves we saw several AOVs in a test/training facility and viewed diagnostic 

viewed during the visit, in testing devices that the plant uses.

NUREG/CR-6654 C-32
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE PALISADES NUCLEAR PLANT

ITEM No. 1 INFORMATION RESPONSE OR INFORMATION 

the plant, undergoing 

maintenance or replacement, 

or in the plant stock system, 

if applicable to this 

interview. 

26 Has the plant made changes In the specific case of CV-3025, a handwheel was added as an emergency 

to valves or systems that operation measure. 

include AOVs, or replaced Based on poor performance of the original valves, and the guidance in 

AOVs with different models NUREG 1275, Vol. 6, several solenoid valves throughout the plant were 

of AOVs or different valves replaced with different models. 

that are not AOVs? If so, According to comments received from Palisades on the draft of this report, describe the changes and the 

a number of changes were made to the air system and procedures related to 

circumstances. the air system. Refer to Attachment 4 to the letter from N. Haskell of 

What prompted the change? Consumers Energy to the NRC dated July 30, 1999, for a list of those 

Was the change made for changes. 

this plant only? 

27 Does the plant follow Industry guidance, including EPRINMAC guidelines, are consulted when 

EPRI/NMAC guidelines for formulating plant procedures. Explicit compliance with industry guidance 

maintaining AOVs and the could not be determined. Palisades is part of an EPRI pilot program on 

air system(s)? If not, AOVs similar to the one described to us in more detail at Fermi 2, and is 

describe differences and using EPRI's Performance Prediction Program devised for motor-operated 

reasons for the differences. valves. Palisades and EPRI are involved in a collaborative effort to 

Provide reference develop design basis AOV calculations. 

information, if possible. 

28 What is the plant doing or See item 27. 

planning to do in response to 

the recent Industry 

correspondence on AOVs. 

Provide reference 

information, if possible. 

29 Do you have any Plant engineers discussed several updates to the air system that had been 

suggestions for improving proposed to management to improve air quality and the subsequent 

the performance of AOVs, performance of air-operated equipment. Among those were installation of 

particularly in the areas of a replacement dryer and relocation of filters in the high pressure air 

surveillance, testing, or system. Recent information received from Palisades in response to the 

maintenance? draft of this report indicated that a replacement dryer is to be installed and 

filters are to be relocated in the near future. See item 24. 

30 Provide a list of 10 CFR None provided beyond the LERs previously found by AEOD/INEEL. 

50.59 and 10 CFR 50.72 

reports on AOVs and AOV 

support systems (air or inert 

gas supply, etc.) that have 

been issued for this plant.

C-33 NUREG/CR-6654
Appendix C

NUREG/CR-6654

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE PALISADES NUCLEAR PLANT 

ITEM No. [ INFORMATION I RESPONSE OR INFORMATION 

31 What thrust or torque A number of "20% "was discussed as an acceptable margin. Details on 

margins are expected for existing margins for AOVs were not provided. 

AOVs? Are different Note: This question was asked to get an idea of what engineers 

marginssed, fortsafetyn - considered to be an acceptable margin. There was no attempt to establish 

related, important non

safety-related, or non-safety- any sort of commitment to a particular value. 

related AOVs? 

32 What maintenance or Surveillance on containment isolation valve accumulators are performed 

surveillance is done to AOV every outage. Accumulator leakage is monitored and trended to ensure 

accumulators to ensure that the accumulator can maintain the valve closed for up to 4 hours. 

air/nitrogen quality and Seismic design is in accordance with the design basis and FSAR 

pressure? Were seismic commitments. 

considerations and size 

verified? 

33 Describe problems with A significant and recent event of interest was the common-cause 

pressure regulators, if any. contamination of nine high-pressure air regulators caused by rust in the air 

lines. This situation was originally reported to AEOD in April 1997 but 

was not covered by an LER. 

34 Describe problems with Not discussed. 

feedwater regulating valves, 

if any. 

35 What, if any, is your Gary Foster and Bob Gambrill are active participants in the AOV Users 

involvement with the AOV Group. 

Users Group? Describe. Palisades personnel made presentations regarding diagnostic testing at 

AUG meetings.

C-34
Appendix C

TRIP No. 4 REPORT 

STUDY OF AIR-OPERATED VALVES 

LaSALLE, DECEMBER 17 AND 18,1997 

We had two days of meetings with the engineers and technicians at LaSalle concerned with the air 

system and air-operated valves. We were escorted through the rooms in the plant that housed the station 

air system equipment. The plant was shut down, and been shut down for an extended period, at the time 

of our visit. We did not have an opportunity to visit the training facilities or see diagnostic equipment 

used by the plant. 

LaSalle has several engineers and technicians dedicated to the station air system and another group 

dedicated to the diesel air start system. A third group is concerned with maintenance and service of 

ventilation dampers. LaSalle also has an Equipment Qualification Department (EQ), an Inservice Testing 

Department (IST), and a group dedicated to implementation of the Maintenance Rule. The interfaces and 

divisions of responsibilities between these groups was not clear to us. 

The engineers at LaSalle provided us with a draft "Administrative Procedure for Air Operated 

Valve Program," a draft "Engineering Qualification Guide Air Operated Valve (AOV) Engineer 

Qualification," and several lists of AOVs in groupings by system. In addition, we were provided with a 

copy of a training document entitled "LaSalle A.O.V. Seminar" that describes procedures for setting and 

maintaining AOVs. An internal document, "Subject: Elastomer Evaluation Guide - NDIT MSD-94

048," was provided as part of the discussion of elastomeric materials during the visit. Results of database 

searches in the LaSalle internal maintenance (Problem Identification Form, e.g., PIF) database for "AOV" 

and "air-operate" were also made and provided to us during the visit. 

The AOV program plan is in the draft and planning stage at LaSalle and they want to have it in 

place by restart. (They were projecting a 4/98 restart date at the time of our visit.) No ranking of AOVs 

at LaSalle, in terms of importance, was presented. 

There are several compressed gas systems used to power AOVs at LaSalle; these are station air 

(consisting of service air and instrument air, which are shared by the two units) and the drywell pneumatic 

systems, one for each unit. The instrument air system is used to supply air to operate AOVs outside 

containment and the drywell pneumatic systems supply nitrogen to operate AOVs inside each drywell. 

The instrument air system supplies both units. The licensee noted that the EDGs are not dependent on the 

operation of any AOVs. The EDGs are dependent on stand-alone SOVs (one per EDG). 

The instrument air system is equipped with continuous dew point monitors at the receivers and 

alarms in the control room. This was the first plant that we visited that had such equipment. The LaSalle 

personnel noted that early in the plant's life, they had numerous problems because they had not paid much 

attention to the dew point monitors. However, when actions were taken to fix the causes of high dew 

point, they experienced a dramatic improvement, and in recent years have had very few problems that can 

be attributed to poor air system quality. The drywell pneumatic system and the diesel starting air system 

do not have dew point monitors or alarms. 

No compressors or their associated equipment are required to safely shut down either unit 

following a postulated LOCA and/or loss of offsite electrical power. AOVs requiring pneumatic supply 

for safe reactor shutdown are provided with individual pneumatic accumulators. 

There are a total of 90 safety-related AOVs for both units at LaSalle. In addition, each unit has 370 

safety-related control-rod drive valves (740 total for both units). There are 1575 non-safety-related AOVs

C-35 NUREG/CR-6654
Appendix C 

at the LaSalle plant (both units). For comparison, there are a total of 200 motor-operated valves in the 

Generic Letter 89-10 program for both LaSalle units. 

We obtained a copy of the LaSalle Summary PRA dated March 1996. According to the Summary 

PRA, "(t)ransients with loss of instrument air, (T1 1), are the largest initiating event category, contributing 

32% of the CDF. These transients are significant because venting containment cannot be performed 

without instrument air. Failure to vent results in the loss of the ADS function (and subsequent loss of the 

low pressure injection systems) and eventual containment failure, causing potential loss of injection 

systems in the reactor building due to severe environments." 

"Loss of offsite power (LOSP) events are the second highest contributor to CDF. Single unit 

LOSP events contribute 6.5% of CDF and dual unit LOSP events contribute 22.9%. If AC power can be 

restored to the emergency busses by the diesel generators or crossties, then the plant response is similar to 

transient events. If both diesel generators or the crossties become unavailable, the unit is considered to be 

in a station blackout sequence. The core damage contribution of those SBO sequences (subset of LOSP) 

is 17.2%." 

It is interesting to note that a loss of offsite power to Unit 1 or to both units leads directly to a loss 

of instrument air and the CDF resulting from such an event should be at least as significant as a loss of 

instrument air alone. The "Accident Sequence Event Descriptions" listed in the tables in the Summary 

PRA include the event described as "LOSS OF INSTRUMENT AIR 1E OR LOSP AT UNIT 1;" 

however, the Summary PRA does not provide a discussion of the relationship or dependencies between 

the two events. 

A summary card, distributed by Commenwealth Edison is reproduced below and shows the 

following information regarding the PRA for LaSalle: 

FRONT OF CARD [BACK OF CARD 

Initiating Events 

Initiator %CDF 

Loss of Instrument Air 32.1 LaSalle Station 

Dual Unit LOSP 22.9 Key PRA Results 

Transient with Bypass 8.0 

Transient w/out Bypass 7.4 

Single Unit LOSP 6.5 

Loss of Division 1 AC 4.8 

Loss of Division 2 AC 4.7 

MSIV Closure 4.3 

Other 9.3 

Key Operator Actions Key Equipment 

System RAW 

Initiate ADS DC Division 1Y 31.7 

Vent Containment DC Division IX 25.1 

Depressurize with bypass valves only 0DG01P 24.5 

FW cntrl to lower power < bypass DG"B"CWP 22.0

NUREG/CR-6654 C-36
Appendix C

FRONT OF CARD f BACK OF CARD 

Capacity (ATWS) 1(2)DGO1P 20.2 

ADS to Restart pumps given FWLC-A DC Division 2Y 20.2 

fails (ATWS) HPCS 18.1 

Restore RPV level to mix boron Instrument Air 16.9 

Initiate SBLC RHR "B" 16.2 

RHR "A" 15.7 

Containment Vent 7.5 

SBLC 7.4 

CD 5.4 

RCIC 5.1 

DG "B" 4.1 

DG "A" 3.5 

LPCS 3.0 

DG "0" 2.5 

After the visit we received NRC Event No. 33434 dated 12/19/97, "Inadequate Turbine Building 

Vent System Exhaust ..." This report was discussed with plant representatives by phone on 1/7/98 (H. 

Ornstein & 0. Rothberg / Shafique R. Khan & Vince Guterrez). The Event Report involves a transient 

analysis performed to predict pressures in the turbine building ventilation exhaust tunnel downstream of a 

postulated high energy line break. A particular set of ventilation system air-operated dampers, whose 

safety function is to remain closed, could remain open under certain circumstances involving failure of 

non-safety-related solenoids that control the dampers. In this circumstance, the generated pressures could 

exceed the structural capability of block walls in the area and this could result in damage to vital 

equipment in the area. During a recent walk-down they found a failed SOV which could be a precursor. 

LaSalle intends to make hardware modifications so that failure of the solenoids cannot compromise the 

safety function of the dampers. This recently discovered postulated event describes a condition where 

failures of non-safety-related AOVs can affect or interfere with equipment that has to perform a safety 

function under certain conditions. We are concerned about the importance of a number of events that 

involve air-operated damper (AOV) failures in nuclear plants and about how the dampers are designed 

and maintained. 

The cooperation, courtesy, and knowledgeable responses from the members of the LaSalle 

engineering and technical staff were noted and appreciated by those of us who are involved in this study 

of AOVs. 

The following tabular summary was prepared to describe the information gathered during the 

subject visit. The NRC, with assistance from the INEEL, is studying the performance of air-operated 

valves (AOVs) in commercial nuclear power plants. The information was collected in accordance with 

the Program Plan dated lC/22/97 (INEEL Letter to H. Ornstein, NRC, from J. Bryce, 10/23/97, Job Code 

E8238, Task Order 15 - JHB-167-97).

C-37 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE LaSALLE NUCLEAR PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

1 Date. December 17 and 18, 1997. 

2 Name of Interviewer. Owen Rothberg, INEEL/LMITCO, 301/816-7773 

Mark Holbrook, INEEL/LMITCO, 208/526-4362 

Hal Ornstein, NRC/AEOD, 301/415-7574 

Joe Colaccino, NRC/NRR, 301/415-2753 

3 Plant Name & Docket No. LaSalle County Station, Docket No.'s 50-373 and 50-374.

4

____________ I

I. 1

Person(s) Interviewed, 

Title(s), Phone Number(s), 

E-Mail address, short 

description of 

organization(s) and duties.

Steve Shields, Com. Edison/AIRC Coordinator 

Guy Campbell, Com. Edison/EPM 

Baron Westphal, Com. Edison/Pump and Valve Supervisor, x2770 

John Kowalski, Corn. Edison/AOV Mech. Engr. 

Tim Sandness, Com. Edison/FLS 

Ernie Bianchetta, Com. Edison/Principal Instructor 

Paul Templet, Com. Edison/FLS 

Mark A. Smith, Com. EdisonIAOV Coordinator, x2323 

Shafique R. Khan, Sargent & Lundy/Sr. Project Engineer, 312/ 

269-7482 

Ivo Garza, Com. Edison/Power Operated Valves 

Steven Smalley, Com. Edison/PCM, Maintenance Rule 

Rodney Delap, Com. Edison/SA/IA Engineer 

Keith Tabel, Com. Edison/SA/IA Group Leader 

Roy Linthicum, Sargent & Lundy/Project Engineer 

Al Carroll, Com. Edison/Coordinator 

Robert Tjernlund, Com. Edison/Design Engineer, x2918 

Robert Cockrel, Com. Edison/DG System Engineer 

Fred Darim, Com. Edison/Plant General Manager 

Michael Reynolds, Com. Edison/IMA 

Len North, Com. Edison/ImCST 

John Pollock, Com. Edison/Support Engineering Supervisor 

Bob Janacek, Com. Edison/PRA (by phone) 

LaSalle Switchboard: 815/357-6761 

LaSalle Mailing Address is: 260lNorth 21st Rd., Marsailles, IL 61341

NUREG/CR-6654 C-38
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE LaSALLE NUCLEAR PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

5 If necessary, and if person(s) Several LERs and event descriptions that appeared pertinent were selected 

interviewed can do so, for discussion with the LaSalle engineering staff prior to the trip. Several 

obtain any missing of these are discussed under item 6, below. 

information not provided The engineers at LaSalle provided us with a draft "Administrative 

prior to the site visit, as Procedure for Air Operated Valve Program," a draft "Engineering described Task in the outline forI 4 above. Note what Qualification Guide Air Operated Valve (AOV) Engineer Qualification," 

Tnformask 4b on ove s o wand ed several lists of AOVs in groupings by system. In addition we were 

information was provided, provided with a copy of a training document entitled "LaSalle A.O.V. 

Seminar" that describes procedures for setting and maintaining AOVs. An 

internal document, "Subject: Elastomer Evaluation Guide - NDIT MSD

94-048," was provided as part of the discussion of elastomeric materials 

during the visit. Results of database searches in the LaSalle internal 

maintenance (Problem Identification Form, e.g., PIF) database for "AOV" 

and "air-operate" were also made and provided to us during the visit. 

Copies of P&IDs for the instrument air and drywell pneumatic system 

were also provided to us. 

The FSAR Section 9.3.1 on Process Auxiliaries was reviewed after the site 

visit. 

6 Describe plant events LER # 37385008, "Secondary Containment Isolation Dampers Found 

involving AOVs and Inoperable" 

provide reference LER # 37385011, "Reactor Scram Caused by High Temperature in the 

information, if possible. Main Steam Tunnel Caused by Solenoid Valve Repairs" 

Recent: LER # 37387032, "Reactor Scram on Low Reactor Level Due To 

Recurring: Difficulty ... Feedwater Regulating Valve" 

Significant: LER # 37389007, "Potential Loss of Control Room Isolation Due to 

Possible Failure of Exhaust Purge Dampers..." 

LER # 37391007, "RWCU Isolation Due to Leaky Filter/Demineralizer 

Valve" 

LER # 37396011, "Pneumatic Valves With Less-Than-Designed 

Diaphragm Area Results in Inadequate Valve Closing Forces..." 

LER # 37486015, "RWCU ... Isolation Due to Relief Valve Lifting" 

LER # 37488010, "Shutdown Due to ADS Nitrogen Backup Pressure 

Regulator Failure" 

LER # 37489018, "Battery Low Temperatures Caused by Failed Damper 

and Low Outside Air Temperatures" 

LER # 37490011, "Loss of Unit 2 North Bank ADS Backup Pressure 

Supply" 

LER # 37492016, "Reactor Scram on Loss of Air..." 

LER # 37493005, "2A Diesel Generator Air Start System Below 200 

psig..."

C-39 39NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

ZTTt VNTT TOTT- THE LaSALAT.F, NUCLEAR PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

LER # 37495005, "Failure of Outboard MSIV..." 

NRC Inspection Report No. 50-373/97011 dated October 15, 1997, 

discusses leaks in Scram Service Pilot Valves. This report was discovered 

after the trip and was not discussed with plant representatives. 

NRC Event No. 33434 dated 12/19/97, "Inadequate Turbine Building 

Vent System Exhaust ...," was discovered after the trip and was discussed 

with plant representatives by phone on 1/7/98 (H. Ornstein & 0. Rothberg 

/ Shafique R. Khan & Vince Guterrez). LaSalle provided a copy of an 

Engineering Design Change Package to NRC after the phone call (received 

on 1/23/98). 

7 Describe AOV or air system LER # 37396011 refers to a situation discovered by the LaSalle engineers 

actual or detected potential in which certain AOVs did not have effective diaphragm areas (EDA) that 

failures at the plant? conformed to those the valve manufacturer publishes. A "Part 21" 

Provide reference notification was issued by LaSalle on 10/4/96. A total of 36 AOVs are 

information, if possible. -involved at LaSalle. This is a generic problem applicable to similar or 

other Anchor Darling/WKM valves at other plants. As learned from other 

AOV Users Group participants, the problem has also been observed with 

other AOV manufacturers. 

Several LERs (37385008, 37385011, 37389007, 37489018) and NRC 

Event Report 33434 refer to various problems with air-operated dampers. 

LER 37492016 describes a SCRAM caused by loss of the instrument air 

system. Two station air compressors tripped on high lube oil temperature, 

which was caused by loss of cooling water, which was caused by 

personnel closing the TBCCW supply and return valves for one unit which 

were cross-tied to another unit. 

LER 37493005 describes a diesel generator failure caused by moisture in 

the air lines. This system is not equipped with moisture sensors or alarms. 

8 Describe actions taken after Refer to the LERs in item 7. 

events or failures involving 

AOVs or the air system. 

Provide reference 

information, if possible. 

9 Were there any actual or Common cause failures were indicated in LER 37396011 that refers to 

potential common mode or incorrect calculation of effective diaphragm areas on 36 AOVs. 

common cause failures in Loss of instrument air or either of the drywell pneumatic systems each has 

the air system or AOVs at a number of common cause ramifications. Refer to the LERs described 

the plant? Describe and above. 

provide reference 

information, if possible. 

10 Describe root cause analysis Root-cause analysis procedures were not described. However, extensive 

procedures for the plant. follow-up to AOV anomalies were performed, including use of diagnostic

NUREG/CR-6654 C-40
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY •~TmE 17TcTmT 'T'C'. TTJr' T ... AT I E' N.T TC'T "I AD DT A ITT

S. Ih E 111 1 ~ 1r I HE L~aoS L.LE IN U LJEL,-tX 1- LA~1 I 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

Provide reference equipment and associated analyses tools. The discovery of the problem 

information, if possible. of the EDA of certain AOVs, described in item 9 of this table, is an 

example. 

11 Describe root-cause Root causes are described in most (since 1986) of the LERs noted above. 

analyses performed for air It appears that the root causes were logically determined and addressed. 

system or AOV failures at 

the plant. Provide reference 

information, if possible. 

12 Describe maintenance The instrument air receiver moisture traps are blown down once per shift 

procedures for the air except during the summer when it is done twice per shift. Evidence of 

system. Provide reference contamination is sought. Filters are changed on a regular schedule and 

information, if possible. there is a regular schedule for dryer maintenance. 

The licensee indicated that AOVs, and pneumatic equipment in general, 

were performing much better as a result of their heightened awareness and 

improved IA system maintenance. There were numerous air system 

contamination events in the 1980s that were attributed to inadequate dryer 

maintenance. 

LaSalle was the first plant visited that continuously monitors air quality 

and our preliminary conclusion is that the reliability of AOVs and other 

pneumatic equipment is improved (because of better quality air) over 

some of the other plants visited. 

13 Describe maintenance We did not visit the training or maintenance facilities at LaSalle. 

procedures for AOVs. LaSalle has several engineering and maintenance groups. They are 

Provide reference 

information, if possible. knowledgeable concerning the equipment and have expertise in 

maintaining it; however, their approach to maintenance appears to be 

Safety-related: reactive. This will probably change because of the impact of the 

Important non-safety- Maintenance Rule, 10 CFR 50.65. LaSalle is in the preliminary stages of 

related: implementation of the Maintenance Rule. 

Non-safety-related: 

14 Describe IST procedures for No IST is done on the air system except for post-maintenance and post

the air system. Provide repair testing. 

reference information, if 

possible. 

15 Describe IST procedures for Section XI of the ASME Code (1989 edition with no addenda) is used to 

AOVs. Provide reference stroke/time test safety-related AOVs. No IST is performed on other 

information, if possible. AOVs. 

Safety-related: 

Important non-safety

related: 

Non-safety-related:

C-41 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE LaSALLE NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

16 Describe diagnostic systems, LaSalle uses a recent version of Fisher FlowScanner for diagnostic testing 

if any, used for AOVs. of AOVs. In addition, the plant engineers have adapted some of the 

Provide reference techniques developed for MOVs to predict the margins on AOVs. Using 

information, if possible. these combined techniques, the engineers discovered the discrepancies 

described in LER # 37396011 and NRC Information Notice 96-68. For 

Description of system: example, certain AOVs did not have effective diaphragm areas (EDA) that 

Specifications: conformed to those the valve manufacturer publishes. The pneumatic 

Data collected and diaphragm actuators are WKM Model 70-13 in sizes 35, 70, 140, and 280. 

LaSalle has 36 of these actuators total in both units. Thirteen of the 

valves, per unit (26 total) are in the RCIC system. 

Vendor assistance provided, 

if any: 

17 Describe design (and Procedures are described in the draft AOV program for a design basis 

analysis) procedures for review. A detailed description of the design and analysis procedures was 

AOVs. Describe how not provided. 

design basis is established 

and maintained for AOVs. 

Provide reference 

information, if possible. 

18 Describe analyses and/or Procedures are described in the draft AOV program for a design basis 

testing for verification of review. A detailed description of the design and analysis procedures was 

operability during postulated not provided. 

transient or accident 

conditions. Provide 

reference information, if 

possible. 

19 Describe training for Training was described in general terms. All technicians are trained to 

installation, maintenance, assemble/disassemble AOVs using hands-on techniques. Diagnostic 

and testing of AOVs. testing of AOVs is performed by a trained group. 

Provide reference 

information, if possible. LaSalle Station has been ComEd's lead plant in the field of AOV 

diagnostics and LaSalle has been "lending out" AOV technical personnel 

to help upgrade those activities at the other CornEd plants. LaSalle 

personnel made presentations at the AUG and international (ICONE) 

meetings, to the benefit of other plants. 

20 Describe databases used to The LaSalle plant uses a number of computerized databases to track 

track maintenance, failures, maintenance, events, and failures. These appear to provide necessary 

and events regarding AOVs. information. We asked during the visit, and in a telephone conversation 

Provide reference shortly thereafter, about sharing information between the CECO plants 

information if possible. and the industry in general. 

On site: 

Company wide: 

Industry:

NUREG/CR-6654 C-42
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE LaSALLE NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

21 Describe the impact of the LaSalle will, in accordance with the Maintenance Rule, examine the 

Maintenance Rule, 10 CFR performance of certain AOVs and the supporting air supply system and 

50.65 on AOV and air establish performance goals. This, in turn, requires specific actions to 

system maintenance and improve performance. At the time of our visit, the plant did not appear to 

testing. Provide reference be as far along in the process as some of the other plants that we visited. 

information if possible. 

22 Is PRA data used for PRA data is not being used to make a decision on the AOV program at 

predictive maintenance or this time. Again, it is expected that PRA and failure data will be used by 

replacement of AOVs? If the plant in accordance with the Maintenance Rule. 

so, how? 

23 Are AOVs serviced on site, AOVs are serviced on site, under the supervision of an AOV coordinator. 

serviced offsite, or replaced Solenoids are normally replaced as piece-parts and are not repaired. 

as piece-parts if found to 

require service? 

24 Identify and describe the Problems with air-operated dampers appear common, although no 

most common recurring common cause was identified. Problems with solenoids, related to organic 

maintenance problem(s) and materials and lubrication incompatibilities, as documented in NUREG

failures regarding AOVs and 1275, Vol. 6, also appeared to be common. 

the air system. What did Contamination and/or moisture in the instrument air system, in contrast to 

you see? Provide reference plants previously visited, has NOT been a problem in recent years. The 

information if possible. operators experienced air system and AOV problems when the plant first 

started up until they paid attention to instrument air quality, as was being 

reported on the dew point monitoring equipment. 

25 Interviewer comments We toured the instrument air room in the plant and were impressed by the 

regarding actual valves cleanliness, large scale, and attention to detailed design that was apparent 

viewed during the visit, in in the installation. The engineers were cognizant of the design and 

the plant, undergoing function of the air system, and its requirements. 

maintenance or replacement, 

or in the plant stock system, 

if applicable to this 

interview. 

26 Has the plant made changes Plant changed ASCO 8323 (see LER 37495005) solenoids to Valcor 

to valves or systems that solenoids on MSIVs after repeated failures related to lubricant/thread

include AOVs, or replaced locking compound-related problems. 

AOVs with different models Change out of the ASCO 8323 SOVs were done in a staggered sequence of AOVs or different valves 

as suggested in NUREG-1275, Vol. 6. 

that are not AOVs? If so, 

describe the changes and the It was noted that the Valcor valves are to be installed, per the 

circumstances. manufacturer's instructions, using Graphoil tape; however, LaSalle uses a 

different thread-locking compound. We questioned the use of thread

What prompted the change? locking compound and were told that the engineers would investigate.

C-43 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE LaSALLE NUCLEAR PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

Was the change made for 

this plant only? 

27 Does the plant follow EPRIINMAC guidelines are used for reference, along with information 

EPRIINMAC guidelines for from a number of industry sources. 

maintaining AOVs and the 

air system(s)? If not, 

describe differences and 

reasons for the differences. 

Provide reference 

information, if possible. 

28 What is the plant doing or We were provided with the draft of a document entitled "LaSalle 

planning to do in response to Administrative Procedure - Air Operated Valve Program." The purpose of 

the recent Industry the procedure, as stated therein, is to categorize AOVs based on their 

correspondence on AOVs. safety significance and functional requirements, prioritize AOVs based on 

Provide reference past performance and industry data, establish AOV design bases by 

information, if possible. reviewing design data, determine as-built configuration by performing 

walkdowns, tracking, and trending. The plan appeared to be 

comprehensive and was based on information from a number of industry 

organizations including the AOV Users Group, NSAC, NMAC, and EPRI. 

29 Do you have any The use of a comprehensive program for AOVs, that includes 

suggestions for improving prioritization based on safety significance, failure experience, diagnostic 

the performance of AOVs, evaluations, industry experience, and test data should be used to improve 

particularly in the areas of the performance of AOVs. LaSalle is in the process of constructing and 

surveillance, testing, or implementing such a program. 

maintenance? 

30 Provide a list of 10 CFR A number of LERs (noted previously in items 6 and 7) and information 

50.59 and 10 CFR 50.72 from the LaSalle PIF database were reviewed. 

reports on AOVs and AOV 

support systems (air or inert 

gas supply, etc.) that have 

been issued for this plant. 

31 What thrust or torque Engineers would like to have 20% margin in their calculations. The 20% 

margins are expected for figure is an informal judgement call and is not a plant requirement. 

AOVs? Are different The question was asked in order to survey knowledgeable engineers about 

margins used for safety- what they thought was an acceptable margin and was not done to establish 

related, important non

safety-related, or non-safety- any commitment. 

related AOVs? 

32 What maintenance or The accumulators are blown down every refueling cycle and the check 

surveillance is done to AOV valves are verified to be functioning at that time. Seismic design, along 

accumulators to ensure with other design considerations, are to be reviewed as part of a 

air/nitrogen quality and comprehensive program. 

pressure? Were seismic I

NUREG/CR-6654 C-44
Appendix C

NUREG/CR-6654

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE LaSALLE NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

considerations and size 

verified? 

33 Describe problems with LER # 37490011, "Loss of Unit 2 North Bank ADS Backup Pressure 

pressure regulators, if any. Supply" indicated the failure or misoperation of a pressure regulator used 

to maintain several ADS accumulators. The North ADS Bottle Bank 

pressure regulator was not opening to control header pressure. A manual 

stop valve was found to have been left in the closed (wrong) position. The 

drywell pneumatic compressors failed to maintain system pressure during 

the event. The stop valve mispositioning affected the three ADS valves 

supplied by the North Bottle Bank. There are a total of seven ADS valves. 

34 Describe problems with LER # 37387032, "Reactor Scram on Low Reactor Level Due To 

feedwater regulating valves, Difficulty Controlling Reactor Water Level with the Feedwater Regulating 

if any. Valve at Low Flow, Low Power Conditions" refers to a problem with the 

system design. Hardware modifications were described in the LER 

including installation of a motor operator on the FRV inlet isolation valve 

along with a control switch in the control room, as well as installation of a 

bypass line and regulating valve around the existing FRV. It is assumed 

that these modifications were made; however, we did not verify this 

during our visit. 

35 What, if any, is your The plant has a engineer who is a representative heavily involved in the 

involvement with the AOV AOV Users group. LaSalle AOV engineers and their contractors have 

Users Group? Describe. made presentations at AOV Users Group and other industry meetings to 

feed back information on their AOV diagnostics and the EDA problems 

that were discovered with their Anchor Darling/WKMI/BSB valves.

C-45
Appendix C

Trip No. 5 REPORT 

STUDY OF AIR-OPERATED VALVES 

THREE MILE ISLAND, UNIT 1, FEBRUARY 12 AND 13, 1998 

We had two days of meetings and interviews with the engineers at Three Mile Island (TMI) who 

are concerned with AOVs and the air systems. We were shown portions of the air systems, as well as 

some of the valves served by them. 

TMI has several engineers dedicated to the service air system and AOVs. Several types of air

operated components such as air-operated dampers, solenoid-operated valves in various services, and air

operated components attached to the diesel generators are served by system engineers for the particular 

systems. We noted that the engineers involved with the air systems that we spoke to seem knowledgeable 

about their own systems and equipment as well as the other air systems and equipment outside of their 

cognizance. 

We were provided with a number of documents and reports including a draft "Air Operated Valve 

Program Description;" an installation specification for "Installation of Air Filters on Critical Plant Control 

Valves;" a copy of the TMI letter to the NRC responding to Generic Letter 88-14 on instrument air 

system problems; several lists of plant events and maintenance reports; and a package of information 

(including sorted lists of valves) used to describe the methods and techniques for risk analysis and risk

based ranking of AOVs and other equipment. 

We were also provided with a copy of a report of the B&W Owners Group (B&WOG) entitled 

"Instrument Air System Review Report." This was part of an effort by the B&WOG for B&W plants that 

corresponded to the work in NUREG-1275, Volume 2, and Generic Letter 88-14. We spoke with Gordon 

Skillman, who was a contributor to the report. 

The purpose of the instrument and control air system at TMI is to continuously deliver clean, dry 

air at 100 psig (-40tF dew point, filtered to 0.9 micron particle size). The instrument and control air 

system consists of three unlubricated air compressors, each discharging through a separate after cooler 

and air receiver. Two of the air compressors utilize a common air dryer (the "old" portion of the 

instrument and control air system). The third air compressor uses a separate air dryer (the "new" portion 

of the instrument and control air system). The "new" portion of the instrument and control air system was 

added to address problems with maintaining air quality, particularly moisture, that the plant experienced 

after startup. The "new" portion is now run continuously and the "old" portion is only run for testing or 

when the "new" portion is taken off line for service. 

Backup for the instrument and control air system is provided by a connection to the plant service 

air system. Air will automatically flow from the plant service air system at a pressure of approximately 

70 psig through an oil removal filter when instrument air pressure drops sufficiently, in order to supply 

components needed to shut down the plant. A two-hour backup air supply is also available to provide 

compressed air for operation of components within the main steam and emergency feedwater systems if 

the plant instrument air system is not available. We were told that no AOVs are powered from the 200 

psig diesel air start system. 

The instrument and control air system is equipped with continuous dew point monitors and alarms 

in the control room. The instrument and control air system is arranged to be restarted and run off the 

diesel generators, from a switch in the control room, in the event of a loss of offsite power. The piping 

for the "new" instrument and control air system is stainless steel up to the dryers and copper downstream. 

The piping for the "old" instrument and control air system is carbon steel up to the dryers and copper 

downstream. All of the receivers are carbon steel with a coated interior. We were told that there is

NUREG/CR-6654 C-46
Appendix C

evidence that the coating may have worn over time. A number of air filters have been added directly 

upstream of critical plant control valves in order to minimize the risk of particulate contamination of these 

components. The performance of the instrument air system is described in Item 7 in the table below. 

The TMI engineers described to us a particularly interesting event involving insufficient margins in 

two AOVs. Calculations for five Crane-Aloyco, 2.5 to 6 inch, 1500 and 150 pound class, flex-wedge and 

split wedge gate AOVs with Miller DA-63-B and A-63-B cylinder actuators were reviewed by one of the 

TMI engineers. The AE requested that the valve manufacturer (Crane-Aloyco) perform thrust 

calculations on the five AOVs, as part of a limit switch upgrade modification, in order to verify that limit 

switch installation would not affect valve operability. The resultant thrust calculations, using "present 

day" methodology, indicated that two of the five valves had negative closing margins for the specified 

differential pressure (d/P). The architect/engineer (AE) then requested that the manufacturer redo the 

calculations using the methodology by which the valves were originally sized. The revised Crane-Aloyco 

calculations, based on the original valve sizing methodology, indicated positive margins in both the 

opening and closing directions, but using a valve factor of zero. Upon review of the Crane-Aloyco 

calculations, TMI convened a review group which verified that the two containment isolation valves 

which had negative margins in the manufacturers' first calculations were, in fact, operable and would be 

able to perform their designed safety function. This conclusion was based on TMI's calculations using a 

0.75 friction factor and d/P of 1600 psi (that d/P required for containment isolation). 

A design modification was proposed to increase the closing margin for one of the valves because 

normal operating d/P (2375 psi) was greater than the d/p needed to perform its safety function (1600 psi). 

The proposed modification included the installation of an accumulator and piping to provide air assistance 

to the spring. Subsequently, TMI performed an analysis of the valve using the EPRI PPM methodology 

(2375 psi and 0.67 friction factor) and confirmed the operability of the valve. Design problems with a 

pressure booster for the AOV were also noted. 

It was learned that Crystal River 3 (a similar B&W plant with the same AE, Gilbert Associates, as 

TMI-1) had a similar AOV margin problem. Both plants have been exchanging information on methods 

to increase the available margins for the AOVs. These conditions were not previously covered by an LER 

or other correspondence with the NRC. As a result of our visit, the TMI engineers agreed to investigate 

to determine if there are Part 21 concerns to be addressed, either by TMI or the valve manufacturer 

(Crane-Aloyco). There may be generic or common-cause issues with these types of AOVs, related to the 

design or the OEM calculations, that need to be addressed. 

Several common-cause solenoid valve (piece parts of AOVs) failures, involving hardening of 0

rings were reported. These failures are described in the Table, below. 

There are 910 AOVs in the TMI plant. Of these, 98 are safety-related (Q class or Class 1), 328 are 

Class 2, and 484 are Class 3, in accordance with the lists provided. By way of comparison, there are 81 

motor-operated valves in the TMI Generic Letter 89-10 program. 

The cooperation, courtesy, and knowledgeable responses from the members of the TMI technical 

staff were noted and appreciated by those of us who are involved in this study of AOVs. 

The following tabular summary was prepared to describe the information gathered during the 

subject visit. The NRC, with assistance from the INEEL, is studying the performance of air-operated 

valves (AOVs) in commercial nuclear power plants. The information was collected in accordance with 

the Program Plan dated 10/22/97 (INEEL Letter to H. Ornstein, NRC, from J. Bryce, 10/23/97, Job Code 

E8238, Task Order 15 - JHB-167-97).

C-47 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

,RITE VISTT TO THREE MILE ISLAND. UNIT 1. NUCLEAR GENERATING STATION

ITEM No. INFORMATION RESPONSE OR INFORMATION 

1 Dates. February 12 and 13, 1998. 

2 Names of Interviewers. Hal Ornstein, NRC/AEOD, 301/415-7574 

Owen Rothberg, INEEL/Rockville, 301/816-7773 

John Watkins, INEEL/Idaho Falls, 208/526-0567 

David Terao, NRC/NRR, 301/415-3317 

3 Plant Name & Docket No. Three Mile Island, Unit 1, Docket No. 50-289. 

4 Persons Interviewed, Titles, Harold Wilson, Supervisor, Maintenance Assessment, TMI, 717/948

Phone Numbers, 8050, FAX 717/948-8598 

E-Mail address, short David Atherholt, Plant Maintenance Director, TMI, 717/948-8838, FAX 

description of organizations, 717/948-8598, datherholt@gpu.com 

and duties. Charles Hartman, GPUN/Engineering Dept., 717/948-8150 

James Gilles, GPUN/Engineering Dept., 717/948-8840 

Howard Crawford, Manager - Programs Engineering Division, TMI, 717

948-8412, FAX 717/948-6822, hcrawford@gpu.com 

Greg Gurican, Nuclear Licensing Engineer, TMI, 717/948-8753, FAX 

717/948-2820, ggurican@gpu.com 

Patrick Bennett, GPUN System Engineer, 717/948-8232 

Charles Adams, GPUN, IOSRG (PRA), 717/948-8055 

Mark Fauber, GPUN Licensing, (Conducted Tour of Plant) 

Gordon Skillman, GPUN, (Stopped by and told us of B&W Owners 

Report on Instrument Air System Review) 

The plant mailing address is: 

GPU Nuclear Inc. 

Route 441 South 

Post Office Box 480 

Middletown, PA 17057-0480 

5 If necessary, and if person(s) No information was provided by TMI prior to our visit. A copy of our 

interviewed can do so, program plan was provided to TMI. 

obtain any missing We arrived with the following materials: 

information not provided 

prior to the site visit, as A copy of Section 9.10.1 of the FSAR (Update-12, 3/94), 

described in the outline for INSTRUMENT AND CONTROL AIR SYSTEM 

Task 4 of the program plan. Reduced GPU Drawing No. 302-271, Rev 62, INSTRUMENT & 

Note what information was STATION SERVICE AIR Flow Diagram 

provided. 

(continued next page)

NUREG/CR-6654 C-48
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THREE MILE ISLAND, UNIT 1, NUCLEAR GENERATING STATION

ITEM No. INFORMATION RESPONSE OR INFORMATION

5 (continued) - Reduced GPU Drawing No. 302-272, Rev. 16, BACKUP 

INSTRUMENT AIR Flow Diagram 

- Reduced GPU Drawing No. 302-273, Rev. 18, EMERGENCY 

FEEDWATER & MAIN STEAM VALVE 

- 2-Hour Backup Supply Air Flow Diagram 

- LER 28986002 Summary, Anticipatory Reactor Trip, Turbine Trip, 

High Moisture Separator Level 

- LER 28986007, Failure to Meet Design Criteria on the Two Hour 

Backup Air Supply 

- LER 28987008, Reactor Trip, Turbine Trip, High Moisture 

Separator Level 

- A list of LERs taken from the Sequence Code Search System 

(SCSS) database 

The following materials were provided by TMI during our visit: 

- AIR OPERATED VALVE PROGRAM DESCRIPTION Draft 

Topical Report 118 

- Letter from GPUN to NRC dated 2/24/89 (C311-89-2016) 

indicating TMI's response to Generic Letter 88-14 

- Letter from GPUN to NRC dated 2/26/92 (C311-92-203 1) 

indicating TMI's completion of actions for their Generic Letter 88

14 commitments 

- A computer list of MAINTENANCE TREND ACTION NOTICE 

(MTAN) REPORTS dated 2/4/98 with AOV items marked 

(individual MTANs noted below) 

- A computer list of POTENTIAL FAILURE OF AIR OPERATED 

VALVES, SOLENOID VALVES, AND AIR SYSTEM AT TMI 

prepared 2/13/98 

- Citations for the ORAM User's Manual, Version 1.5 (DOS), EPRI 

TR-102819, Nov. 1993 and the ORAM SENTINEL Users Manual, 

Version 3.0, TR 107018, Sept. 1997, ALL MODES 

MAINTENANCE AND SAFETY FUNCTION ADVISOR 

- Cover page and Introduction to the TMI OUTAGE FUEL 

PROTECTION CRITERIA, Topical Report 097, Revision 3, dated 

1/16/95 (Howard Crawford) 

- Exhibit 5 to TMI Administrative Procedure 10780, Rev. 13, TMI 

PSA SUMMARY RESULTS, Contains listing of core damage 

sequences and initiating events in order of importance 

(continued next page)

C-49 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THREE MILE ISLAND, UNIT 1, NUCLEAR GENERATING STATION

ITEM No. INFORMATION [ RESPONSE OR INFORMATION

5 (continued) Materials provided by TMI (continued): 

- An undated paper (one figure is dated 7/15/93), TMI-1 PRA INPUT 

TO MAINTENANCE RULE - RISK SIGNIFICANT SYSTEM 

LIST, Developed by Risk Analysis Section, C. Adams 

- A list of valves headed TABLE-i, INSERVICE TESTING 

VALVES RANKED BY FUSSEL-VESELY (WITHOUT 

COMMON CAUSE) 

- A list of valves headed TABLE-2, TMI-1 INSERVICE TESTING 

VALVES RANKED BY RISK ACHIEVEMENT (WITHOUT 

COMMON CAUSE) 

- A list of valves headed "FVRArk" indicating Fussel-Vesely 

Importance and Rank 

- A list of valves headed TABLE-3, INSERVICE TESTING 

VALVES RANKED BY FUSSEL-VESELY (WITH COMMON 

CAUSE) 

- A list of valves headed TABLE-4, TMI-1 INSERVICE TESTING 

VALVES RANKED BY RISK ACHIEVEMENT (WITH 

COMMON CAUSE) 

- A list of valves headed SHEET 6, showing a summary of the 

rankings (H i.e., high, M i.e., medium, L i.e., low, N i.e., none, T 

i.e., truncated, and NM i.e., not modeled) in Tables 1 through 4 for 

each valve 

- A list of classes of valves headed SHEET 7, summarizing total 

number of valves in the rankings (H i.e., high, M i.e., medium, L 

i.e., low, N i.e., none, T i.e., truncated, and NM i.e., not modeled) in 

Tables 1 through 4 

- MTAN No. 93-15 dated 12/21/93 to investigate the reliability of 

make-up valve MU-V-0017 as the result of a history of slow or 

sluggish operation 

- MTAN No. 93-03 dated 7/29/93 to investigate the failures of 

several solenoid failures for valves CM-V-0002, 3, 4, 7, and 8 

(includes test data) 

- TMI-1 SPECIFICATION Tl-IS-123267-001, REGULATORY 

REQUIRED INSTALLATION SPECIFICATION FOR 

INSTALLATION OF AIR FILTERS ON CRITICAL PLANT 

CONTROL VALVES 

(continued next page)

NUREG/CR-6654 C-50
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THREE MILE ISLAND, UNIT 1, NUCLEAR GENERATING STATION 

ITEM No. INFORMATION RESPONSE OR INFORMATION

5 (continued) Materials provided by TMI (continued): 

- Several diagrams and excerpted pages from TMI Report 643, 

REQUIRED THRUST ANALYSIS, for limit switch modifications 

to valves MU-V-3, MU-V-26, WDL-V-304, WDL-V-534, and 

WDL-V-535 

- INSTRUMENT AIR SYSTEM REVIEW REPORT, B&W Owners 

Group, Safety and Performance Program Report No. 47-1165965

00 dated 12/86 

Materials provided by TMI by mail after the visit: 

- A list of SAFETY EVALUATIONS REPORTS with those 

pertinent to the AOV study marked. Attached were copies of 

GPUN forms entitled "Technical Functions, Safety/Environmental 

Determination and 50.59 Review (EP-016)" and numbered with 

"SE" numbers. The attached forms were: 

- SE No. SE-CMR-1 13202-397, Replacement of instrument air 

dewpoint recorder, 1/28/93 

- SE No. SE-CMR-1 13202, Replacement of Failed IWT Regulators, 

2/5/90 

- SE No. 115101-002, Instrument Air As-Found Conditions, 5/21/91 

- SE No. 123267-001, Installation of Air Filters on Critical Plant 

Control Valves, 8/11/89 

- SE No. SE-128961-002, Two Hour Backup Instrument Air 

Switching Valve Modification, 3/16/87 

- SE No. 412512-005, Vent Valve Installation for Testing Air 

Accumulators, 1/22/90 

- SE No. 412012-002, EF and MS 2 Hour Air Supply, Truck 

Connection Modification, 9/23/83 

- The following tabular listings dated 7/19/93: 

- Table 1, Modified Master Frequency File (TMI-1 NOHUMAN 

Model) 

Materials provided by TMI by mail after the visit: 

- The following tabular listings dated 7/19/93: 

- Table 2, Initiating Event Contributions (TMI-1 NOHUMAN 

Model) 

- Table 3, TMI-1 NOHUMAN Model top 50 Output Sequence Lists 

(continued on next page)

C-51 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THREE MILE ISLAND, UNIT 1, NUCLEAR GENERATING STATION

ITEM No. INFORMATION RESPONSE OR INFORMATION 

5 (continued) - Table 4, Split Fraction Importance (Sorted by Importance) 

- Table 5, Split Fraction Importance (Sorted by Achievement) 

- Table 6, Split Fraction Importance (Sorted by Risk Reduction) 

- Table 7, Split Fraction Importance (Sorted by Fraction Designator) 

- Table 8, Top Event Importance (Sorted by Importance) 

- Table 9, Top Event Importance (Sorted by Risk Achievement) 

-_ Table 10, Top Event Importance,(Sorted by Risk Reduction Worth)

Describe plant events 

involving AOVs and 

provide reference 

information, if 

possible. 

Recent: 

Recurring: 

Significant:

- LER 28986002 described a reactor trip caused by dirt and rust in the 

air ports of a valve controller. 

LER 28986007 indicated that a backup air supply might not have 

met the single failure criterion. 

LER 28987002 described a reactor trip caused by failure of an air

operated controller due to undefined causes. 

MTAN No. 93-15, dated 12/21/93, documents the licensee's effort 

to investigate the reliability of make-up valve MU-V-00 17 as the 

result of a history of slow or sluggish operation. The valve and 

operator were rebuilt and several items, including the positioner, 

were replaced. 

MTAN No. 93-03 dated 7/29/93 documents the licensee's effort to 

investigate the failures of several solenoid failures for valves CM

V-0002, 3, 4, 7, and 8, sample supply and return valves for the 

Reactor Building air monitor (included test data). Several SOVs 

were replaced with nuclear grade valves and several were rebuilt. 

The root cause was reported as hardening of the O-rings, although 

the mechanism that caused the deterioration was not reported. This 

was a common-cause failure mechanism. 

Several diagrams and excerpted pages TMI Report 643, 

REQUIRED THRUST ANALYSIS, for limit switch modifications 

to valves MU-V-3, MU-V-26, WDL-V-304, WDL-V-534, and 

WDL-V-535. Deficient design margins of the AOVs were 

calculated during a design review (see the summary at the 

beginning of this trip report). The engineers were asked if they 

made an effort to determine if AOVs might have inadequate design 

margins or suspect calculations. They said that they did a search 

and did not find any. They also said that Crystal River 3 (similar 

plant with same AE) had a similar problem and the two plants 

discussed ways of increasing the margins of the affected AOVs. 

Several 50.59 reports (see item 5 above for a list) describe a number 

of modifications and additions to the air systems prompted by a 

desire to improve performance of the equipment served.

NUREG/CR-6654

6

C-52
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THREE MILE ISLAND, UNIT 1, NUCLEAR GENERATING STATION 

ITEM No. INFORMATION - RESPONSE OR INFORMATION 

7 Describe AOV or air system Prior to 1990, the instrument air system performed poorly from the 

actual or detected potential standpoint of providing clean, dry air. There was only one dryer, which 

failures at the plant? had to be bypassed in order to service it. In response to GL 88-14, 

Provide reference NUREG-1275, Vol. 2, and the B&WOG Instrument Air System Review 

information, if possible. Report, 47-11659565-00, TMI installed an additional instrument air 

system at that time. The "new" instrument air system, equipped with its 

own dryer, became the main supplier of instrument air for the plant and 

the "old" system is still used for backup. All of the air systems, including 

the service air system, are equipped with dryers and both of the instrument 

air systems have continuous dew point monitors and alarms in the control 

room. The "old" instrument air system has carbon steel piping upstream 

of the dryers while the "new" system has stainless steel piping. Both 

systems use copper piping downstream of the dryers. Receivers for both 

instrument air systems are carbon steel and the plant installed filters 

directly upstream of "critical" valves in order to protect against any 

particulate contamination. In addition, instrument air can be restored after 

a loss of offsite power event by operating a switch in the control room that 

powers the instrument air system from the diesel generators. A backup 

supply of air bottles provides two hours of air for the main steam and 

emergency feedwater systems. These features ensure reliable performance 

of the instrument air system, as well as the quality of the air provided. As 

a result of all the improvements in the instrument air system design, 

operation and maintenance, TMI has had few problems and no reportable 

problems related to air quality or quantity in the last few years. 

8 Describe actions taken after See item No.s 6 and 7 for a discussion of problems and actions taken. 

events or failures involving The TMI engineers were preparing an Air-Operated Valve Program at the 

AOVs or the air system. time of the visit, and furnished us with a draft copy. We do not know 

Provide reference when the plan will be finished or put into operation. This plan is similar to 

information, if possible. several that we have seen in visits to other plants and is related to the 

licensee's effort to meet the requirements of the Maintenance Rule, 10 

CFR 50.65. The scope of the program includes: 

- Categorizing AOVs by importance. 

- Determining design-bases of AOVs. 

- Identifying methods to determine required and actual AOV 

capabilities (i.e., margins). 

- Methods for making any required hardware, software, or procedural 

modifications to ensure design basis operability of AOVs. 

- Determining test methods and test accuracy for AOVs. 

- Developing training methods, standards, and devices. 

- Upgrading and improving periodic maintenance and testing of 

AOVs to ensure long-term operability. 

- Developing methods and data to evaluate AOV failures, diagnostic 

system performance, and preventative maintenance frequency.

C-53 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THREE MILE ISLAND, UNIT 1, NUCLEAR GENERATING STATION

ITEM No. INFORMATION I RESPONSE OR INFORMATION 

9 Were there any actual or Solenoid valve failures described in MTAN No. 93-03 (see ITEM No.s 5 

potential common mode or and 6) were caused by hardening of the 0-rings. The root cause for the 

common cause failures in deterioration of the 0-rings was described as aging, without further 

the air system or AOVs at details. No further problems were reported since the corrective actions 

the plant? Describe and were completed. 

provide reference 

information, if possible. 

10 Describe root cause analysis No specific or formal root cause analysis procedures were described or 

procedures for the plant. presented to us. From examination of the licensee's materials, it appears 

Provide reference that root cause analyses are routinely attempted for all failures and events. 

information, if possible. Our impression was that the methods and results have improved over the 

years. 

11 Describe root-cause One root cause analysis described to us involved the calculation of 

analyses performed for air insufficient margins in two Crane-Aloyco AOVs. Diagrams and excerpted 

system or AOV failures at pages from TMI Report 643, REQUIRED THRUST ANALYSIS, were 

the plant. Provide reference provided to us and described the investigation of margins for the AOVs. 

information, if possible. (Refer to ITEM No.s 5 and 6.) Deficient design margins for the AOVs 

were calculated during a design review (see the summary at the beginning 

of this trip report). 

As noted above, Crystal River 3 had the same or a similar problem. This 

may be a common-cause (AE and/or valve manufacturer) problem. 

12 Describe maintenance Preventive maintenance of the air system consists of filter replacement and 

procedures for the air dryer maintenance on a regular schedule. Dew point monitors are checked 

system. Provide reference on a regular schedule and we saw information being recorded on 

information, if possible. automatic recorders attached to the monitors. 

13 Describe maintenance TMI has a system for investigating trends of recurring and common-cause 

procedures for AOVs. failures. We were given several MAINTENANCE TREND ACTION 

Provide reference NOTICEs that are pertinent examples of their system and applicable to our 

information, if possible. interest. A list of MTANs was also provided. TMI considers valves in all 

categories. The plant also provides standard preventive and corrective 

Safety-related: maintenance for AOVs. The plant engineer in charge of the air system has 

Important non-safety- been involved with this system for quite some time and appears to be quite 

related: knowledgeable concerning AOV and air system maintenance problems. 

Non-safety-related: The maintenance of air-operated dampers and the AOVs for the diesel 

generators is done by the engineers and technicians in charge of the 

particular system. The TMI AOV program plan specifically excludes air

operated dampers. 

Preventive (and predictive?) maintenance of AOVs is to be accomplished 

as part of the Air-Operated Valve Program (see ITEM No. 8). 

14 Describe IST procedures for No inservice testing of the air system was indicated, other than post

the air system. Provide maintenance testing. We did note that records of air system moisture 

reference information, if content are produced and we were told trends are monitored. 

possible. I

NUREG/CR-6654 C-54
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THREE MILE ISLAND, UNIT 1, NUCLEAR GENERATING STATION 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

15 Describe IST procedures for Inservice testing of safety-related AOVs is accomplished in accordance 

AOVs. Provide reference with the applicable requirements of the ASME Code (stroke-timing testing 

information, if possible. under no-load conditions for those safety-related AOVs not exempted for 

Safety-related: some acceptable reason). No IST is accomplished on other AOVs. 

Important non-safety

related: 

Non-safety-related: 

16 Describe diagnostic systems, .TMI uses the MOVATS universal diagnostic system for their MOVs. The 

if any, used for AOVs. plant engineers said that they would use a similar or the same system, 

Provide reference adapted to AOVs, but they did not provide specific plans or details for 

information, if possible. their approach to diagnostic testing of AOVs. The TMI draft AOV 

Description of system: program plan indicates that diagnostic testing of AOVs is planned. 

Specifications: 

Data collected and 

frequency of collection: 

Vendor assistance provided, 

if any: 

17 Describe design (and The draft AOV program plan provides for a review of the design bases 

analysis) procedures for and design calculations for AOVs. The engineers reviewed design 

AOVs. Describe how calculations as a result of discovery of suspect design margins in two 

design basis is established AOVs (see the discussion of Report 643 in Item 6 of this table). 

and maintained for AOVs. 

Provide reference 

information, if possible. 

18 Describe analyses and/or We were not told of any testing done by GPU or TMI. Analyses of AOV 

testing for verification of designs were previously described (see previous entries this table) in TMI 

operability during postulated Report 643, REQUIRED THRUST ANALYSIS for two AOVs that 

transient or accident appeared to have deficient design margins. 

conditions. Provide 

reference information, if 

possible. 

19 Describe training for Training regarding AOVs was not described in any detail. The AOV draft 

installation, maintenance, program plan describes training practices to be used but our impression is 

and testing of AOVs. that they are still in the planning stages of specific training on diagnostic 

Provide reference testing of AOVs. Other maintenance training for AOVs involves hands

information, if possible. on, supervised training on sample valves and equipment, in accordance 

with standard training methods. 

20 Describe databases used to TMI uses internal databases such as the MTAN and SE reports, as 

track maintenance, failures, described in Item 5 of this table. The databases are used to track recurring 

events regarding AOV s. failures that require additional m aintenance and/or m odifications. TM I is

C-55 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THREE MILE ISLAND, UNIT 1, NUCLEAR GENERATING STATION

ITEM No. INFORMATION RESPONSE OR INFORMATION 

Provide reference recording failures and analyzing the data for trends and common-cause 

information if possible. events. 

On site: 

Company wide: 

Industry: 

21 Describe the impact of the As with the other plants that we visited, TMI is actively involved in 

Maintenance Rule, 10 CFR compliance with the Maintenance Rule. Their model for guidance is 

50.65 on AOV and air Regulatory Guide 1.160 and NUMARC 93-01. TMI has ranked events in 

system maintenance and terms of calculated core-damage frequency (CDF) and is ranking systems 

testing. Provide reference and components, including AOVs, in terms of their importance to risk. 

information if possible. An undated paper (one figure is dated 7/15/93), TMI-1 PRA INPUT TO 

MAINTENANCE RULE - RISK SIGNIFICANT SYSTEM LIST, 

developed by Risk Analysis Section, C. Adams, includes a description of 

the plant's compliance with the requirements of the Maintenance Rule. 

22 Is PRA data used for Valves have been categorized, in terms of importance to safety and 

predictive maintenance or operations, in accordance with the TMI-1 PRA INPUT TO 

replacement of AOVs? If MAINTENANCE RULE - RISK SIGNIFICANT SYSTEM LIST, 

so, how? developed by Risk Analysis Section, C. Adams. It was noted that 5.3% of 

calculated core-damage frequency was attributed to loss of instrument air. 

Loss of instrument air was ranked sixth in importance of initiating events, 

in terms of CDF (small LOCA was the highest at 18.8% of CDF). 

23 Are AOVs serviced on site, The methods of repair vary with the circumstances. Vendor input is 

serviced offsite, or replaced sought if the engineers need additional information about the design 

as piece-parts if found to history, etc. AOVs and SOVs have been sent out in the past for repair or 

require service? failure research. 

24 Identify and describe the The list of Maintenance Trend Action Notices that were provided to us 

most common recurring indicate the valves and systems that have the most attention due to 

maintenance problem(s) and repeated or common-cause failures. MTAN 93-03 on solenoid valves and 

failures regarding AOVs and MTAN 93-15 on a sluggish valve were discussed with us. 

the air system. What did In addition, the problem described in TMI Report 643, REQUIRED 

you see? Provide reference THRUST ANALYSIS, for two AOVs that had been calculated to have 

information if possible. insufficient design margins, although not covered by an MTAN, is 

significant because of the generic considerations. 

25 Interviewer comments No AOVs undergoing maintenance were seen during our brief visit inside 

regarding actual valves the plant. 

viewed during the visit, in 

the plant, undergoing 

maintenance or 

replacement, or in the plant 

stock system, if applicable 

to this interview.

NUREG/CR-6654 C-56
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

QITT VISTIT TO THREE MILE ISLAND. UNIT 1 NUCLEAR GENERATING STATION

ITEM No. INFORMATION RESPONSE OR INFORMATION 

26 Has the plant made changes Several SOVs were changed to improve reliability. The old valves had 

to valves or systems that deteriorated O-rings. See MTAN 93-03. 

include AOVs, or replaced As noted above, a second instrument air system and filters upstream of 

AOVs with different models "critical" valves were added. This was prompted by the lack of 

of AOVs or different valves redundancy and poor performance of the "old" service air system. 

that are not AOVs? If so, 

describe the changes and the 

circumstances. 

What prompted the change? 

Was the change made for 

this plant only? 

27 Does the plant follow The recommendations of EPRI/NMAC, along with other industry groups, 

EPRI/NMAC guidelines for is reviewed for applicability to the plant. TMI has incorporated a number 

maintaining AOVs and the of improvements to enhance the quality of instrument air. 

air system(s)? If not, 

describe differences and 

reasons for the differences. 

Provide reference 

information, if possible. 

28 What is the plant doing or TMI reviewed the Industry correspondence no specific response is 

planning to do in response to planned. 

the recent Industry 

correspondence on AOVs. 

Provide reference 

information, if possible. 

29 Do you have any The B&W Owner's Group Report (47-1165965-00) dated December 

suggestions for improving 1986, INSTRUMENT AIR SYSTEM REVIEW REPORT, was provided 

the performance of AOVs, to us. Gordon Skillman, who was part of the B&W Safety and 

particularly in the areas of Performance Improvement Program team discussed the report with us. A 

surveillance, testing, or number of recommendations in the report are pertinent to all plants as well 

maintenance? as B&W designs. TMI used a number of the recommendations in 

modifications to its air system. The specific recommendations for TMI 

included: 

- In a major loss-of-air event (e.g., air pipe break), high air flow rates 

pass through the driers and filters. The desiccant could break up 

and clog the filters. A bypass procedure or automatic bypass of the 

dryers and filters, in a major loss-of-air event, was recommended. 

- The local AOV accumulators were not being checked for 

operability. The recommendation was to inspect the accumulators 

and their associated check valves. 

- Local accumulators had no water blowdown method available to 

remove water accumulation. Any water present reduces the

C-57 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THREE MILE ISLAND, UNIT 1, NUCLEAR GENERATING STATION

ITEM No. INFORMATION RESPONSE OR INFORMATION 

available air volume. The recommendation was to check the 

accumulators for water (at the next outage). 

Failure of an air compressor could lead to a rapid decrease in air 

system pressure if the system has no reverse flow check valve 

between each compressor and the rest of the system. It was 

recommended that a particular valve (IA-V 12) should be positioned 

normally closed. 

Refer to item 32 for additional comments on the actions taken in response 

to the B&WOG report. 

30 Provide a list of 10 CFR A list of 50.59 reports is provided in Item 5 of this table. No 50.72 reports 

50.59 and 10 CFR 50.72 were provided. 

reports on AOVs and AOV 

support systems (air or inert 

gas supply, etc.) that have 

been issued for this plant. 

31 What thrust or torque TMI expects margins better than 10%. They are in the process of 

margins are expected for categorizing valves as those having 10% or greater margin (acceptable), 

AOVs? Are different those with 5% to 10% margin (to be improved, if practical), and those 

margins used for safety- with less than 5% margin (unacceptable, must be improved). They are in 

related, important non- the planning stages of this process, for the most part. Refer to the TMI 

safety-related, or non-safety- program plan for AOVs. 

related AOVs? 

32 What maintenance or No maintenance, other than corrective maintenance is performed on 

surveillance is done to AOV accumulators. The B&WOG Report on air systems noted the need to 

accumulators to ensure ensure that accumulators are clean and dry. Generic Letter 88-14 requires 

air/nitrogen quality and verification of the capability of air-operated components to perform their 

pressure? Were seismic safety function. A TMI-1 50.59 report, SE 412512-005, indicates that 

considerations and size several vent valves were added to the air system for testing the functional 

verified? capability of accumulators. However, no inspection of the interior of the 

accumulators or means for blowdown is indicated. 

The engineers at TMI said that they added drains to those accumulators 

that had moisture in them when they applied the recommendations in the 

B&WOG report; however, they had made no modifications to other 

accumulators. 

33 Describe problems with SE No. SE-CMR-l113202, Replacement of Failed IWT Regulators, 2/5/90 

pressure regulators, if any. indicated that several failed pressure regulators were replaced although the 

cause of failure was not noted. Several of the 50.59 reports noted 

previously above describe the installation of filters upstream of critical 

valves. These would also protect the positioners and regulators associated 

with the valves.

NUREG/CR-6654 C-58
Appendix C

NUREG/CR-6654

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THREE MILE ISLAND, UNIT 1, NUCLEAR GENERATING STATION 

ITEM No. INFORMATION I RESPONSE OR INFORMATION 

34 Describe problems with There were no problems since the instrument air system improvements 

feedwater regulating valves, were made. 

if any. 

35 What, if any, is your The plant engineers have a representative in the AOVUG who is also an 

involvement with the AOV active participant in the Joint Owners Group AOV efforts. 

Users Group? Describe.

C-59
Appendix C 

Trip No. 6 REPORT 

STUDY OF AIR-OPERATED VALVES 

INDIAN POINT, UNIT 3, MARCH 10 AND 11, 1998 

We had two days of meetings with the engineers concerned with the air systems and air-operated 

valves (AOVs) at Indian Point 3 (IP3). We were escorted through portions of the plant and saw some of 

the air systems and AOVs in the plant. Indian Point 2 (IP2) is adjacent to IP3 but is owned and operated 

by another company, Consolidated Edison. Indian Point 1 (IP1), also adjacent to IP3, is shut down. IP3 

was operating at full power at the time of our visit. 

IP3 has maintenance engineers and system engineers who are charged with maintaining the 

operational readiness of the air system and the operability of AOVs for the plant. 

One portion of the instrument air (IA) system at IP3 is served by two 225 scfm compressors (IAC 

#31 and #32) that serve a common receiver. Two banks of two desiccant-type dryers are downstream of 

the receiver. A cross connection to the station air system is upstream of the dryers to allow use of station 

air in an emergency. A third 350 scfm IA compressor (IAC #33) was added to the system and has its own 

receiver and a desiccant-type dryer. Two other IA compressors (designated Admin #31 and #32) were 

also added to the system and each of these has its own receiver and one desiccant-type dryer. The #31 

and #32 IA system compressors and dryers can be loaded on to the station diesel generators in the event 

of a loss of offsite power. 

The service air (SA) system is served by a compressor, a backup diesel powered compressor, and a 

tie-in from IP2. The diesel SA compressor has a deliquescent-type dryer downstream. A separate 

polisher building was added and contains a separate polisher air system. It is served by two redundant 

400 scfm compressors (Polisher #31 and #32). The polisher air system has one air receiver and a bank of 

two refrigerant-type air dryers. 

The components or systems essential to plant safety and serviced by the IA system are: 

I. containment isolation valves, 

2. cooling water valves for containment building fan coolers, 

3. condensate storage tank shut-off valves, 

4. auxiliary boiler feed pump control valves, 

5. steam dump valves to atmosphere, 

6. low pressure steam dump valves to main condenser, 

7. weld channel and containment penetration pressurization system (WCCPPS), 

8. emergency diesel cooling water valves, 

9. spray additive tank outlet valves, 

10. boron injection tank recirculation valves, and 

11. control room air conditioning dampers.

NUREG/CR-6654 C-60
Appendix C

In the event of low pressure in the IA system, air is automatically supplied to the IA system from 

the SA system. In the event of low pressure at any or all of the components listed as items 4, 7, or 11, dry 

nitrogen cylinders automatically supply gas pressure to those components required for safe shutdown or 

continued plant operation. In addition, a manual supply of dry nitrogen is available, in the event of low 

IA system pressure, to operate the steam dump valves. In the event of an IA system line rupture in the 

conventional plant, a restriction orifice limits the flow to 225 scfm. One compressor supplies air to 

compensate for the break and the spare compressor supplies the primary plant. 

Originally, the IP3 IA system had carbon steel piping upstream of the dryers and copper piping 

downstream. A major portion of the carbon steel piping was replaced with stainless steel, with the 

exception of the pipe connections near the aftercoolers, and a section of pipe for the IA containment 

isolation valve. 

The weld channel and containment penetration pressurization system (WCCPPS) is an unusual 

innovation. This system provides gas. to the containment penetrations and the liner weld channels at 

higher than containment pressure in order to provide an additional barrier against containment leakage. 

IP2 has a similar system. Zion also has such a system, although of different design. We were told that 

moisture intrusion from the instrument air system, prior to actions taken by the licensee to improve air 

system quality, caused some contamination of this system, some corrosion damage to valves and piping, 

and back leakage of moisture into portions of the nitrogen supply piping. These problems were corrected 

by upgrading the original IA system. 

The IA system at IP3 is equipped with continuous dew point monitors with digital readout on a 

small instrument adjacent to the compressors. This readout is monitored frequently and anomalies are 

noted. In addition, if the dew point exceeds -1 50F, an alarm is activated in the control room. 

IP3 has 578 AOVs in their plant. Of these 263 AOVs are classified as safety-related. Also, 215 of 

the 578 AOVs are classified as within the scope of the Maintenance Rule. By way of comparison, 89 

motor-operated valves are within the scope of Generic Letter 89-10. 

IP3 initiated a program plan to improve the performance of their AOVs based on failure rates that 

they estimated to be twice the industry average. This plan includes categorization of AOVs by 

importance, engineering evaluations to establish design requirements, and extensive use of diagnostic 

testing to identify maintenance requirements. They intend to implement their program fully by 1999. 

The cooperation, courtesy, and knowledgeable responses from the members of the IP3 engineering 

and technical staff were noted and appreciated by those of us who are involved in this study of AOVs. 

The following tabular summary was prepared to describe the information gathered during the 

subject visit. The NRC, with assistance from the INEEL, is studying the performance of air-operated 

valves (AOVs) in commercial nuclear power plants. The information was collected in accordance with 

the Program Plan dated 10/22/97 (INEEL Letter to H. Ornstein, NRC, from J. Bryce, 10/23/97, Job Code 

E8238, Task Order 15 - JHB-167-97).

C-61 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT. UNIT 3. NUCLEAR POWER PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

1 Date. March 10 and 11, 1998. 

2 Name of Interviewer. Hal Ornstein, NRC/AEOD, 301/415-7574 

Owen Rothberg, INEEL/Rockville, 301/816-7773 

John Watkins, INEEL/Idaho Falls, 208/526-0567 

Jit Vora, NRC/RESEARCH, 301/415-5833 

3 Plant Name & Docket No. Indian Point, Unit 3, Docket No. 50-289.

4

________ L

Person(s) Interviewed, 

Title(s), Phone Number(s), 

E-Mail address, short 

description of 

organization(s) and duties.

Plant Address is: Indian Point 3 Nuclear Power Plant, P.O. Box 215, 

Buchanan, NY 10511. Phone numbers listed below are 914/736-xxxx 

unless otherwise noted. 

K. Peters, Licensing Manager, IP3, x8029 

J. Comiotes, General Manager, Operations, IP3, x8002 

J. Odendahl, Acting General Manager, IP3, x8701 

M. Carmichael, QA Manager, IP3, x8501 

R. Schmitt, Maintenance Eng. Supervisor, IP3, x8632 

K. Eslinger, Systems Eng. Supervisor, IP3, x8993 

M. Dinelli, Performance Engineer, IP3, x8315 

Jim Werner, Systems Engineer, IP3, x8319, FAX 914/734-6031 

Joe DeRoy, Director of Engineering, IP3, x8006 

Bob Barrett, Site Executive Officer, IP3, x8001 

Victor P. Rizzo, Maintenance Engineer, IP3, x8536 

Tat Chan, System Engineer, IP3, x8874 

Michael J. Dries, System Engineer, IP3, x8382 

Becky Green, ORG Engineer, IP3 

Bob Dolansky, IST Engineer, IP3, x8458, FAX x8350 

Andrew Mihalik, Senior Systems Engineer, x8362 

Anthony Dicescaro, Engineer involved with Maintenance Rule 

implementation 

David Lew, NRC Senior Resident Inspector, IP3, P.O. Box 337, 914/739

8565, FAX 914/739-8624 

Laura Dudes, NRC Resident Inspector, IP3, P.O. Box 337, 914/739-8565, 

FAX 914/739-8624

NUREG/CR-6654 C-62
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT, UNIT 3, NUCLEAR POWER PLANT 

ITEM No. INFORMATION [ RESPONSE OR INFORMATION

If necessary, and if 

person(s) interviewed can 

do so, obtain any missing 

information not provided 

prior to the site visit, as 

described in the outline for 

Task 4 of the program plan. 

Note what information was 

provided.

5

NUREG/CR-6654

A copy of our program plan was provided to IP3 prior to the visit. 

Slides of the IP3 AOV Program Overview by Victor Rizzo, IP3, and slides 

of the IP3 Field Testing Experience by Rusty Gasser, Crane-Movats, 

provided to the NRC TM at a MOV/AOV conference in early December 

1997, were reviewed before the visit and used in the discussions at the 

plant. 

We arrived with copies of the following materials: 

- Rev. 2 of Section 9.6.3 of the IP3 FSAR, Compressed Air System 

(including a flow diagram of the instrument air (IA) system) 

- General facility description taken from the NRC Internet home page 

- LER 28685002, Reactor Trip Resulting from Feedwater Regulating 

Valve Positioner Failure 

- LER 28685006, Reactor Trip Resulting from Feedwater Regulating 

Valve Positioner Failure 

- LER 28688006, Reactor Trip Resulting from failure of Main 

Turbine Air Pilot Valve Failure Due To Foreign Materials on the 

Valve Internals 

- LER 28688009, Failures of ASCO Solenoid Valves Due to 

Lubricant Problems 

- LER 28693013, Damper Failures Caused by Dirty Linkages 

- LER 28693036, Improper Seismic Mounting of Damper Actuators 

in the Control Room 

- LER 28693045, Failure Mode of Central Control Room Dampers 

Upon Loss of Instrument Air 

LER 28693050, Solenoid Valves Can Be Over-Pressurized 

- LER 28696002, Loss of EDG After Loss of Offsite Power Due to 

Pressure Regulator Failure for Ventilation Control System 

- LER 28696004, Two Diaphragm AOVs in Series Discovered to be 

Incapable of Closing at System Pressure 

- LER 28696008, Nitrogen Pressure Below Minimum to Adequately 

Supply Isolation Valve Seal Water System 

- LER 28696013, AOV Prevented From Fully Opening Due to Crud 

Buildup Inside Valve 

(continued next page)

C-63
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT, UNIT 3, NUCLEAR POWER PLANT 

ITEM No. INFORMATION [ RESPONSE OR INFORMATION

5 (cont.) Information Provided The engineers at IP3 provided us with the following information: 

(continued) - An informal sketch of the Unit Air System (for information only) 

- IP3 Drawing #9321-F-20363, Rev. 52, Flow Diagram, Instrument 

Air, Sheet 1 

- IP3 Drawing #9321-F-20363, Rev. 4, Flow Diagram, Instrument 

Air, Sheet 2 

- IP3 Drawing #9321-F-27233, Rev. 36, Flow Diagram, Nitrogen to 

Nuclear Equipment 

- IP3 Drawing #9321-F-27623, Rev. 39, Flow Diagram, Penetration 

& Liner Weld Joint Channel pressurization System 

- NYPA Memorandum RET-96-3 11 dated 11/5/96, IP3 Maintenance 

Rule Basis Documentation Revised Importance Ranking with a 47

page table attached entitled IP3 IPE Risk Achievement/Reduction 

Worths 

- Table entitled Air Operated Valve Listing by system-component 

dated 3/10/98, 65 pages 

- Table entitled Air Operated Valve Listing by manufacturer

component dated 3/10/98, 65 pages 

- IP3 Action Plan IDSE-APL-96-061, ACTS No. 96-22957, Rev. 0, 

dated 1/6/97, entitled The Return of 32 IA Compressor to 

Maintenance Rule Status "(a)(2)" and the Resolution of 32 & 32 IA 

Compressor Replacement or System Upgrade 

- Customer Service Guide and Technical Data for Drain-All 

Instrument Air Drain Traps, (3/17/94) 

- IP3 Engineering Memo dated 5/5/95, Subject: Evaluation of PCV

13 10A and B (MSIVs) Operation and Performance This memo 

describes the resolution of a concern with the operation of the 

MSIVs 

The plant engineers sent us a box of event reports and other data after the 

trip (5/28/98). A number of work requests were included that they 

considered to be of interest for this study of AOVs. These were: 

- WR 93-10265-00, Replace diaphragm on aux. feed regulating 

control valve operator (PCV- 131 OA). 

- WR 93-10284-03, FCV-447-SOV-1 could be over-pressurized. 

- WR 94-00005-03, CVCS SOV-LCV-1 12A could be over

pressurized. 

(continued next page)

NUREG/CR-6654 C -64
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT, UNIT 3, NUCLEAR POWER PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION

5 

(cont.)

Information Provided 

(continued)

- WR 94-02963-04, Diaphragm leak in 32 cation bed first injection 

AOV. 

- WR 94-04801-00, PCV- 1193 in WCCPPS does not respond to 

adjustments. Regulator does not control pressure. 

- WR 95-04694-02, Follow-up to WR-04801-00 on PCV-1 193. 

- WR 96-00020-00, Air leak on air-regulator to AOV FCV-1 173. 

- WR 96-01430-00, Valve 4EX-2 operator stuck. 

- WR 96-01717-01, Excessive air pressure required to stroke AOV 

LCV0 1127C, heater drain tank to condenser #32 bypass. 

- WR 96-01770-01, Repair pressure regulator for PCV-1200 in 

WCCPPS. 

- WR 96-02488-02, SG#31 MSIV supply SOV air leaks. 

- WR 96-03865-00, IA regulator to MS-FCV-1 155 has a diaphragm 

leak. 

- WR 96-05736-00, #31 anion bed rinse control valve PM inspection 

request on AOV. 

- WR 96-06664-00, CVCS FCV-l 1 A strokes slowly. 

- WR 97-00797-00, #32 condenser high level control valve packing 

leak. (Package contains IP3 Generic Valve Packing Procedure). 

- WR 97-04166-06, Need to blow down piping in WCCPPS. 

Repaired and cleaned pressure regulator. 

- WR 97-04166-07, Need to blow down piping in WCCPPS. Rebuilt 

pressure regulator. 

- WR 97-04684-00, SGBDR recovery outlet PCV erratic behavior 

causes flow and pressure spikes. (Package contains Copes-Vulcan 

instruction manual for diaphragm-actuated control valve.) 

- WR 97-05214-32, #32 cation bed backwash control valve air 

operator, bad diaphragm. 

- WR 97-05643-01, Valve BFD-FCV- 1116 identified to be leaking. 

- WR 97-05953-46, Primary water containment isolation valve failed 

stroke test. 

- WR 97-05957-35, Pressure regulator gage for N2 supply broken. 

- WR 97-05957-81, Cation regeneration vessel backwash inlet valve 

positioner leaking air. 

- WR 97-05957-82, Resin mixing and hold vessel rinse inlet valve 

positioner leaking air. 

- WR 97-05957-84, Cation regeneration vessel rinse inlet valve 

leaking air on positioner. 

(continued next page)

C-65 C6NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT, UNIT 3, NUCLEAR POWER PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

5(cont.) Information Provided WR 97-05957-85, Resin mixing and hold bottom water inlet valve 

(continued) positioner leaking air. 

WR 97-06959-01, Resin mixing and hold vessel rinse and drain 

outlet valve leaked air. 

- WR 97-06959-40, SGBDR purification system bypass valve supply 

pressure regulator broken. 

- WR 98-00506-00, #32 anion bed rinse control valve air operator 

flange leak. 

6 Describe plant events We discussed the LERs and other events listed in item 5, above. Problems 

involving AOVs and included solenoid valve lubrication, dirt in working parts of valves, and 

provide reference mechanical parts failures. Problems with packing friction and other 

information, if possible. maintenance adjustments are common. 

Recent: 

Recurring: 

Significant: 

The plant engineers sent us a box of event reports and other data after the 

trip (5/28/98). Events or conditions of interest for this study of AOVs 

included the following items: 

- DER 93-734, SOVs can fail due to over-pressurization. This DER 

and associated LER 28693050 indicate that a number of non-safety

related IA regulators were installed. Over-pressurization of SOVs 

[exceeding designed maximum operating pressure differential 

(MOPD)] was not recognized by the plant designers as a failure 

mechanism. Generic Letter 91-15, Information Notice 88-24, and 

NUREG- 1275, Vol. 6, described the problem but the plant 

engineers did not recognize the problem and make the necessary 

changes promptly. In addition to replacement of SOVs 1276 and 

1276A, corrective actions were taken for 109 SOVs having 

insufficient MOPD. 

- DER 94-097, Failure of Appendix R emergency diesel air-start 

motor supply valves, AOV-1 12 and -113 due to inadequate operator 

pressure rating. This was an original plant design mistake. 

- DER 94-109, AOV-1 13 was reinstalled in wrong orientation 

(includes an analysis of the event and critique of procedures). 

- DER 94-397, The controls (SOVs) for AOVs FCV-1204 and -1205, 

quench-water supply valves to the 31 heater drain pump, were not 

the proper design for the intended function and had to be replaced. 

- DER 94-894, Feedwater valve stuck open. 

- DER 94-1105, Improper maintenance and reassembly of AOV 

BFD-FCV-406 C and D. 

DER 95-0033, Critique of post-installation tests of diaphragms on 

Copes-Vulcan D-100 AOVs.

NUREG/CR-6654 C-66
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT, UNIT 3, NUCLEAR POWER PLANT 

ITEM No._I INFORMATION RESPONSE OR INFORMATION

6 (cont.) Describe plant events DER 95-0233, Investigation of the use of Teflon tape on pipe 

(continued) threads in the plant. Includes a number of background references. 

Contamination of stainless steel piping from migration of flourides 

in Teflon products and blockage of small orifices from "plate-out" 

of Teflon on hot surfaces are maj or concerns. 

DER 95-0395, Investigation of the (incorrectly) reported failure of 

the station air/instrument air backup valve (PCV- 1142). 

DER 95-1302, Failure of PCV- 1310A, Main Steam Isolation Valve 

to Aux. FW Pump, to stroke. See DER 96-2747. 

DER 95-1392, Spring washers were shown on a drawing for SWN

PCV- 1205 through 1210 but were deleted by the manufacturer. The 

washers are not needed and are not installed. 

DER 95-2244, High failure rate for instrument air dewpoint probes 

was investigated. 

DER 95-2317, Failure of CH-AOV-204B, normal charging 

isolation valve to open. Operator tapped the SOV and the valve 

operated normally afterward. Results of investigation not included. 

DER 95-2517, Fail-safe position of valve WD-AOV-1610 listed in 

the FSAR did not agree with system drawings. 

- DER 96-0064, Normal charging valve, CH-AOV-204B, failed to 

open. No analysis or cause included. 

- DER 96-2056, Investigation of crud buildup in service water system 

valves (PCV-1296 and 1297, 2 inch WKM Type 70-28-1D pressure 

control valves) which resulted in inadequate service water supply to 

the control room AC system. LER 28696013 was included. NRC 

Generic Letter 89-13, "Actions for Low-Flow SWS Lines 

Susceptible to Crudding" applies. DERs 96-1267, 96-1422, 96

1756, 96-1763, 96-1776, 96-1779, and 96-1780 were noted as 

related. 

- DER 96-0393, AOV, BFD-FCV-406A, B, and C, feedwater 

regulating valves did not stroke smoothly without system pressure 

and did not open fully during testing., The I/P controllers were 

recalibrated and the "popping" operation of the valves was 

investigated and subsequently considered satisfactory. See DER 

96-0812 for a comprehensive history of these and related AOVs. 

- DER 96-0416, An exhaustive discussion of the merits and 

objections to protecting accumulator vent valve, HCV-943, from 

foreign material intrusion. DERs 95-1299 and 94-0809, both 

referring to the same issue, were included. 

(continued next page)

C-67 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

QrrjQ T XTT T- TMnTAN POINT UNIT 3. NUCLEAR POWER PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION

6 (cont.) Describe plant events - DER 96-0557, An AOV, DA-AOV-112, DG starting air inlet 

(continued) control valve, failed because a label was applied by the licensee 

over the (very small) vent port. Other similar valves were 

investigated to ensure that the problem was resolved. Receipt 

inspection procedures were modified. Drawings of the valve were 

included. 

DER 96-0687, A nitrogen pressure regulator for AOV PCV-1 196 in 

the WCCPPS was replaced with a regulator from IP-2. Debris and 

corrosion products in the re-used regulator caused it to leak during a 

post-installation test. It was inferred that the deteriorated condition 

should have been discovered prior to installation. 

DER 96-0812, This DER contains a history of the leakage problems 

associated with the auxiliary feed pump discharge flow control 

valves FCV-406C and D, and the plant operation problems caused 

by their excessive leakage. It appears that the basic problem is that 

the operators are undersized for maximum pressure conditions. See 

DER 96-0393 for another problem with these or related AOVs. 

DER 96-1029 indicated that FCV-406-A and D stuck on opening 

with no pressure in the line. 

DER 96-0696, Fine mesh screens were apparently supposed to have 

been provided in the inlets of Weld Channel and Containment 

Penetration Pressurization System (WCCPPS) valves PS-PCV

1194, 1196, 1198, and 1200. The screens were found to be not 

required because filters were provided when regulators were 

changed. The design-basis documentation was updated. 

DER 96-0905, Failure of main turbine stop valve equalizing valve 

to close. The horizontal orientation of the valve caused stem 

binding. The AOV manufacturer (Copes- Vulcan) recommended 

additional bushings and packing redesign to reduce stem binding. 

- DER 96-1029, AOV BFD-FCV-406A was reported to have stuck 

closed on a demand to open with no pressure or little open-direction 

pressure on the valve seat. DER 96-1367 (copy not provided) was 

referenced in this package and was described as the document for 

tracking implementation of an AOV program. 

DER 96-1267, CCRAC pressure control valve, SWN-PCV-1297 

moved to mid-position on demand to move to full open and resulted 

in a 72-hour LCO condition. A blocked sensing line in the 

pneumatic controller was cleared. 

(continued next page)

NUREG/CR-6654 C-68
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT, UNIT 3, NUCLEAR POWER PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION

6 (cont.) Describe plant events DER 96-1422, Repair of sensing line and pneumatic controller for 

(continued) PCV-1297 (DER 96-1267). Dirt from the service water system 

clogged the line. In addition, a mistake was found in the calculated 

effective pressure drop on the manufacturer's data sheet. This AOV 

is a WKM 2 inch model 70-28-1D pressure control valve. Refer to 

DER 96-1776. See DER 96-1756 for repair of the controller. 

DER 96-1674, No. 32 EDG starting air pressure regulator DA

PCV-14-3 was not maintaining pressure. Foreign material was 

found in the space between the diaphragm and the valve seat. 

Debris from the air start system (corrosion product perhaps?) was 

suspected. The foreign material was removed by cycling the valve 

open and closed. The corrective actions note that the air tank was 

scheduled to be cleaned and piping had been changed to stainless 

steel to prevent foreign material intrusion in the future. 

DER 96-1756, Repair of pneumatic controller PC-1432 for AOV 

PCV-1297 (see DERs 96-1267 and -1422). Several components 

and linkages were broken and an orifice was clogged. 

DER 96-1776, Valve internals were removed from PCV-1297 in 

order to obtain required flow. Investigation confirmed that the 

manufacturer's original data sheet was not correct. See DER 96

1267 and -1422. LER 28696013 refers to, and was included in this 

package, but does not explicitly mention the mistake in the 

manufacturer's data sheet. 

DER 96-1779, Critique of valve vendors data and plant procedures 

related to insufficient flow in AOV PCV-1297. See DERs 96-1776, 

-1422, and -1267. 

DER 96-1780, AOV PCV-1297 was not stroke tested prior to retest. 

See DERs 96-1267, -1422, -1776, and -1779. Package includes 

control room AC diagrams and a copy of the General 

Troubleshooting Procedure (IC-AD- 13) for IP3. 

DER 96-1973, The FSAR indicated that a single failure in the Weld 

Channel and Containment Penetration Pressurization System 

(WCCPPS) would not prevent fulfillment of the design function of 

the system. On 8/30/96, the plant engineers realized that the FSAR 

was incorrect; a single failure of the air or nitrogen regulators would 

render the WCCPPS incapable of performing its design basis 

function. The primary function of the WCCPPS is to prevent 

leakage from the containment to the surrounding environment. An 

operability determination investigation is included in the package 

and the conclusion was that the single air or nitrogen regulator 

failure would not impact public health and safety. 

(continued next page)

C-69 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT, UNIT 3, NUCLEAR POWER PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION

6 (cont.) Describe plant events 

(continued)

- DER 96-1985, Evaluation of the differences and similarities 

between PCV-1310B and A. PCV-1310A did not open properly but 

PCV-131OB did. These are two identical AOVs in series. These 

are steam isolation valves to the 32 auxiliary boiler feed pump 

turbine. See DER 95-1302 and 96-2747. 

- DER 96-2056, Debris found in PCV-1296. See DER 96-1267, 96

1780, and others that refer to these AC compressor control valves. 

- DER 96-2239, Failure of nitrogen pressure regulator in Weld 

Channel and Containment Penetration Pressurization System 

(WCCPPS). LER28696010 was included. Particulate 

contamination in the system was suspected to have caused the 

regulator failure. 

- DER 96-2747, Investigation of the failure-to-open of PCV-13 1A, 

Steam Isolation Valve to Aux. FW Pump, failure to open during a 

stroke-timing test. This DER included a detailed analysis of the 

operation, maintenance, and design of the AOV. 

Other DER reports in the box sent on 5/28/98 included: 94-0398, 94-0765, 

95-0384, 95-0561, 96-1763, 96-2241, and 96-2649. 

Also included in the box sent on 5/28/98, along with the event reports, 

were: 

- Listing of DER history on IA systems through 3/2/98 

- Listing of DER history on AOVs through 3/2/98 

- Listing of maintenance work history on air operators through 3/2/98 

- Listing of maintenance work history on SOVs through 3/2/98 

- Listing of DER history on SOVs through 3/2/98 

- Listing of maintenance work history on compressors through 3/2/98 

- Listing of maintenance work history on regulators through 3/2/98

- Listing of maintenance work history on valves

7 Describe AOV or air IP3 initiated an AOV program because their failure rate was found to be 

system actual or detected about twice the industry average over the past few years. See the slides of 

potential failures at the the IP3 AOV Program Overview by Victor Rizzo, IP3, and slides of the 

plant? Provide reference IP3 Field Testing Experience by Rusty Gasser, Crane-Movats, provided to 

information, if possible. the NRC TM at a MOV/AOV conference in December 1997. Their 

program includes engineering reviews, diagnostic testing, and equipment 

overhaul. IP3 made a number of repairs and adjustments to their AOVs 

based on this program. 

8 Describe actions taken after Before the late 1980s, IP3 had numerous problems attributable to 

events or failures involving contaminated air (mostly moisture) in the past. A number of 

AOVs or the air system. improvements were made including on-line dew point monitoring and

NUREG/CR-6654 C-70
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT. UNIT 3. NUCLEAR POWER PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

Provide reference alarms in the control room, use of automatic drains, and replacement of 

information, if possible. carbon steel piping with stainless steel. It appears that IP3 now provides 

clean, dry air to AOVs in the plant, thereby eliminating the potential for 

common-cause AOV failures from instrument air contamination. 

9 Were there any actual or One very recent event that occurred at Dresden and Quad Cities involving 

potential common mode or potential failure of AOV diaphragms on Copes-Vulcan Dl100 valve 

common cause failures in operators was discussed with the maintenance engineers. The problem is 

the air system or AOVs at that the elastomer covering over the fibers in the diaphragms is too thin 

the plant? Describe and and the diaphragms wear out as the valve is operated. IP3 engineers had 

provide reference not been aware of this problem and since they have a large number of this 

information, if possible. type of valve they plan to investigate immediately. This is a potential 

common-cause failure mechanism. 

Another problem, mentioned above, involved the lubricant used on ASCO 

solenoid valves. This problem was reported at other plants and is covered 

in NRC correspondence. 

There appears to be no way to determine the internal condition of AOV 

accumulators, including the presence of corrosion products or moisture. 

This is a common problem at most, if not all, of the plants visited. 

10 Describe root cause IP3 has a formal root cause analysis procedure (Administrative Procedure 

analysis procedures for the 8.2) based on standard industry practices and references. Except for some 

plant. Provide reference of the early ones, LERs identify root causes. 

information, if possible. 

11 Describe root-cause Several root-cause analyses were described to us including the IP3 

analyses performed for air determination of problems having to do with ASCO solenoid valves and 

system or AOV failures at their subsequent replacement. The plant's adoption of diagnostic system 

the plant. Provide technology indicates their dedication to determining root causes of AOV 

reference information, if failures. 

possible. 

12 Describe maintenance The plant has upgraded their air system over the past few years to improve 

procedures for the air capacity and ensure air quality. As noted above, the air system has been 

system. Provide reference equipped with adequate drying capacity and filtration, both general and 

information, if possible. local, to the valves. A major portion of the carbon steel piping between 

the 31 and 32 IA compressors and the copper air headers was replaced 

with stainless steel. One of the plant's objectives was to reduce the 

number of compressors and dryers, while increasing capacity, and to 

replace compressors with less maintenance-intensive ones. 

13 Describe maintenance AOV maintenance procedures are to be based on diagnostic testing. IP3 

procedures for AOVs. has a two-phase program in place to determine the condition of their 

Provide reference AOVs. The first phase is a pilot program for selected AOVs and is based 

information, if possible. on their refueling outage needs, their priorities from the Maintenance Rule, 

1 Safety-related: safety-related importance, and plant efficiency. The second phase, as yet

C-71 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT, UNIT 3, NUCLEAR POWER PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

to be fully implemented, includes a comprehensive program of industry 

Important non-safety- involvement, engineering reviews, diagnostic verification, and training. 

related: 

Non-safety-related: 

14 Describe IST procedures No IST is done for the air system because it is considered to be a non

for the air system. Provide safety related system. 

reference information, if 

possible. 

15 Describe IST procedures IST procedures for AOVs consist of ASME stroke-time testing of safety

for AOVs. Provide related AOVs. IP3 uses the provisions of Section XI of the 1983/Summer 

reference information, if 1983 ASME Code. They are also implementing provisions of OM-10 as 

possible. allowed by NUREG-1482. No other IST is mandated. The remark was 

made that it is rare to pick up an incipient valve failure during a stroke

Safety-related: timing test. The IP3 program for AOV diagnostic testing is not mandated 

Important non-safety- by IST considerations. Rather it is being driven by high failure rates and 

related: their Maintenance Rule and QA requirements. 

Non-safety-related: 

16 Describe diagnostic IP3 uses the CRANE/MOVATS diagnostic system for AOVs. Their AOV 

systems, if any, used for program includes heavy dependence on diagnostic testing to identify 

AOVs. Provide reference maintenance and adjustment problems with AOVs. They are doing 

information, if possible. engineering evaluations to determine design basis requirements and it is 

Description of system: not clear how they plan to relate that to available design margins. They 

will have to rely on the manufacturer's data to determine the designed 

Specifications: capability of their AOVs. Their program will at least ensure that their 

Data collected and AOVs are capable of opening and closing without loss of margin due to 

frequency of collection: misadjustment or deterioration. IP3 cannot assess margins for AOVs 

without knowing the design bases. If the manufacturer's data and methods 

Vendor assistance are used without verification, the design bases and the margins for the 

provided, if any: AOVs may not be conservatively estimated. 

17 Describe design (and The AOV program plan calls for a review of the design basis and a design 

analysis) procedures for review. Additional calculations may be required. The design basis was 

AOVs. Describe how established when the plant was designed and the calculations of the valve 

design basis is established manufacturers will have to be reviewed to try to establish if acceptable 

and maintained for AOVs. margins were provided. This effort is not complete at IP3. 

Provide reference 

information, if possible. 

18 Describe analyses and/or No testing is being done or is planned to determine design-basis 

testing for verification of operability. Analyses, or examination of existing analyses, is planned as 

operability during part of the AOV program plan for a number of safety-related and 

postulated transient or important valves and operators. 

accident conditions. 

Provide reference 

information, if possible.

NUREG/CR-6654 C-72
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY SITE VISIT TO INDIAN POINT. UNIT 3. NTICJ.EAR POWER PILANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

19 Describe training for Training improvement is part of the AOV program plan. The training was 

installation, maintenance, described as mostly "on-the-job" with close supervision. 

and testing of AOVs. 

Provide reference 

information, if possible. 

20 Describe databases used to Copies of several printouts from the databases for AOVs were provided. 

track maintenance, failures, Sorts by manufacturer and system were provided and it is possible to 

and events regarding construct sorts in a wide variety, including safety-related, containment 

AOVs. Provide reference isolation, PRA ranking, etc. 

information if possible. 

On site: 

Company wide: 

Industry: 

21 Describe the impact of the As with all of the plants visited, the impact of the Maintenance Rule is 

Maintenance Rule, 10 CFR apparent. IP3 has had to determine those valves important from a risk 

50.65 on AOV and air perspective and compare those valves with their safety-related AOVs. In 

system maintenance and addition, the plant realized that a number of AOVs are significant from an 

testing. Provide reference economic perspective. The AOV program plan is based heavily on the 

information if possible. rankings needed to satisfy the requirements of the Maintenance Rule. No 

(a)(1) goals for the air system or AOVs were in place in the summary 

dated 12/30/97 that we looked at. An action plan, dated 1/6/97, for return 

of the #32 IA compressor to (a)(2) status was reviewed. 

22 Is PRA data used for Predictive maintenance is being considered by the plant and PRA data is 

predictive maintenance or now used for ranking the importance of AOVs. The AOV program plan 

replacement of AOVs? If does not include predictive maintenance or replacement of AOVs based on 

so, how? PRA data or techniques. 

23 Are AOVs serviced on site, IP3 services valves on site or offsite, depending on the individual 

serviced offsite, or replaced circumstances. Solenoids are generally replaced as piece-parts, although 

as piece-parts if found to some have been sent to the manufacturer for analysis (lubricant problem). 

require service? 

24 Identify and describe the The most common AOV problems, as identified from the diagnostic field 

most common recurring test experience on 10 AOVs, appear to be related to packing, 

maintenance problem(s) seating/unseating difficulties and positioner calibration. 

and failures regarding 

AOVs and the air system. 

What did you see? Provide 

reference information if 

possible. 

25 Interviewer comments Although a number of AOVs were viewed, there were none shown to us 

regarding actual valves that were undergoing replacement or maintenance. The plant was at full

C-73 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT, UNIT 3, NUCLEAR POWER PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

viewed during the visit, in power at the time. 

the plant, undergoing 

maintenance or 

replacement, or in the plant 

stock system, if applicable 

to this interview. 

26 Has the plant made changes The plant has not changed their AOVs although several ASCO solenoids 

to valves or systems that were replaced with Valcor solenoids in an effort to improve performance 

include AOVs, or replaced (sticking). The changes were considered successful. 

AOVs with different 

models of AOVs or 

different valves that are not 

AOVs? If so, describe the 

changes and the 

circumstances. 

What prompted the 

change? 

Was the change made for 

this plant only? 

27 Does the plant follow The EPRI/NMAC guidelines were consulted during the preparation of the 

EPRI/NMAC guidelines for AOV program plan at IP3. 

maintaining AOVs and the 

air system(s)? If not, 

describe differences and 

reasons for the differences. 

Provide reference 

information, if possible. 

28 What is the plant doing or Although not the driving force for preparation of the IP3 AOV program 

planning to do in response plan, the recent industry correspondence and interest in AOVs did serve to 

to the recent Industry focus management attention on the issue. 

correspondence on AOVs. 

Provide reference 

information, if possible. 

29 Do you have any The preparation of a comprehensive program plan appears to be the 

suggestions for improving centerpiece of the IP3 effort to improve AOV performance. Their 

the performance of AOVs, objective is to reduce AOV failures and corrective maintenance. 

particularly in the areas of 

surveillance, testing, or 

maintenance? 

30 Provide a list of 10 CFR No reports were reviewed. 

50.59 and 10 CFR 50.72 

reports on AOVs and AOV 

support systems (air or

NUREG/CR-6654 C-74
Appendix C

NUREG/CR-6654

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO INDIAN POINT, UNIT 3, NUCLEAR POWER PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

inert gas supply, etc.) that 

have been issued for this 

plant. 

31 What thrust or torque Various figures were discussed from 0 to 20%. Certainly no negative 

margins are expected for margins would be tolerated. The plan is to look at the results and decide if 

AOVs? Are different a margin is acceptable based on the individual circumstance. Referring to 

margins used for safety- item 16 in this table, if the design bases are not conservatively established 

related, important non- then the margins will not be meaningful. 

safety-related, or non

safety-related AOVs? 

32 What maintenance or No specific maintenance is done on AOV accumulators. There is no 

surveillance is done to direct way to inspect the conditions inside the accumulators. 

AOV accumulators to 

ensure air/nitrogen quality 

and pressure? Were 

seismic considerations and 

size verified? 

33 Describe problems with Refer to DERs included in supplemental material forwarded on 5/28/98. 

pressure regulators, if any. 

34 Describe problems with Two 1985 LERs indicated problems with FRVs. These problems were 

feedwater regulating related to the poor quality of the air. No recent incidents were discovered. 

valves, if any. Refer also to DERs included in supplemental material forwarded on 

5/28/98. 

35 What, if any, is your The maintenance engineers are active in the industry groups as part of 

involvement with the AOV their implementation of the AOV program plan. This plan was recently 

Users Group? Describe. presented and discussed at an AOV Users Group meeting.

C-75
Appendix C

Trip No. 7 REPORT 

STUDY OF AIR-OPERATED VALVES (AOVs) 

TURKEY POINT, UNITS 3 AND 4, MARCH 24 AND 25, 1998 

We had two days of meetings and interviews with the engineers at Turkey Point 3 & 4 (TP) who 

are concerned with AOVs and the air systems. We were shown portions of the air system, as well as 

some of the valves served by it. 

TP has several engineers dedicated to the air system and the operability of AOVs. However, diesel 

generator AOVs and dampers are under the cognizance of system engineers and technicians assigned to 

care for the particular system. 

A number of reports and training aids were provided. These are described in the tabular listing 

below. The plant air system is described in TP System Description #155, which was provided. 

The instrument air (1A) system is intended to provide clean, dry, oil free air at 100 psig for 

instrumentation and control as well as pneumatic actuators throughout the plant. This is a non-safety

related system. Electrically driven compressors supply each unit and diesel driven compressors are also 

available for each unit. Present plant configuration is one electrically driven compressor (3CM and/or 

4CM) supplying air to both units, with the remaining electrically driven compressor and the diesel driven 

compressors (3CD and 4CD) on standby. Service air is available as a backup in case of a loss of IA. The 

IA header is divided into branch lines that supply the steam dump valve accumulators, turbine area, intake 

structure and water treatment plant, auxiliary building and control room, and the containment and 

blowdown area. Air supplied to the intake structure/water treatment plant and the auxiliary 

building/control room is supplied from both units to preclude losing air to those areas if 1A from a single 

unit is lost. Temporary diesel compressors can also be attached to the IA system. 

The IA system is equipped with filters for the system and for individual AOVs. Continuous digital 

readout dew point monitors are installed which are not alarmed; however, the dryers signal a local alarm 

for high moisture on outlet. (There is no alarm in the control room for high dew point but there is an 

alarm for loss of IA.) Automatic drain traps are installed. The dryers are powered from emergency 

power to ensure that dry air will be provided under accident conditions. Air system samples are taken for 

particulates, dew point, and hydrocarbons/toxins at 20 different locations on the system. The distribution 

piping for the system is galvanized steel with the instrument supply tubing being stainless steel or copper. 

TP had problems with its IA system some years ago related to considerable amounts of water in the 

instrument air system. These problems prompted action by the management to improve the quality of 1A, 

and are considered by the operators to be very successful, based on performance since the modifications 

were made. 

There are 174 AOVs at TP classified as Category 1 (98 active and 76 passive) and 53 AOVs 

classified as Category 2 (34 active and 19 passive). Of the 227 Category 1 and 2 AOVs, 191 are 

classified as safety-related, 34 as quality-related, and 2 as non-nuclear-safety. The total number of AOVs 

at TP is 836. By way of comparison, there are 111 motor-operated valves in the TP Generic Letter 89-10 

program. 

Several functional failures and degraded conditions of AOVs were identified by the engineers. 

These are discussed in Item 7 of the table below. Corrective actions were taken for the degraded 

conditions.

NUREG/CR-6654 C-76
Appendix C

Condition Report 96-1598 indicated that the plant engineers examined the 10 CFR Part 21 

notification from LaSalle regarding mistakes made by the AOV manufacturer in computing pneumatic 

actuator diaphragm areas for WKM Model 70-13 pneumatic actuators. CR 96-1598 applies to BS&B 

actuators at Turkey Point. CR 96-1598 does not mention NRC Information Notice 96-68, which also 

describes this deficiency. The engineers reviewed diaphragm areas and calculations for 28 (including 20 

safety-related) AOVs. Their conclusion was that the valves at Turkey Point that might be affected by the 

Part 21 notification would be capable of performing their intended functions. 

The cooperation, courtesy, and knowledgeable responses from the members of the Turkey Point 

technical staff were noted and appreciated by those of us who are involved in this study of AOVs. 

The following tabular summary was prepared to describe the information gathered during the 

subject visit. The NRC, with assistance from the INEEL, is studying the performance of air-operated 

valves (AOVs) in commercial nuclear power plants. The information was collected in accordance with 

the Program Plan dated 10/22/97 (INEEL Letter to H. Ornstein, NRC, from J. Bryce, 10/23/97, Job Code 

E8238, Task Order 15 - JHB-167-97).

C-77 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

I Date. March 24 and 25, 1998. 

2 Name of Interviewers. Owen Rothberg, INEEL/LMITCO, 301/816-7773 

Mark Holbrook, INEEL/LMITCO, 208/526-4362 

Hal Ornstein, NRC/AEOD, 301/415-7574 

Jerry Jackson, NRC/RESEARCH, 301/415-6656 

Kahtan N. Jabbour, NRC/NRR/PM, 301/415-1496 (sat in on many of the 

sessions) 

3 Plant Name & Docket No. Turkey Point Nuclear Plant, Units 3 and 4. Docket Nos. 50-250 and 50

251. 

4 Person(s) Interviewed, Olga Hanek, TP Licensing Engineer, 305/246-6607 

Title(s), Phone Number(s), Inman Lanier, CSI Valve Engineer, Juno 

E-Mail address, short 

description of Tim Miller, TP AOV Engineer, 305/246-6620 

organization(s) and duties. Hal McKaig, TP Component Engineering Supervisor, 305/246-6739 

Gary Hollinger, TP Licensing Manager 

Dan Tomaszewski, TP System Engineering Manager, 305/ 

246-6158 

T. V. Abbatiello, TP Quality Manager 

Steve Hellriega, TP Works Control Supervisor 

Tom Carter, TP Maintenance 

Devin Ryan, TP Air Systems Engineer, 305/246-6612 

Tom Jernigan, TP Plant General Manager 

Jalal Zamanali, FPL Supervising Engineer, PRA, 561/694-3857 

E. A. Thompson, TP Engineering Manager 

Scott Turner, St. Lucie Senior Engineer 

J. W. York, NRC Resident Inspector 

Rogerio Reyes, NRC Senior Resident Inspector, 305/245-7669

NUREG/CR-6654 C-78
Appendix C 

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT

ITEM No. I INFORMATION I RESPONSE OR INFORMATION

5 If necessary, and if We arrived with the following materials: 

person(s) interviewed can - LER 25084031, Containment Isolation Valve for the nitrogen 

do so, obtain any missing supply system to the Safety Injection Accumulators did not close. 

information not provided Solenoid valve failed. 

prior to the site visit, as 

described in the outline for - LER 25085002, Containment Isolation Valve for Steam Generator 

Task 4 above. Note what Blowdown Isolation would not close because dirt had clogged the 

information was provided. SOVs. 

- LER 25085020, MSIVs might not close because of insufficient 

instrument air combined with low steam flow. 

- 25085021, eedwater Control Valve and Aux. Feed Flow Control 

Valve malfunctioned due to moisture in the instrument air system. 

- LER 25085024, Aux. Feed Pump Discharge Control Valve failed to 

close fully because of a misadjusted positioner. 

- LER 25086005, MSIV partially closed and would not close fully 

because of malfunctioning SOV. 

- LER 25086028, Potential loss of SI pumps due to potential failure 

of AOVs from a loss of instrument air. 

- LER 25086031, Unit Shutdown due to Aux. Feed Flow Control 

Valve malfunction due to SOV failure. 

- LER 25086036, Both Emergency Diesel Generators out of service 

due to governor solenoid out of adjustment. 

- LER 25087002, Containment Isolation Valve was out of service due 

to a failed solenoid coil. 

- LER 25088002, Reactor Coolant System pressure decreased due to 

failure of Control Valve's controller. 

LER 25092005, MSIVs declared inoperable due to misalignment of 

valves during nitrogen bottle replacement. 

- LER 25094001, SOV failed during surveillance, 3rd of this type. 

SOVs replaced and configuration changed. 

- LER 25185014, Containment and control room isolation occurred 

and pressurizer relief tank drained due to leakage of Pressure 

Control Valve. 

- LER 25185021, Turbine runback occurred due to erratic behavior of 

Main Feedwater Flow Control Valve. 

- LER 25186018, Auxiliary Feedwater declared out of service due to 

Aux Feed Control Valve Failure to open. 

(continued next page)

C-79 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT

ITEM No. I INFORMATION [ RESPONSE OR INFORMATION

5 (cont.) Information obtained - LER 25186025, Reactor Trip due to Feedwater FCV failing closed 

(continued) due to a failed solenoid coil. 

- LER 25188009, Failure of Differential Pressure Switch on 

Condensate Polishing Vessel Resulted in SG Feed Pump Trip and 

Aux. Feed Initiation. 

- LER 25191006, Autostart of Aux. Feed Pumps Following a low 

suction pressure trip of MF pump due to malfunction of Condensate 

Polishing Vessel inlet Valve close limit switch. 

- LER 25192007, Auto Aux. Feed start on MFP trip due to 

malfunction of limit switch on AOV. 

- USNRC FACILITY STATISTICS AND GENERAL 

INFORMATION, taken from the NRC Bulletin Board on the 

INTERNET. 

We were provided with the following materials by the site engineers: 

- Slides describing the Turkey Point AOV Program, including 

diagrams of the instrument air system and instrument air dryers. 

- A list, generated from the licensee's database of safety-related 

AOVs at Turkey Point showing descriptive information and 

categories. 

- TP Student Handout 1710800, Pneumatic Measurement and Control 

Applications. 

- TP Student Lab Exercise Guide 1708805, Troubleshoot a Pneumatic 

Pressure Measurement Channel. 

- TP Student Lab Exercise Guide 1708808, Calibrate a Pneumatic 

Level Measurement Channel. 

- TP Student Lab Exercise Guide 1708809, Troubleshoot a Pneumatic 

Level Measurement Channel. 

- TP Student Lab Exercise Guide 1708810, Align a Pneumatic 

Controller. 

TP Student Lab Exercise Guide 1708812, Align a Control Valve 

and Actuator. 

- TP Student Lab Exercise Guide 1708829, Calibrate a Positioner 

(Fisher 3582/3583). 

TP Student Lab Exercise Guide 1708524, Pressure Process Lab 

Orientation. 

(continued next page)

NUREG/CR-6654 C-80
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT

ITEM No. [ INFORMATION RESPONSE OR INFORMATION

5 (cont.) Information obtained - TP Student Handout 1710600, Process Control Fundamentals. 

(continued) - TP Student Lab Exercise Guide 1708601, Determine the 

Characteristics of a Temperature Process. 

- TP Student Lab Exercise Guide 1708602, Determine the 

Characteristics of a Pressure Process. 

- TP Student Lab Exercise Guide 1708603, Determine the 

Characteristics of a Flow Process. 

- TP Student Lab Exercise Guide 1708604, Determine the 

Characteristics of a Level Process. 

- TP Student Lab Exercise Guide 1708606, Determine the 

Characteristics of Controlling a Level Process With a Two-Position 

Controller. 

- TP Student Lab Exercise Guide 1708609, Analyze the Effects of 

Changes on a Proportional Controller for a Flow Process. 

- TP Student Lab Exercise Guide 1708613, Analyze the Effects of 

Changes on a Proportional Plus Integral Controller For a Level 

Process. 

- TP Student Lab Exercise Guide 1708615, Analyze the Effects of 

Changes on a Proportional Plus Integral Plus Derivative Controller 

for a Pressure Process. 

- TP Student Lab Exercise Guide 1708617, Bench Check a 

Controller. 

- A chart and card entitled "Turkey Point System Importance," 

(showing instrument air ranked 7 of 17 systems in risk significance 

and showing IA as about 0.4 of the highest risk significance, by 

scale). 

- Plant Event Report for all plants, dated 3/23/98. 

- FP 0-OSP-200.1, Schedule of Plant Checks and Surveillances. 

- FP 4-OSP-072.4, MSIV Air Accumulator Check Valve Test. 

- FP 3-OSP-072.3, MSIV N-sub-2 Backup System Consumption 

Test. 

- FFP 3-OSP-072.2, MSIV N-sub-2 Backup Periodic Test. 

- FP 3-OSP-072, MSIV Closure Test. 

- lP 4-OSP-072, MSIV Closure Test. 

- FP 3-OSP-075.3, AFW Nitrogen Backup System Low Pressure 

Alarm Setpoint Verification. 

- FP 3-OSP-075.6, Auxiliary Feedwater Train I Backup Nitrogen 

Test. 

(continued next page)

C-81I NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT

ITEM No. INFORMATION T RESPONSE OR INFORMATION

Information obtained 

(continued)

5 (cont.)

NUREG/CR-6654

- FP 3-OSP-075.7, Auxiliary Feedwater Train 2 Backup Nitrogen 

Test. 

- FP System Description No. 155, Plant Air Systems. 

- FP O-ADM-518, Condition Reports (Preparation Instruction). 

- FP 3-OSP-041.4, Overpressure Mitigating System Nitrogen Backup 

Leak and Functional Test. 

- FP 0-ADM-059, Root Cause Analysis. 

- FP 3-OP-041.4, Overpressure Mitigating System. 

- FP 3-OSP-075.10, AFW Flow Control Valve Test. 

- FP O-OSP-200.5, Miscellaneous Tests, Checks and Operating 

Evolutions. 

- FP 4-OSP-200.3, Secondary Plant Periodic Tests. 

- FP 0-PPM-041.4, RCS Pressurizer Main Spray Valves Overhaul. 

- Memo from Jalal Zamanali, FPL to Tim Miller (ENG-NR-98-023) 

Risk Ranking of the AOVs for Turkey Point Nuclear Plant, dated 

March 3, 1998. 

- TP3 Drawing 5613-M-3030, Component Cooling Water. 

- TP3 Drawing 5614-M-3041, RCS PORV Control. 

- TP3 Drawing 5613-M-3075, Aux. Feed System Nitrogen Supplies 

to AFW Control Valves. 

- TP3 Drawing 5614-M-3072, Main Steam System MSIV Control. 

- TP3 Drawing 5613-M-3013, Instrument Air System Air 

Compressors. 

- Summary of Condition Reports Involving AOVs, Tracking and 

History. 

- Condition Report 96-0304, 3A TPCW HX Inlet Isolation Valve 

(POV-3-4882) failed to fully close. 

- Condition Report 96-0535, Pilot Lockup Valves stuck closed. 

- Condition Report 96-1202, RCS letdown isolation valve failed to 

close. 

- Condition Report 96-1598, Investigation of Part 21 notification on 

vendor mistakes in computing AOV disk areas. 

- Condition Report 97-0754, AFW FCV failed stroke-timing test. 

- Condition Report 98-141, Auxiliary Spray to Pressurizer Spray 

isolation valve SOV failure. 

- Two Reports entitled "Instrument Air Compressor Upgrade" dated 

6/92 and 12/93. 

(continued next page)

C-82
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT

ITEM No. INFORMATION RESPONSE OR INFORMATION 

6 Describe plant events Feedwater bypass failures in 1985 due to moisture in the Instrument Air 

involving AOVs and system; valve failures caused a S/G water level transient; moisture also 

provide reference caused AFW valves to fail during the same time period. These multiple, 

information, if possible. simultaneous AOV failures affected all of the three trains of AFW on site. 

This was a significant common-cause failure event. It was found to be an 

Recent: important accident sequence precursor (ASP) event (see Appendix C) and 

Recurring: also caused steam generator water level transients. 

Significant:

7 Describe AOV or air Five AOV functional failures have been identified since January 1994, as 

system actual or detected follows: 

potential failures at the 1. Pilot Operated Lockup Valves (POLVs) for Emergency 

plant? Provide reference Containment Cooler (ECC) Valves. 

information, if possible. The function of the POLVs is to open the ECC outlet valves when air 

pressure drops to about 45 psig, Unit 4, or 60 psig, Unit 3. The ECC 

outlet valves are moved to the open position on loss of IA because the 

actuators require air pressure to function and the IA system is not 

safety-related. Failure of the POLVs to shift on loss of IA pressure 

could result in insufficient CCW flow to support the containment 

temperature/pressure control design basis safety function. Therefore, 

the POLVs are considered safety-related. There are 12 POLVs (6 per 

unit) at Turkey Point. One failure of a POLV to shift was confirmed in 

1996. As-found examination of the other POLVs indicated that 3 of 

the 12 were stuck. Several contributors to failure, included O-ring 

distortion and grease caking, were identified as causing excessive drag. 

These are considered to be common-cause, common-mode failures. 

Increased exercise and spring modifications were implemented as fixes 

and the failures had not recurred. (Condition Report 96-0535 applies.) 

2. Intake cooling water isolation valve failures. 

POV-3-4882 and POV-4-4883 (safety-related AOVs) failed to fully 

close in two separate incidents. The problems were considered not 

reportable under 10 CFR 50.72 or 50.73. Minor wear was found in the 

housing and cover guides. Corrosion was found on the lower portion 

of the yoke and housing below the 0-ring seal. The corrosion was 

identified as the root cause of failure. (Condition Reports 96-0304 and 

96-0735 apply.) POV-4882 is a risk significant, key component. 

These are considered to be common-cause failures. The PM frequency 

and monitoring were increased and the licensee indicated these 

measures had been effective.

_____ I _________ __ I

C-83 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

7 (cont.) (continued) 3. RCS letdown isolation valve failure. 

Describe AOV or air CV-3-204 (safety-related AOV) failed to close when remotely 

system actual or detected operated from the control room. The SOV pilot was determined to 

potential failures at the be defective and was replaced; however, no definitive root cause 

plant? Provide reference was identified. A small amount of Teflon tape was found in the 

information, if possible body of the SOV. Condition Report 96-1202 applies. 

4. AFW control valve exceeded allowable stroke time. 

CV-3-2832 (safety-related AOV) was stroked a number of times 

and the stroke times were found to be erratic and slow. The most 

likely problem was wear and debris generated from that wear in the 

positioner. Condition Report 97-0754 applies. 

5. Potential failure of auxiliary spray to pressurizer spray line. 

CV-3-3 11 (safety-related AOV) was reported to have a solenoid 

pilot blowing air. The SOV was replaced but no definite root cause 

was identified. 

Degraded AOV Conditions: 

1. Copes Vulcan AOVs (PORVs) had diaphragm problems due to a 

high ambient temperature environment. This was termed 

"diaphragm creep." A short time after installing a new diaphragm, 

it leaked and the hold-down bolts were found to be loose. A new 

silicon rubber diaphragm was installed to correct the problem. This 

is a common-cause failure mechanism. Condition Report 94-1192 

applies. 

2. The pressurizer spray valves were moved to a cooler environment in 

the 1980s in order to improve operation. No reference was found. 

8 Describe actions taken TP added auto drain traps to several low points in the instrument air 

after events or failures system. The licensee performs air system samples at 20 different locations 

involving AOVs or the air for particulates, dewpoint, and hydrocarbons/toxins. The air dryers are 

system. Provide reference powered from emergency power to provide assurance that dry air will be 

information, if possible. available during accident conditions. FPL added continuous on-line dew 

point monitoring as well as alarms on the outlets of the dryers. FPL also 

installed new oil-free rotary screw air compressors which are much more 

efficient and require less maintenance, and added 2 diesel-driven air 

compressors (all with auto-start capability) to increase air system 

reliability. 

9 Were there any actual or Turkey Point has several Copes-Vulcan AOVs; these valves may 

potential common mode or potentially be affected by the diaphragm creep problem recently identified 

common cause failures in at Dresden. Refer to Dresden LER 23798003 dated 2/27/98, Dresden 

the air system or AOVs at 50.72 Report 33620 dated 1/28/98, and Dresden MR No. H-98-0045 dated 

the plant? Describe and 3/6/98. 

provide reference 

information, if possible.

NUREG/CR-6654 C-84
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT

ITEM No. INFORMATION [ RESPONSE OR INFORMATION 

10 Describe root cause Refer to FP&L O-ADM-059, Root Cause Analysis. This Procedure 

analysis procedures for the describes a formal process for determining root causes of failures. 

plant. Provide reference 

information, if possible. 

11 Describe root-cause A formal root cause analysis program is in place and is described in FP&L 

analyses performed for air 0-ADM-059. An examination of the LERs for the plant indicates that root 

system or AOV failures at causes for events are investigated and reported. 

the plant. Provide 

reference information, if 

possible. 

12 Describe maintenance Periodic maintenance is performed on the air compressors (quarterly and 

procedures for the air semi-annually) and controls (annually, auto-start pressure switches semi

system. Provide reference annually); air dryer (desiccant) heaters and controls (annually); after-filter 

information, if possible. (annually), and point of use filters are routinely changed out. Dryer 

performance testing is done quarterly. Air quality sampling is performed 

every 18 months. Water traps are blown down and the dew point is 

checked each shift. 

13 Describe maintenance TP did maintenance in the past based on experience; are moving toward a 

procedures for AOVs. more standardized set of activities; depends on environment; will do 

Provide reference routine actuator overhauls in some cases. 

information, if possible. 

Safety-related: 

Important non-safety

related: 

Non-safety-related: 

14 Describe IST procedures No formal IST program is used for the IA system, although a number of 

for the air system. system and equipment checks are performed on a routine basis. The IA 

Provide reference system is considered to be a non-safety-related system. 

information, if possible. 

15 Describe IST procedures Stroke time testing is performed in accordance with ASME Section XI 

for AOVs. Provide code requirements. 

reference information, if 

possible. 

Safety-related: 

Important non-safety

related: 

Non-safety-related: 

16 Describe diagnostic Turkey Poinit uses the Fisher FlowScan system. This system has been 

systems, if any, used for used in the past as a troubleshooting tool. The licensee also recently 

AOVs. Provide reference purchased the Crane.MOVATS Universal Diagnostic System (UDS) to 

information, if possible. provide direct force measurements. This system uses Teledyne strain 

Description of system: gauges.

C-85 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

Specifications: Crane MOVATS recently provided UDS training to Turkey Point 

Data collected and personnel. Turkey Point engineers plan to use the UDS as a monitoring 

frequency of collection: tool for AOVs. 

Vendor assistance 

provided, if any: 

17 Describe design (and As of the time of the site visit, TP is using the original air settings that 

analysis) procedures for were provided when the plant was constructed. The licensee believes that 

AOVs. Describe how this approach is adequate based on experience (i.e., few design-related 

design basis is established problems have been observed). FPL engineers will develop design-basis 

and maintained for AOVs. requirements for a sample of their Category 1 AOVs to see if there are any 

Provide reference problems. FPL will also look at specific valves if a design-related failure 

information, if possible. occurs. However, the licensee does not intend (at this time) to do a 

design-basis review for all program AOVs. For example, none of the 

Category 2 valves will be reviewed unless there are problems identified as 

part of the Category 1 AOV review. FPL engineers indicated that they 

also plan to review any future JOG recommendations to develop design

basis requirements as part of any industry-wide AOV program. 

This approach assumes that the fluid conditions experienced by an AOV 

during normal operations are in fact the worst-case conditions that a given 

valve will ever have to operate under. This assumption cannot be verified 

unless a design-basis review is performed. This approach also assumes 

that the valve vendors' sizing methodologies are conservative. Industry 

MOV experience has shown that, in many cases, the valve vendors have a 

poor understanding of how the valves will perform under design-basis 

conditions. 

18 Describe analyses and/or No specific test program is in place for verifying design-basis operability 

testing for verification of of AOVs, other than diagnostic testing under static conditions and 

operability during comparisons to existing vendor requirements. See Item 17. 

postulated transient or 

accident conditions. 

Provide reference 

information, if possible. 

19 Describe training for Training for use of the Fisher FlowScan diagnostic system is addressed by 

installation, maintenance, on-the-job training in the plant. Three days of training are set aside to 

and testing of AOVs. cover use of the new MOVATS UDS system. 

Provide reference New I&C personnel receive eight weeks of Lab training on in-plant 

information, if possible. equipment. AOVs are covered during part of this lab period. Training 

includes disassembly, reassembly, and actuator setup of the more common 

AOV actuators found in the plant. Industry events and emerging technical 

issues are addressed during Annual Continuing Training sessions, which 

last two weeks.

NUREG/CR-6654 C-86
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

20 Describe databases used to TP provided several runs from their database of AOVs and AOV events 

track maintenance, for the site. They track data from their own plant and from the industry. 

failures, and events 

regarding AOVs. Provide 

reference information if 

possible. 

On site: 

Company wide: 

Industry: 

21 Describe the impact of the The licensee has reviewed the plant systems to determine those functions 

Maintenance Rule, 10 that are risk significant. Using these functions, the licensee reviewed the 

CFR 50.65 on AOV and AOV population to identify those valves that are needed to support the 

air system maintenance risk-significant functions. These valves were identified as Category 1 

and testing. Provide AOVs. If a valve was safety related, but not identified as Category 1, it 

reference information if was classified as Category 2. Other AOVs that were not Category 1, but 

possible. were considered important to plant operation were also classified as 

Category 2. 

22 Is PRA data used for Predictive maintenance of AOVs was not discussed. 

predictive maintenance or 

replacement of AOVs? If 

so, how? 

23 Are AOVs serviced on AOVs are typically serviced on site by licensee maintenance personnel. 

site, serviced offsite, or Some outside personnel are added during outages to complete the 

replaced as piece-parts if increased work scope. These additional personnel receive training and are 

found to require service? supervised by licensee maintenance personnel. 

24 Identify and describe the TP noted some environmental effects on the PORVs (i.e., "diaphragm 

most common recurring creep") which was caused by high environmental temperature conditions. 

maintenance problem(s) 

and failures regarding 

AOVs and the air system. 

What did you see? 

Provide reference 

information if possible. 

25 Interviewer comments We noticed several "green tags" on valves and pressure switches denoting 

regarding actual valves maintenance outstanding issues. Some of these tags dated back to 1997. 

viewed during the visit, in Time did not allow further investigation to determine how effective this 

the plant, undergoing tag system is for scheduling needed work activities. 

maintenance or 

replacement, or in the 

plant stock system, if 

applicable to this 

interview.

C-87 NUREG/CR-6654
Appendix C

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT

ITEM No. INFORMATION [ RESPONSE OR INFORMATION 

26 Has the plant made The Feedwater regulating valves were modified to put a Fisher actuator on 

changes to valves or a Copes-Vulcan valve when the original actuator proved unreliable. 

systems that include 

AOVs, or replaced AOVs 

with different models of 

AOVs or different valves 

that are not AOVs? If so, 

describe the changes and 

the circumstances. 

What prompted the 

change? 

Was the change made for 

this plant only? 

27 Does the plant follow The plant personnel are aware of the industry efforts toward AOVs. They 

EPRI/NMAC guidelines have consulted the industry guidelines as reference material. 

for maintaining AOVs and 

the air system(s)? If not, 

describe differences and 

reasons for the differences. 

Provide reference 

information, if possible. 

28 What is the plant doing or At the time of the visit, we were told that TP had no plans to change their 

planning to do in response existing approach to evaluation or maintenance of AOVs. Since that time, 

to the recent Industry we have been told that TP is putting together an AOV program. 

correspondence on AOVs. 

Provide reference 

information, if possible. 

29 Do you have any TP relies heavily on personnel training to ensure proper maintenance and 

suggestions for improving adjustment of AOVs, and this was emphasized. As noted in item 5 in this 

the performance of AOVs, table, TP has specific procedures in place to control maintenance on 

particularly in the areas of safety-related and important non-safety-related AOVs. 

surveillance, testing, or 

maintenance? 

30 Provide a list of 10 CFR A list was requested and was provided near the end of the plant visit. See 

50.59 and 10 CFR 50.72 item 7 in this table. 

reports on AOVs and 

AOV support systems (air 

or inert gas supply, etc.) 

that have been issued for 

this plant.

NUREG/CR-6654

C-88
Appendix C

NUREG/CR-6654

TOPICS TO REVIEW FOR AIR OPERATED VALVE STUDY 

SITE VISIT TO THE TURKEY POINT NUCLEAR PLANT 

ITEM No. INFORMATION RESPONSE OR INFORMATION 

31 What thrust or torque Turkey Point has not established any formal margin guidance for AOVs. 

margins are expected for Licensee personnel indicated that an informal target of 20% margin is 

AOVs? Are different desired. However, this margin is not known to exist for all AOVs in the 

margins used for safety- program because calculations have (as of the time of the visit) not been 

related, important non- completed. 

safety-related, or non

safety-related AOVs? 

32 What maintenance or Turkey Point has several AOVs that have nitrogen or air accumulators 

surveillance is done to (e.g., MSIVs, AFW regulating valves, steam dump valves, and PORVs). 

AOV accumulators to These accumulators are tested via consumption tests and intersystem (i.e., 

ensure air/nitrogen quality nitrogen/IA) check valve isolation tests. 

and pressure? Were 

seismic considerations and 

size verified? 

33 Describe problems with In the discussions, it was noted that settings for pressure regulators and 

pressure regulators, if any. controllers need to be formally tracked and monitored in order to be 

assured that settings are proper. An extensive set of training materials 

was provided related to troubleshooting and adjustment of AOVs and their 

control hardware. Settings are tracked and controlled by plant internal 

procedures. 

34 Describe problems with A large amount of water in the instrument air system in 1985 caused 

Feedwater regulating several Feedwater and Aux. Feedwater problems as identified in the LERs. 

valves, if any. These events led to improvements in the supply and quality of IA at TP. 

35 What, if any, is your FPL personnel routinely participate in the AOV Users' Group meetings. 

involvement with the AOV They were also to have participated in the summer of 1998 ASME/NRC 

Users Group? Describe. pump and valve symposium.

C-89
NRC FORM 338 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vol., Supp., Rev., 

NRCM 1102, and Add.lum Numbers, fBUanH.) 3•01,3M2 BIB012GRAPHIC DATA SHEET 

(See ,nsbuolions on iere) NUREG/CR-6654 

2. TITLE AND SUBTITLE INEEL/EXT-98-00383 

A Study of Air-Operated Valves in U.S. Nuclear Power Plants 3. DATE REPORT PUBLISHED 

MONTH YEAR 

February 2000 

4. FIN OR GRANT NUMBER 

E8238 

5. AUTHOR(S) 6. TYPE OF REPORT 

0. Rothberg, S. Khencha, J. Watkins, M. Holbrook Technical 

7. PERIOD COVERED (inclusive Dtes) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC provide Division, Ofie or Region, U.S. NulerRAgulia Commision, and rnih'ng address; if con-'cd, 

provide name ,and met ades.) 

Idaho National Engineering and Environmental Laboratory 

Bechtel BWXT Idaho, LLC 

Idaho Falls, ID 83415-3129

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if NRC, type "Sarne as above, iffconkkctr provide NRC Division, Offce or Region, U.S Nuclwer Regulatory Commission, 

&nd maiing addins.)

Division of Systems Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

10. SUPPLEMENTARY NOTES 

H. Omstein, NRC Project Manager 

11. ABSTRACT (200 words or low) 

A study of air-operated valves in nuclear power plant applications was conducted for the NRC Office of Research (the project was 

initiated by NRC/AEOD). The results of the study were based on visits to seven nuclear power plant sites, literature studies, and 

examination of event records in databases available to the NRC. The purpose is to provide information to the NRC staff 

concerning capabilities and performance of air-operated valves (AOVs). Descriptions of air systems and AOVs were studied along 

with the support systems and equipment. System and equipment that contain AOVs and solenoid-operated valves (SOVs) were 

studied to determine their dependencies. Applications of AOVs and SOVs were listed along with current NRC requirements.

12. KEY WORDSDESCRIPTORS (/.st words orpkses OhWt wil earstresew-chers ino eooti e npat) IS. AVAILABILrTY STATEMENT 

air-operated valves; solenoid-operated valves; licensees; design basis loads; common-cause failures; unlimited 

safety systems; accident sequence precursor, probabilistic risk assessment; system reliability 14. SECURITY CLASSIFiCATION 

(This PaOe) 

unclassified 

(Thi Report) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (24%0