or NUREG/CR-5406

EGG-2569

Vol. 3

BWR Reactor Water Cleanup System

Flexible Wedge Gate Isolation

Valve Qualification and High

Energy Flow Interruption Test

Review of Issues Associated with

BWR Containment Isolation Valve Closure

Prepared by K.G. DeWall, R. Steele, Jr.

Idaho National Engineering Laboratory

EG&G Idaho, Inc.

Prepared for

U.S. Nuclear Regulatory Commission
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NUREG/CR-5406

EGG-2569

Vol. 3

R1

BWR Reactor Water Cleanup System

Flexible Wedge Gate Isolation

Valve Qualification and High

Energy Flow Interruption Test

Review of Issues Associated with

BWR Containment Isolation Valve Closure

Manuscript Completed: September 1989

Date Published: October 1989

Prepared by

KG. DeWall, R. Steele, Jr.

Idaho National Engineering Laboratory

Managed by the U.S. Department of Energy

EG&G Idaho, Inc.

P. O. Box 1625

Idaho Falls, ID 83415

Prepared for

Division of Engineering

Division of Safety Issue Resolution

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

NRC FIN A6857, B5529
ABSTRACT

This report presents the results of research performed to develop technical insights for the NRC effort regarding

Generic Issue 87, 'Tailure of HPCI Steam Line Without Isolation." Volume m of this report contains the data and

findings from the orgiial researchperformedto assess the qualification ofthe valves and reportedinEGG-SSRE-7387,

"Qualification of Valve Assemblies in High Energy BWR Systems Penetrating Containment" We present the original

work here to complete the documentation wail. The recommendations contained in Volmne M of this report resulted in

thetestprogamdescribedinVohmleI andIl. Theresearchbeganwithasurveytocharacterizthepopuaonofnormally

open containment isolation valves in those process lines that connect to the primary system and penetrate containment.

The qualification methodology used by the various manufacturers identifiedinthe survey is reviewed and deficiencies in

that methodology are identified. Recommendations for expanding the qualification of valve assemblies for hig energy

pipe break conditions are presented.

A6322-Environmental Qualification of Mechanical and Dynamic Qualification of Mechanical and Electrical

Equipment Program

iii
EXECUTIVE SUMMARY

Volume m of this report discusses research performed

to develop technical insights for the NRC effort regarding

Generic Issue 87, "Failure of HPCI Steam Line Without

Isolation." The work was performed under FINA6322.

The Office of NuclearRegulatory Research sponsors the

Mechanical Equipment Qualification Research Program

(FIN A6322) and is assisting in the resolution of this

issue.

Four BWR systems, the Emergency Cooling System,

the Hfigh Pressure Coolant Injection System, the Reactor

Core Isolation Cooling System, and the Reactor Water

Cleanup System, were included in the valve assembly

aracterization. The "typical" containment isolation

valveisa3 to lOin.,600to900lb, gatevalve. Themost

common design is a cast steel, flexible wedge,

pressure-sea valve with a Limitorque operator (AC

inside and DC outside of containment). The

Anchor/Darling Valve Company manufactures

approximately 40% of the valves identified.

The mitigation ofahigh energy pipe break iswitbin the

design basis for the above valve assemblies, with typical

systemdesignconditionsof 1250psi andS750F. Noflow

testing has been performed under these conditions to

verify the presumptions used by manufacturers in the

qualification analysis calculations. Operator torque

switch settings are determined using calculations

supplied by the valve vendor, which could lead to

inadequate torque settings to close the valve if the

original calculations are not conservative.

Most of the valve and operator manufacturers use the

same equation to size operators with minor variations in

coefficients. In this equation, the required thrust to close

the valve is equal to the sum of the disc drag load due to

differential pressure, the stem end pressure load, and the

packing drag load. The service conditions used in the

thrust equation are supplied by each individual plant.

Four areas have been identified as having the most

influence on stem thrust requirements. Observations

concerning these four areas are noted below.

1. Repeated cycling can have a significant

effect on valve thrust requirements.

2. The typical industry 0.3 disc friction

coefficient is not conservative for all cases.

3. Mass flow/momentum influence on valve

thrust requirements may be significant.

4. Increased temperature causes a significant

increase in valve closure loads

The limited number of tests performed to assess gate

valve flow interruption capability with high pressure

steam have resulted in a relatively frequent inability to

isolate. The data now available suggest that industry may

be using nonconservative friction factors and possibly

under-estimating valve stem thrust requirements.

Additional work is needed to determine whether present

qualification practices are adequate.

V
CONTENTS

ABSITACr ....................................................................... . iii

EXEIc.TFVESUM ARY .............................................................. 'V

ACMKOWLEDGM[EMNrS . .............................................................. ix

1. INTRODUCTION . ................................................................ 1

1.1 Background ................................................................. 1

2. OBJECTIVES ..................................................................... 3

3. GI-87 VALVE ASSEMBLIES ................................... S

4. PRESENT VALVE ASSEMBLY QUALIFICATION ...................................... 7

4.1 Valve OperatingDesign Basis .............. ..................................... 7

4.2 Utility Qualification Prograis ................................................... 7

4.3 Vendor Qualification Methoidology ............................................... 10

5. WEAKNESSES IN VENDOR MEHODOLOGY ....................................... 13

5.1 EPR/M~ashrall Test Program Results ............................................. 13

5.2 Westinghouse lest Program Results .............................................. 15

53 Ontalio Hydro How Test ....................................................... 15

6. CONCLUSIONS .................................................................. 17

7. RECONMMENDA1ONS ................................... 19

8. REFERENCES . .................................................................. 21

APPENDIX A-MDENTJFICA7lON OF GI-87 VALVE ASSEMBLIES ....... .................... A-I

APPENDIXB-PSARJPSARDATA . . . B-1

FIGURES

1. lypical motor operated gate valve ..................................................... 8

A-1. Typical Emergency Cooling System ................................................... A-8

A-2. Typical High Pressue Coolant Injection (HPIC) system ................................... A-9

A-3. Typical Reactor Coe Isolation Cooling (RCIC) system ........ ............................. A-10

Vii
A-4. Typical Reactor Water Ceanup (RWCU) system .................................... A-11

A-5. HPCI isolaion valve manufacturers bysie ................. ................... A-12

A-6. RCIC isolation valve manufactuxers bysz s.................... ............... A-12

TABLES

1. Flow inptruptoncapablity test .................................... 9

2. Valve/operatormatrix .................................... 14

A-l. Systems for BWR-2 .................................... A-3

A-2. Systems for BWRs-3 and -4 .................................... A-4

A-3. Systems for BWRs-S and -6 .................................... A-5

A-4. BWR plant listing ..................................... A-6

viii
ACKNOWLEDGMENTS

Mr authorwisfes toexp=isssnereygmeidontoft followingpemm fortdrasemame isresearcheffmt

Robert Bergman (Anchor/Darling Company), Paul Lawson (Crane Company), Robert Kestler(Wm. Powell Company),

Bik Neilson (Velan Incorporated), John Fernell (Velan Incorporated), and Cliff Johnson (Walwomth Company) all

contributed by providing information about their respecive valve qualification methodology.

ix
REVIEW OF ISSUES ASSOCIATED WITH BWR CONTAINMENT

ISOLATION VALVE CLOSURE

1. INTRODUCTION -

The United States Nuclear Regulatory Commission

(USNRC) has assigned a HIGT' priority to Generic

Issue No. 87, 'Failure of HPCI Steam Line Without

Isolation."I Tle issueconcernsapostuatedbreakinthe

High Pressure Coolant Injection (HPCI) steam supply

line in Boiling Water Reacts (BWRs) and the

uncertainty regarding the capability of the HPCI steam

supply line isolation valves to close under those

conditions. A similar situation can occur in the Reactor

Core Isolation Cooling (RCIC) System and the Reactor

Water Cleanup (RWCU) System, along with other high

energy steam lines coming off of the Main Steam Line

(MSL). Without isolation, such breaks have high

potential consequences became other emergency

equipment located in the vicinity of the break would be

exposed to an environment which could result in

common-cause failure. To resolve this issue one must

answer two questions: (a) have the subject valves been

qualified for the conditions expected to result from a high

energy pipe break and (b) were the methods used to

qualify the valve assemblies adequate to assure

operability under pipe break conditions.

The Office of Nuclear Reactor Regulation (NRR),

Division ofSafety Review and Oversight is coordinating

the actions necessary to resolve this licensing issue and

has requested assistance from the Office of Nuclear

Regulatory Research (RES), within ongoing work on the

Mechanical Equipment Qualification Research Program

(FINA6322). The Idaho National Engineering

Laboratory (INEL) is the contractor for this program.

1.1 Background

The PC[ steam supply line typically has two

containment isolation valves in series, usually one inside

containment and one on the outside of containment

These valves are normally openinmost plants. The HPCI

supply valve, located adjacent to the turbine, and the

turbine stop valve are normally closed. The RCIC and

RWCU each have two isolation valves which are

normally open. The RWCU valves must remain open if

the system is to operate.

The gate valve is designed for use in a system where a

positive shut-offis required with minimal pressure drop.

It is ideally suited to those situations where isolation of

one part of a system from another is required and control

of the dynamic properties of the fluid (throttling) is

unnecessary. Withthedisc(orgate)inthe raisedposition,

the run of the valve is free of any obstruction with

approximately the same head loss as in the adjacent

piping. When the disc is lowered into the seat, the

upstream pressure forcesit against the seat creating aseal

and isolating the downstream system from the fluid. The

thrust required to close and open the valve is not

dependent on flow direction; however, the thrust

requirement may be affected by the mass flow through

the valve.

Failure to close, defined as the inability of the valve

operatorto move the gate from the full open to full closed

position in the specified time duration, can result from

many causes.2 Under 01-87 concerns, the two most

important reasons for valves failing to close are

1. Excess stem seal loads

2. Large pressure or flow induced forces.

The first of these, excess stem seal loads, most often

result from pressing the stem packing too tightly against

the stem by overtightening the packing compression

bolts. Tis condition may develop during packing

maintenance either inadvertently or in an attempt to

overcome leaks due to stem scoring. Functional testing

after maintenance is typically performed to guard against

over tightening of the packing.

Largepressure orflow induced forcescanoccurwhena

valve must close to shut offflow from a downstream pipe

break, precisely the concern of GI-87. Under these

conditions, the flow through the valve can reach critical

velocity as the valve closes. The result is large

I
differential pressure and inertiaJmomentum load on the

disc forcing the disc against the seat and increasing

fiction.

Due to flow limitations at the valve manufacturers'

facilities, only the opening chaacteristics ofthe valve are

typicallytestedunderoperatingconditions. Aspartofthe

utilities' In-Service Testing ProgrAm, the operation of

the valves is tested periodically but without steam flow.

The capability of the valves to close when exposed to the

forces resulting from a break downstream has typically

not been fully tested.

2
2. OBJECTIVES

The overall purpose of the INEL research is to provide

a technical basis for the resolution of G187. The

following research objectives were developed to guide

the research toward this end.

1. Identify (by manufacturer and model) the

specific valve assemblies used in the BWR

systems which fail under the concerns of

01-87. (Documented in "Summary of Valve

Assemblies in High Energy BWR Systems

Outside of Containment-Interim Report.")

2. Determine the conditions for which those

valve assemblies have been qualified and

identify valve assemblies that have adequate

qualification to assure isolation of a high

energy line break.

3. Review the qualification methods used by

vendors and identify deficiencies in that

methodology.

4. Recommend appropriate follow-up effons

required to assure adequate qualification of

questionable valve assemblies.

3
3. G1-87 VALVE ASSEMBLIES

A review of available information sources to identify

the systems applicable to GI-87 was performed and a

determination was made as to the valve and operator

manufacturers, types, and sizes used in those systems.

The following is a summary of the results presented in

AppendixA. The BWR systems containing isolation

valves of concern under GI-87 are the following.

1. Emergency Cooling System (steam leaving

the reactor-BWR-2 only)

2. High Pressure Injection System (HPCI)

(turbine steam supply-BWR-3 and 4 only)

3. Reactor Core Isolation Cooling (RCIC)

(turbine steam supply)

4. Reactor Water Cleanup (RWCU).

Preliminary and Final Safety Analysis Reports

(PSAR/FSAR) and data from the Institute of Nuclear

PowerOperations (NPO) NuclearPlant Reliability Data

System (NPRDS) were used to determine specific valve

assembly information. With only a few exceptions, the

valves thatmustbe qualifiedinorder to resolve GI-87 are

pressuro-seal, caststeel, flexible wedge gate valvesindte

3 to 10 in. range and 600 and 900 lb. class.

The most predominant valve manufacturer is the

Anchor/Darling Valve Co. with 41% of the containment

isolation valves. The other manufacturers are

Borg-Warner (2%), Crane Co. (18%), William Powell

Co. (11%), Velan Inc. (16%), and Walworth Co. (12%).

Limitorque Corporation manufactures 94% of the valve

operators. The remaining 6% are identified as

Philadelphia Gear Corporation operators (predecessor of

Limitorque Corporation).

5
4. PRESENT VALVE ASSEMBLY QUALIFICATION

The following paragraphs discuss the results of

reshperformed to determine the conditions forwbich

the valve assemblies identified in the previous section

have been qualified and todetermine the methods used by

utilities and vendors to provide this qualification. An

essential part of this discussion will be a review of

operator sizing and torque switch settingpractices, since

these items directly control valve disc movement.

4.1 Valve Operating Design

Basis

The second objectivelistedinSection2ofthisreportis

to determine the conditions for which the valve

assemblies have been qualified and identify the valve

assemblies that have adequate qualification to assure

isolationof a high energy line break. Inorderto complete

this objective, a number of utility submittals in response

to 1E Bulletin 85-03 (Reference 3) were reviewed to

identify maximum valve design differential pressure and

temperature. Although the bulletin addressed valve

torque switch settings exclusively, the valve design

information requested covers the containment isolation

valves of interest to G147 in the steam lines for the HC[

and RC[C systems.

The design basis for each valve consists of (a) the

maximum differential pressure expected durng opening

andlorclosing of the valve for both normal and abnormal

events, and (b) the temperature corresponding to these

conditions. At most plants, the maximum expected

differential pressure is conservatively considered to be

the maximum upstream pressure. No credit is taken for

the downstream pressure. Thus, the maximum expected

differential pressure will be the most conservative

enveloping differential pressure that could be

experienced by the MOVs during various plant

operational modes.

Of the plants responding to lE Bulletin 85-03, most

identified thepipe break condition as a design basis event

for the containment isolation valves in the HPCI and

RCaC steamlines and the RWCU suction line. At the

Perry Nuclear Power Plant for example, the control

switch settings for these valves take into account line

breaks, and are designed to provide positive valve

actuation up to the maximum differential pressures

expected to be seen across the valve in either the open or

close direction during a design basis accident condition.

This envelopes single equipment failure or inadvertent

equipment operation.

The utility submittals in response to E Bulletin 85-03

indicated that the upstream (and thus maximum

differential) pressure ranged from 1100 to 1375 psig and

the corresponding temperatures ranged from 540 to

585°F. The submittals verify the FSAR information

found in the first part of the GI-47 study and lead to the

conclusiond thighenergypipebreak isolationis within

the valve's design basis. 6

4.2 Utility Qualification

Programs

Specific information on the valves identified in the

GI-87 valve survey were obtained from a representative

nuclear power plant. The system design pressures and

temperatures, valve sizes, and valve and operator

manufacturers at the plant are typical of the majority of

operational BWRs.

The valves used in the HPCI, RCIC, and RWCU

systems are manufactured by Anchor/Dadling Co. and

utilize Limitorque operators. They are of the same sizes,

type, and class as those listed in Section 3. These flexible

wedge gate valves consist essentially of a one piece

wedge with the areas behind the seating surfaces

hollowed out to allow more flexibility to conform to the

seat alignment. The bodies of these valves have cast-in

disc guides.

The purchase specifications and requirements include

environmental conditions, thermal transients, and

pressure, temperature, flow and differential pressure

requirements.

The program forselecting correct valve switch settings

consists of the following elements:

1. Calculation of design differential pressures

during the preparation of equipment

specifications.

7
2. Developmentofinitialtorqueswitchsettings

by the valve or motoroperated vendors.

3. Vendor testing of representative valves at

design flows and differential pressures to

verify adequate performance at the

conditions specified in (1), and the switch

settings selected in (2).

4. Stroke testing (with no differential pressure

present) of all valves, using the Motor

Operated Valve Analysis and Test System

(MOVATS) to verify proper torque and limit

switch settings.

The torque switch, limit switch, and stem packing

adjustments are specified by the manufacturer in the

Anchor/Darling Instructions for the Installation,

Operation, and Maintenance Manual.

The following list details the testing that was

performed.

1. Hydrostatic Testing

a Test is performed in accordance with the

Code.

b. Tbe valve must be stroked six times

following the hydrostatic test.

2. Functional Testing

a. Valve is oriented for the most adverse

conditions.

b. The SSE deflection is imposed on the

operator

c. Valve is internally pressurized to the

maximum design pressure.

d. The valve assembly is actuated using the

minimum actuation supply voltage

e. The valve must open and close within the

specified time.

3. Seat Leakage

a. Leakage shall not exceed two cchr per inch

of nominal valve size.

b. The duration of the test shall be at least

four minutes.

The documentation from the representative BWR

plant included a copy ofa data sheet fom avalve closure

test conducted by Wyle Laboratories and a comparison of

the design versus "realistic" valve movement torque

requirements. The data sheet contained information

gained aspant of a Flow Interruption Capability Test and

is reproduced in this report as Table 1. The only

conclusions one can make are the following: (a) the test

began with the system at the design pressure of 1370 psig

and adifferential pressure across the valve of zero, (b)the

valve closed in 2.09 seconds with a final upstream

pressure of 1205 psig and downstream pressure of

390 psig, and (c) the largest differential across the valve

disc during closure was 815 psL This is much less than

the full system pressure one would expect given a GI-8

type pipe break immediately downstream of the valve.

The information given did not indicate the presence of

high fluid flow during the test. In all, the test provided

insufficient information to assure valve operability under

high energy pipe break conditions.

No tests or analysis under blowdown conditions were

performed for these valves by the valve manufat

8
Table 1. Flow interruption capability test

Vaive cosing Time

(sec)

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1.0

1.1

1.2

1.3

1A

1.5

1.6

1.7

1.8

1.9

Upstream Steam

Pressure

Cysg)

1370

1365

1345

1325

1320

1305

1280

1270

1260

1250

1240

1225

1215

1205

1195

1190

1195

1200

1215

1230

Downstream Steam

Pressure

(psig)

1370

1365

1345

1325

1320

1305

1280

1270

1260

1250

1240

1225

1215

1195

1175

1140

1095

1025

910

770

TIst Valve

Differential Pressure

(psid)

0

-0

0

0

0

0

0

0

0

0

0

0

0

10

20

50

100

175

305

460

2.0'

2.1

2.2

1220

1205

1220

560

390

270

660

815

950

a. NOTE: Valve dosed at 2.09 seconds.

9
4.3 Vendor QualifIcation

Methodology

The vendors of the most commonly used components

were contacted andthe utility submittals in response toWE

Bulletin 85-03 were reviewed to better understand

vendor qualification methodology, and to identify

possible flow interruption test data sources. The results

of this review are described below.

A gate valve operator must overcome a force equal to

the differential pressure times a coefficient of friction

(generally 03 for a wedge type gate and 0.2 for aparallel

seat gate). Figure 1 shows a cutaway of a typical motor

operated gate valve.7jA The equation used throughout

most of the valve and operator sizing literature equates

the closing stem thrust to the disc friction load plus the

stem end load plus the packing drag load, as detailed in

Equation (1).

T = ,Fi + F, + Fp (1)

where

T = required stem thrust

pc = Seat coefficient of friction

Fd = Disc differential pressure load

Fs = Stem pressure end load

Fp = Packing drag load.

The exact equation used by each vendor is proprietaiy

asistheseatcoefficientoffriction. One vendorhowever,

uses the following equation instead of Equation (1).

T= [F+F, +FFJAP+Fp (2)

Gate

(Closed position)

Figure 1. Typical motor-operated gate valve.

10
where

T = Required stem thrust

# = Seat coefficient of friction

Famtra = oArea factor

Fa, = Seat factor

AP' =

Stem factor

Differential pressure

F= Packing drag load.

The seat coefficient of friction used for wedge-type

gate valves in Equation (2) is 02. Equation (2) is based

on seating and unseating loads.

Valve vendors place varying emphasis on the

importance of other phenomena in their methods of

determining valve thrust. The majority of the vendors do

not take into account the effects of valve cycling, mass

flow, and temperature. The only testing performed are

the standard tests outlined in the ASME B&PV Code

(pressure and shell tests). They believe that mass flow

through the valve does not produce a significant disc load

andconsideronly differential pressure effects on the gate.

All of the valve operator sizing equations are proprietary

includingthediscfrictionfactor. Mostbelievethatvalve

opening loads exceed and will therefore bound closing

loads.

A minority of the valve vendors take a different stand.

Two vendors have observed instances where the valve

thrust increased with cycling, attributed to temperature

effects on the valve and operator, fluid type, valve design,

and packing design. Their equations include additional

force terms to account for mass flow through the valve

and closing load versus opening load. The only vendor

with high energy flow test experience has observed that,

as the valve closes the mass flow through it adds a

significant force resisting valve closure. For this reason

closing thrust requirements are greater than opening

thrust requirements for the same differential pressure

across the gate. The gate friction factor used varies with

conditions and valve design over a wide range in contrast

with a single value used by most other vendors.

Under-estimating the valve loads stated above will

most often result in an undersized motor operator or low

torque switch settings on the motor operator. The

GeneralElectric Company standarddesign (forthenewer

BWR plants) for the motor-operated valves used in the

systems of interest under 01-87 employs the following

control switch scheme. In the opening direction a

position limit switch contact is used to control valve

stroke. The use of only a limit switch contact eliminates

the possibility of the valve not opening on demand due to

an incorrectly set torque bypass switch. In the closing

direction both torque and limit switches, connected in a

parallel arrangement are used to control valve stroke.

This arrangement allows for positive valve closure by

using the limit switch to control valve disc movement

until the pointjust prior to disc seating. At this point the

limit switch drops out of the circuit and the torque switch

controls disc seating thereby preventing valve disc

damage due to overtorquing the disc into the valve's seat.

For the older BWR plants, the torque switch controls

valve movement throughout the entire closure stroke.

As part of the valve procurement process, the vendors

were required to prove valve operability at maximum

system pressures. Flow interruption tests of valves in the

size range of interest to 01-87 are very expensive, time

consuming andrequire a large flow facility. As indicated

above, only one vendor uses equations backed by actual

flow interruption testing. In this case a 14 inch gate valve

was tested with steam flow. All details about the test are

considered proprietary by the vendor.

The remainder of the vendors use a substitute test to

provevalveoperability. In thistestthevalveisclosedand

full pressure is applied across the gate. The valve is then

opened, the inference being made that the thrust required

to open the valve is greater than that required to close it.

The argument for this is that the pressure drop across the

gatewhileclosingoff flowtoabrokenpipecannotexceed

the full pressure and threfore the valve's capability to

close is demonstrated.

11
5. WEAKNESSES IN VENDOR METHODOLOGY

5.1 EPRI Marshall Test Program

Results

Recent test programs suggest that the simplified

approach described in the previous section may not be

justified. In 1980 the Electric Power Research Institute

(EPRI), on behalf of the participating PWR owners,

conducted full flow steam testing on seven typical PWR

PORV Block Valves at Duke Power's Marshall Steam

Station. The results of this testing are described in the

-EPRJMarshall Electric Motor Operated Valve (Block

Valve) Interim Test Data Report"9

The project objectives were to obtain preliminay

information on electric motor operated valves by

performing full flow steam testing. All seven PORV

Block Valves tested were 3 inch 1500 lb class gate valves

of similar design to those identified for GI-87. The

valves were instrumented to measure motor curent and

valve stem position. Pluid pressures and temperatures

were determined from instruments in the test piping;

valve inlet temperature and body temperature were not

monitored. Valvestemstraingaugeswereinstafledonthe

Westinghouse valves at special request from

Westinghouse.

Three manufacturers' valves (Velan, Borg-Warner,

and Rockwell Intertional), as supplied, met the desired

acceptance criteria during the test program. The valve

assemblies fully closed and opened with little seat

leakage for fall flow and differential pressure conditions.

One manufaiturer's valve (Anchor/Darling) failed to

dose during preevaluation testing with the supplied

operator Excessive seat leakage was also observed. The

valve was returned to the manufacturer where the seats

were modified to increase the seat area, the valve stem

and bonnetreplaced, and a modified operatorofthe same

model was instafledL Retesting with the modified valve

and operator still indicated closure trouble, so a larger

operator capable of greater torque was installed. The

valve assembly was successfully tested. This operator

was then replaced with an operator of the same size as

originally supplied. Afterverification of correct operator

to stem alipment and setting of the closingtorque switch

settings to approximately maximum, the valve dosed

completely under fill flow test conditions.

A second manufacturer's valve (Westinghouse) also

experienced closure failures on two different models

tested. Testing indicated that the Model 30M88 valve

with the vendor-recommended operator and torque

switch setting was insufficient to reliably dose the valve.

Increasing the closing torque switch setting allowed the

valve to completely close reliably with ltlde or no seat

leakage for the full flow steam test conditions. The

model 3GM99valve with the recommendedoperatorand

torque switch settings would not completely close the

valve under full flow conditions. Based on valve stem

strain measurements, a larger operator was installed and

the valve passed the EPR/Marshall testing sequence (the

Model 3GM99's operator was also rewired to close using

the dose limit switch instead of the close torque switch).

Additionaltestingwasperformedwiththelargeroperator

rewired in its normal mode, i e., to deenergize the motor

onthe close torque switch. Again, the valve did not close

completely under full flow conditions.

Table 2presents a matix of the valves testedversus the

operators used and indicates whether they conpletely

opened or closed. Valve functionability was successfully

demonstrated for three of the five valve manufacturers,

even though the valves with closure problems used

equivalent operators. Stem load is then a function of not

only the fluid conditions but also the valve design (ie.,

wedge seat, materials, surface finishes, guilding, etc.). It

is evident that, for some valve manufacturers, the actual

stem load required to close the valve is quite different

from the calculated stem load used for sizing the

operators. All failures occurred during the closing cycle,

casting serious doubt on the appropriateness of using

valve opening tests at full differential pressure to prove

dosing cycle operability in a pipe break environment.

13
Table 2. Valve/operator matrixW

Operator

Limitorque Rotork

SB-O0-15 SMBf-00010 14-NAI 16-NAI 16-NAX1 30-NAI

Valve Open _ OClosep CLO Open aose Open lose Open Close

Velan B 10-3054B-13MS

Westinghouse 3GM88

Westinghouse 3GM99

Anchor/Dadling DoubleDisc 5J-1512

Borg Warner 79294

Rockwell Inter. 1309460

Velan B 10-3054B-13MS

Yes

Yes

"vow

Yes

Yes

XNnc

- - Yes No - - Yes Nob

%F-. %,T-d .IV* ItsO LISIV- -

- - - - Yes - Noa

- Yes Yes - - - -

- - Yes Yes - - - -

Yes Yes

Yes Yes

a. The valve did not completely close on demand with the Rotork 14-NAI operaLor.

since the SB-O15 was not electically compatible with the Marshall Facility.

The Rotoik 14-NAl was substituted for the Limitorque SB-00-15

b. The Rotoirk 14-NAl was replaced by a Rotoik 16-NAX1. The valve completely closed on demand only when the torque switch was bypassed.

c. The Limitorque SB-0-15 replaced the SMB-OO10. The valve completely closed on demand only when the SB-W15 was rewired to close using the

limit switch instead of the torque switch.

d. The Limitorque SMB 10 did not completely close the valve on demand

e. The Rotork 16-NAXl even when modified could not completely close the valve on demand.
5.2 Westinghouse Test Program

Results

At the completion of the EPRI/Marshall test program,

Westinghouse conducted additional testing on the

Wesndghouseelectricmotoroperatedvalves. The"EPRI

Summary Report: Westnghouse Gate Valve aosure

Thsting Program" contains the results of this test

program. 11 Although Westinghouse valves were not

identfiedinthe valve survey and are probably notusedin

BWR plants, they are similar enough in design to those

valves utilized in BWR systems to make the following

information of generic importance to GI87.

To determine the causes of the higher than expected

stem thrust measured during previous tests, the

Westinghouse Electro-Mechanical Division undertook

three testing programs.

1. A series of 50 separate water flow tests were

conducted against 60 to 600 gpm flow and

1500 to 2600 psi differential pressure.

2. A mechanical fixture test was conducted

using ahydraulic cylinder to apply simulated

flow loads to the valve disc.

3. Friction factor tests were performed,

utilizing small sanples cut from the faces of

actual discs and seat rings.

The test results indicate that:

1. The friction factor at room temperature will

increase from as low as 0.12 until a level of

0.4 to 0.75 is fhed at 100 to 200 cycles.

2. The magnitude of the friction factor at 1800F

is higher than at room temperature with peak

values of 0.64 to 1.00.

3. Dry data indicates little change in friction

factor occurs with cycling, and that the

friction level is approximately 0.3.

4. With 5500F steam, at a %0.l-inch stroke

length, the friction factor starts in the 0.5 to

0.6 range and drops quickly to

approximately 0.35.

5. Pause time under load (closed position)

increases the friction factor, while pause time

under no load (open position) decreases the

friction factor.

As a result of the EPRI/Marshall, Almarez, and

Westinghouse test programs, Westinghouse concluded

that the valve closure problems were the result of

under-predicting the friction load and therefore

under-estimating the stem thrust required to close the

valve against high differential pressures. Although tests

showed friction factors ranging from 0.1 to 1.0,

Westinghouserecommendedthat afriction factorof 0.55

be used in Equation (1).

It should be noted that several of the other valves in the

EPRI program closed successfully even though their

operators were most likely sized using the 03 disc

friction factor. Westinghouse explained this as most

likely resulting from the difrence in operator sizing

philosophy between Westinghouse and most other valve

companies. Most other companies allow iUmitorque

Corporation to perform their operator sizing.

Westinghouse suggests that the standard Limitorque

technique may have sufficient margin built into it at other

points of the sizing calculation that the final operator size

is adequate and most valves would close at the higher

actual loads. These added margins can result in operator

stall output loads that can damage a valve not designed to

accept them. Westinghouse attempted to minimize the

potential for damage by reducing operator margins,

making the Westinghouse design less tolerant of

underestimation of closing thrusts.

5.3 Ontario Hydro Flow Test

An additional valve flow internrption testprograrn has

been performed. The bulk of the test results are

proprietary, however, a few general results are available.

Ontario Hydro perfonned aflow internuption test of an

8 inch, 900 lb, wedge type gate valve with an electrical

motor operator. The test was performed for New

Brunswick Power, at the Ontario HydroNuclearProcess

Components Testing Facility in Toronto, Ontario,

Canada.

15
The valve test was a blow-down type test, with

saturatedwaterat approximately 252Fflashingtosteam

through the valve. The water source was limited and

could not maintain maximum flow throughout valve

closure. Actual test measurements are proprietary and

only the following "bottom-line" result is public. The

valve failed to operate with the recommended operator

torque settings supplied by the valve and operator

manufacturers.

16
6. CONCLUSIONS

The valves that must be qualified to resolve 01-47 are

pressure-seal cast steel, gate valves in the 3 to 10 inch

range and 600 and 900 lb. class. The most common

manmfactureris AnchorD)arling Valve Company. Valve

operatoesin use on these valves are electric motor driven

(AC and DC) operators, manufatured by Limnitorque

Corporation. Typical system design conditions average

1250 psi and 575F

The mitigation of high energy pipe breaks are within

the design basis for the above valves. Utilities typically

purchase motor-operated valves which are certified

under the manufacturer's Quality Assurance program to

meet the design requirements established by the plant

designer. Tbeir method for establishing the qualification

of the valve assemblies is to confirm that the certified

performce of the motor-operated valve meets the

design requirements of the system.

The same equation forsizing operators is used by most

ofthe valve and operatormanufacturers. This equation is

simply the sum of three terms, the disc drag due to

differential pressure load, the stem endpressure load, and

thepadkng drag load. Fow through the valveistypically

not factored into these equations The equations depend

heavily on the value used for the disc friction factor,

which varies with vendor. iypical values are 0.2 and 0.3.

This is inconsistent with recent test data, where disc

friction factors anged from 0.1 to 1.0.

The study of vendor methodology has identified

several important parameters to be considered in the

prediction of valve stem thrust loads. The specific

relatonship between these parameters and the stem thust

are not well understood. Differences of opinion exist in

the following areas:

1. The effects of high mass flow on valve

closure loads.

2. The ability to bound closing loads with

substitute tests where the valve is opened

starting at full differential pressure.

3. The correct disc friction factorforgate valves

as a function of the other valve and operator

parameters.

4. The effects of valve cycling on stem loads.

Very few tests under actual high energy pipe break

conditions have been performed by utilities or valve and

operatormanufacturers Only one vendorhasblowdown

isolationtestexperience, the others quotepastexpenience

in the commercial power industry to justify their

methods. Operability of the valve assemblies is

demonstrated using a substitute test where the valve is

opened against fall differential pressure. No data was

found supporting the presumption that opening load with

full differential pressure will bound the closing load at

full system pressure and mass flow.

The few flow interruption tests that have been

conducted, althoughnot specificaly designedto measure

these phenomena, have identified the following general

trends.

1. Repeated cycling tends to increase the valve

thrust required to operate the valve.

2. The industry standard 0.3 disc friction

coefficient is not conservative for all cases

and may vary significantly from this nominal

value. Coefficients have been measured

from 0.1 to 1.0.

3. Mass flowAnomentum could have a

significant effect on valve stem thrust loads.

4. Increased temperature causes a significant

increase in the required valve operating

thrust

The qualification of the isolation valves in the HPCI

and RCIC steamlines and the RWCU suction line to close

under high energy pipe break conditions is questionable.

Evidence exists that, for some manufacturers, the actual

stem load required to close the valve is quite different

from the calculated stem load. Valves have failed to fally

close in test programs where the valve assemblies were

specificallydesigned for the test conditions usingpresent

qualification methods.

17
7. RECOMMENDATIONS

Ihe review oftestdata andqualificationtechniques has

provided information suggesting deficiencies in current

closure load prediction and qualification practices.

Further work is recommended as described below.

Additional independent test data should be obtained to

dearly quantify the influence of the various parameters

on valve closure loads. Based on the testing reviewed in

this report, more information is required to provide

confidence in our ability to define a conservative value

forthe frictionloadonthe disc. Specifically the effects of

cycling, seat and disc material specification, and

temperature on the friction load should be evaluated In

addition, the previous results indicating that the friction

load is proportional to pressure drop and independent of

flow rate should be confirmed. Test data should be

obtained through two methods:

1. Evaluate existing data from test laboratories,

vendors, and the open literature. Test reports

have been identified that, although

proprietary, are available forreview on-site.

2. Generating data from new independent tests.

The new testing would be designed to confirm selected

importantresults from utility-orvendor-sponsoredtests

and to address anticipated deficiencies (gaps) in the

existing experimental results.

19
8. REFERENCES

1. NRC memorandum from Harold R. Denton, Director Office of Nuclear Reactor Regulation to Robert B. Minogue,

Director Office of NuclearRegulatory Research, "Licensing Need ForSelected Tasks UnderMechanical Equipment

Qualification Research Program (FIN A6322)," March 11, 1986.

2. F J. Molterus, et. al., Qual[#cation of Active Mechanical Equipment For Nuclear Plants, EPRI NP-3877,

March 1985.

3. USNRC EE Bulletin No. 85-03, "Motor-Operated Valve Common Mode Failures During Plant Transients Due to

Improper Switch Settings."

4. Letter from L. G. Kunel, Nebraska Public Power District to Robert D. Martin, Region IV, USNRC, "Response to lE

Bulletin No.85-3, Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46" May 15, 1986.

5. Letter from Murray R. Edelman, The Cleveland Electric illuminating Company to James G. Keppler, Region m,

USNRC, "Perry NuclearPowerPlant, Docket Nos. 50-400 and 50-401, lE Bulletin 85-03, Motor-Operated Valve

Failures," May 14, 1986.

6. Letter from Corbin A. McNeill, Jr., Public Service Electric and Gas Company to Dr. Thomas E. Murley, Region I,

USNRC, "Motor-OperatedValve CommonMode Failures DuringPlantTransients Due tolmproperSwitchSettings

Salem and Hope Creek Generating Station Facility Operating Licenses DPR-70, DPR-75 and NPF-50," May 27,

1986.

7. Anchor/Darling Valve Company, "Valve Seminar"

8. General Physics Corporation, "Nuclear Valve and Valve Operators:' 1983.

9. MarshaU Electric Motor Operated Valve (Block Valve) Interim Test Data Report, EPRI NP-2514-LD, July 1982.

10. C. A. Seaquist, EPRI PWR PORV Block Valve Test Program: Final Review Report, EGG-ED-6386,

September 1983.

11. Westinghouse Electro-Medhanical Division, EPR! Summary Report: Westinghouse Gate Valve Closure Testing

Program, March 31, 1982.

21
APPENDIX A

IDENTIFICATION OF GI-87 VALVE ASSEMBLIES

A-1
APPENDIX A

IDENTIFICATION OF GI-87 VALVE ASSEMBLIES

Two general tasks were undertaken to assess the

population of the containment isolation valves usedinthe

BWR systems ofinterest underGI-87. First was areview

of available information sources to identity the systems

applicable to GI-87 and to determine the valve and

actuator manufacturers, types, and sizes used in those

systems. The second task was a survey of vendors to

deternine industry methods of qualification. These tasks

are discussed in the following paragraphs.

A.1 Identification of Systems

Covered by G0-87

Information obtained from ftiminazyjFina Safety

Anaysis Reports (PSAR*/FSARs) permitted the

identification of those systems that penetrate

containment and directly communicate with the reactor

vessel or recirculation lines. Tables A-1, A-2, and A-3

list the systems that meet these criteria. Table A-1 shows

the systems for the BWR-2s, Table A-2 covers the

BWR-3s and BWR-4s, and Table A-3 covers BWR-5s

and BWR-6s. The fifth column in each table lists the

operational status of the valves in each of the systems.

Since GI-87 is concemed with the capability of isolation

following a line break, only those systems with valves

normally open were chosen for further study. The ines

with checkvalves toprevent flow out of the actorvessel

were not investigated. The Main Steam Lines are also not

included under GI-87.

The systems chosen for further study are listed below

with a brief description of the specific line under

consideration

1. Emergency Cooling System-steam leaving

reactor (BWR-2 only)

2. HPCI-turbine steam supply (BWR-3&4

only)

3. RCIC/Isoladon Condenser-turbine steam

supply

4. RWCU/Cleanup-water leaving reactor.

Table A-4 lists BWR plants and the plant-specific

systems covered in this study.

Figures A-1 through A-4 are typical schematic

drawings of these systems showing connections to the

primary system and valve location and status. Hollow

valve symbols indicate that the valve is open during

normal plant operation.

A.2 Valve Sizes and Design

Conditions

The PSAR/FSAR system data contained limited

information about the containment isolation valves and

operators. Restricting the search to the four systems

previously identified, 84 process lines were studied.

Each line has two containment isolation valves. With the

exception of two plants where both isolation valves are

located outside of containment, one valve is inside

containment and the other is outside containment All

PSARs/FSARs containing operator information

identified the inside containment isolation valve as

having an AC power source, while the outside

containment isolation valve had a DC source. Gate

valves were identified as the type of valve used in all but

two plants where globe valves were used. Complete

system descriptions were not provided in all FSARs;

however, the information available was very consistent

from plant toplant and variation in those plants without a

detailed FSAR are expected to be minor. The following

paragraphs discuss the results of this literature search for

the four chosen BWR systems.

The Emergency Cooling system is used only on

BWR-2s. The system consists of two lines penetrating

containment, each with two isolation valves located

outside of containment. The system design pressure and

temperature are 1250 psi and 5750F respectively. The

pipe size for this system was not identified.

The HPCI is a 10-inch system with design pressures

and temperatures ranging from 1120 to 1250 psi and 558

to 5750F respectively. All valves identified are gate

valves.

A-2
Table A-1. Systems forBWR-2

Line or System

Number

of Lines

vaves

Connection per Line

Status

(Normal Position)

Main Stream 2

Main Stream

Wam-Up

Emergency Cooling Vents

2

2

RPV

RPV

RPV

RPV

RPV

RECIRC

2

1

2

2

2

2

Open

Closed

Open

OpenlChecc

Open

Closed/Check

Feedwater 2

Emergency Cooling

Steam Leaving Reactor

Cond. Return to Reactor

2

2

ReactorCleanup

Water Leaving Reactor

Water Return to Reactor

Shutdown Cooling

Water LIaving Reactor

Water Return to Reactor

Reactor Head Spray

1

1

1

1

I

RECIRC

RECIRC

RECIRC

REC1RC

RPV

2

2

2

2

2

Open

Opentxeck

Closed

Closed

Closedaheck

Liquid Poison

Control Rod Drive Hyd.

Core Spray

1

1

RPV

RPV

RPV

2

2

Check

2 3 2-OpewClosed

A-3
Table A-2 Systems forBWRs-3 and -4

Line or System

Main Steam

Main Steam Drain

Feedwater

Reactor Water Sample

Control Rod Drive Return

RWCU/Cleanup

Water Leaving Reactor

Water Return to Reactor

Number

of Lines

4

1

1

1

4

1

1

Connection

RPV

RPV

RPV

RPV

RPV

RECIRC

RECIRC

Valves Status

perLine (Normal Position)

2 Open

2 Closed

2 Open/Check

2 Closed

2 Check

2

2

Open

Openoiheck

RHR

Shutdown Cooling

Supply

Return

LPCI Return to Reactor

Reactor Head Spray

Standby Liquid Control

IC/RCaC

Steam Supply

Con& Return

Core Spray

HPCI Steam Supply

1

2

2

1

1

1

1

2

1

RECIRC

RECIRC

RECIRC

RPV

RECIRC

RVP

RECIRC

RPV

RPV

2

2

2

2

2

2

2

2

2

Closed

Closed

Closed/Check

Closed/Check

Check

Open

ClosedlOpen

Closed/Check

Open

A-4
Table A-S. Systems forBWRs-S and-6

Line or System

Main Steam

Main Steam Drain

Feedwatr

Reactor Water Sample

Control Rod Drive Retrun

RWCU/Cleanup

Water Leaving Reactor

Water Return to Reactor

Number

of Lines

4

1

1

1

Connection

RPV

RPV

RPV

RPV

RPV

RECIRC

RECIRC

Valves Status

per Line (Normal Position)

2 Open

2 Closed

2 OpeAl/becck

2 Closed

2 Check

2

2

Open

Openjcheck

REIR

Shutdown Cooling

Supply

Return

LPCI Return to Reactor

HPCS Return to Reactor

Standby Liquid Control

RCIC Steam Supply

RC1C RPV Head Spray

Core Spray

1

2

2

1

1

1

1

2

RECIRC

REC~IRC

RPV

RPV

RECIRC

RPV

RPV

RPV

2

2

2

2

2

2

2

2

Closed

Cosed/Check

aosed/Check

Closed/chck

Check

Open

Check

aosed/Check

A-5
Table A-4. BWR plant listing

FWCI, HPCI or Emergency

Plant Name BWRClass TypeCo nt IC orRCIC HPCS Cleanup or RWCU Cooling

Oyster Creek 2 Mark I IC FWCI Cleanup ECCS

Nine Mile Point 1 2 Mark I IC FwCI Cleanup ECCS

Dwesden 2 and 3 3 Mark I IC HPCI Cleanup

Millsone 1 3 Mark I IC FWCI Cleanup ECCS

Monticello 3 Madr I RaC HPCI Cleanup

Quad Cides 1 and 2 3 Mak I RCIC HPCI RWCU

Pilgrim 3 Mark I RaC HPCI RWCU

Brown's Ferry 1, 2, ad 3 4 Mark I RaC HPCI RWCU

Vermont Yankee 4 Mark I RaC IPC1 RWCU

Duane Arnold 4 Mark I RaC HPC1 RWCU -

Peach Bottom 2 and 3 4 Mark I RaC HPCI RWCU

Cooper 4 Mak I RaC HPCI RWCU -

Hach and 2 4 Mark I RCIC HPCI RWCU -

Bmnswick 1 and 2 4 Mark I RCIC HPCI RWCU -

Fitzpatrick 4 Mark I RaC HPCI RWCU -

Enrico Fermi 2 4 Mark I RaC HPCI RWCU -

Hope Creek 4 Madk I RaC HPCI RWCU -

Susquahanna 1 and 2 4 Mark II RCIC HPCI RWCU -
TAble A-Z (continued)

Plant Name BWR Class Type Containment

PWCL HPCL or

IC or RCaC HPCS Cleamnp or RWCU

Emergency

Cooling

Shorcham

Limerick 1 and 2

La Salle County 1 and 2

WNP 2

Nine Mile Point 2

Grnd Gulf 1 and 2

Perty 1 and 2

RiverBend 1

Clinton 1

4

4'-

5

5

5

6

6

6

6

Malik

Makikl

Makikl

Malkfl

Mark il

MOiM

Markm

MOrM

Markm

RaC

RCC

RaC

RaC

RCC

RaC

RaC

RaC

RaC

HPCI

HpCl

HPCS

HPCS

HPCS

HPCS

HPCS

HPCS

HPCS

RWCU

RWCU

RWCU

RWCU

RWCU

RWCU

RWCU

RWCU

RWCU

2J
Emergency cooling

supply

Emergency cooling

return /

!z

gYj10

510 312

Figure A-1. Tpical Emergency Cooling System.

A-8
Turblne

Hoag stop

valve

Turbine

HO control

valves

tank 1 ElE

Feedwater

Input

6 10311

Figure A-2. Typical High Pre== Coolant Iection systm.

A-9
Feedwater

input

6 10 310

Figure A-3. ipical Reactor Core Isolation Cooling system.

A-10
C

0

a

kV

ft

fedae

'1

iP

I0

Primary

containment Cleanup

recire.

pumps

Cleanup Cleanup

regen. non-regen.

heat exchangers heat exchangers

feedwater

return

Main

Waste _

collection

and surge tank

6s10 3
The RCIC and Isolation Condenser systems range in

size from three to 14 inches with the majority being three

and four inch lines and gate valves. Valves greater than

four inches were identified in only three plants, one of

which gave the valve configuration as two 10-inch gate

valves with a 1-inch by-pass globe valve. The Isolation

Condenser system identified in the BWR-2s and early

BWR-3s conanedthe majority of the large (greater than

4 inches) valves System design pressures and

temperatures covered the same range as those for the

HPCI system.

The majority of the RWCU systems include six-inch

gate valves; thre- and four-inch gate valves were

identifiedin two plants each. The fouroldest plants inthe

study use the Cleanup System which includes four-, sixand eightinch valves. The design pressures and

temperatures range from 1250 to 1450 psi and 564 to

5750 respectively.

Plant-specificsystem details areprovided, as available

from the FSARs, in Appendix B.

The Institute of Nuclear Power Operations (INPO)

Nuclear Plant Reliability Data System (NPRDS) was

usedto determine specific valve assembly information. It

provided the valve manufacturer, model number, type,

size, maximum pressure and maximum temperature for

the High Pressure Coolant Injection system (HPCI) and

the Reactor Core Isolation Cooling system (RCIC). The

NPRDS also provided valve operator manufacturer,

model number, type, power source, maximum force, and

maximum torque for these same two systems. The data

base contained HPCI valve and operator data for 22

plants or 81% of the BWRs having that system; it

contained RaC valve and operator data for 24 plants or

67% of the plants having the RCIC system.

The most predominate valve manufacturer for both

systems is the Anchor/Darling Valve Co. with 41% ofthe

containment isolation valves. The other manufacturers

are Borg-Warner (2%), Crane Co. (18%), William

Powell Co. (11%). Velan Inc. (16%), and Walworth Co.

(12%).

Limitorque Corporation manufactured 94% of the

valve operator The remaining 4 valve operators are

identified as Philadelphia Gear Corporation operators.

Figures A-5 and A-6 show the distribution of HPa

and RCIC valve sizes among the various manufaturrers.

The HPC systems (Figue A-5), with the exception of

one plant, contain 10-inch gate valves exclusively while

the RCIC systems contain 3-,4-, 8-, and 104nch valves.

The containment isolation valves in the BWR-3s and

BWR-4s are 3- and 4-inch gate valves, with the 34nch

valve being slightly more predominate. The RCIC lines

were combined with the Residual Heat Removal (RHR)

System in the BWR-Ss and BWR-6s resulting in an

increase inthe pipe size to 8 and 10 inches. One plant has

8-inch valves while 3 plants have 10-inch valves.

The NPRDS data also contained the model number or

vendorfigurenumberforeachvalve. Vendormarketing

literature and direct communication with vendor

representatives identified the "typical GI-7

containment isolation valve:

Type: Gate Valve

Size: 3 to 10 inches

Class: 600 and 900 lb

Body: Cast Steel

Bonnet Pressure-Seal

Disc: Flexible Wedge.

A-12
0I

Manufacturers by code

Figure A-5. HPC[ isolation valve manufacturers by size.

25

20

IS

10

I

z8

AID B-W CC WPC VI WC

Manufacturers by code

Figure A-6. RCIC isolation valve manufacturers by size.

A-13
APPENDIX B

PSAR/FSAR DATA

B-1
PSAR/FSAR DATA

Design

Plant

Nine Mile PT.--

Oyster Creek

Dresden-2

Dresden-3

Monticello

NMine Mile PrT.-

BWR

Clas Service

2 Cleanup Supply

2 Cleanup Supply

3 Cleanup Supply

3 Cleanup Supply

3 Cleanup Supply

2 ECCS Steam

Supply

Pipe

Size 6889

Type Status PSIG

Gate

Gate

Open 1300 - 1250

Open 1250

Open 1250

Open 

-

OF

575

575

Inside

Valve

ID

1201-1

1201-1

MO-2397

Outside

Valve

ID Notes

1201-1 

-

1201-1 

-

M02398 

-

-

- Open 1250 575 ab

DIesden-2

w Dresden-3

Monticello

Pilgrim-I

Quad Cities-i

Quad Cities-2

Browns Ferry-I

Browns Ferry-2

33333344

HPC: Steam

Supply

BPCa Steam

Supply

HPa Steam

Supply

HPC[ Steam

Supply

Hpa Steam

Supply

HPCa Steam

Supply

H!pCI Steam

Supply

iPa steam

Supply

10

10

iA

44

1.1V -

4

-4

-4

10

10

3ate Open 1125

3ate Open 1125 - COpen 1125

Gate Open 

-

Gate Open 

-

3ate Open 

-

- Open 1120 - Open 1120

558

558

558

2301-9

2301-9

MO-1S

2301-5

2301-5

MO-16

2301-5

2301-5

- 2301-4 - 2301-4

a. Two lines penetrate containment.

b. Both valves located outside containment
PSAR/FSAR DATA (Continued)

Valve

Outside

Valve

PLV9

Browns Feuy-3

Brunswick-i

Brunswick-2

Cooper

Amold

Enrico Fenmi-2

T Hateh-4

Hatch-2

Fltrick

m 1ck-1

Limedck-2

Peach Bottow-2

Peach Bottom-3

BWR

Class Sennce

Pipe

size Type Stawu PSIG Op ID ID Notes

HPCI Steam

Supply

HPa Steam

Supply

HPa Steam

Supply

HPCI Steam

Supply

HP[ steam

Supply

HPl Steam

Supply

HPa Steam

Supply

HPCI Steam

Supply

HPCI stem

Supply

HPC Stema

Supply

HPCI Steam

Supply

HPC Steam

Supply

10 

- Open 1120

-- Open 

-

-- Open -

- Gate Open 

-

10 Gate Open 1250 - Gate Open 1250 - Gate Open 1250

10 

- Open 

-

-

- Open 

-

-- Open 

-

10 Gate open 

-

575

575

575

E41F002 1341F003

4 HPCI Steam

Supply

10 Gate Open
PSAR/FSAR DATA (Continued)

Design

Inside

Valve

Outside

Valve

Plant

VeWrmont Yankee

BWR

Class Service

4 HPC Steam

Supply

Pipe

Size Type Status I'51 OF ID ID Notes

10 

- Open

Oyster Creek

Millstone-I

23

IC Return

IC Return

10 Open 1250

Open 1250

575

575

OysterCreek

Dresden-2

Dresden-3

Millstone-1

Monticello

Pilgrim-1

Quad CitesQuad Cities-2

Brow Ferry-l

Browns Ferny-2

Browns Ferry-3

IC Steam

Supply

IC Steam Supply

IC Steam Supply

IC Steam StpPly

RCIC Steam

Supply

RCIC Steam

Supply

RCIC Steam

Supply

RCIC Steam

Supply

RCIC Steam

Supply

RCYC Steam

ISupply

RCIC Stean

Supply

10 

- Open 1250

14 Gate Open 

-

14 Gate Open 

-

-

- Open 1250

3 

- Open 1135

3 Gate Open 1340

- Gate Open 1135 - Gate Open 1135

3 

- Open 1146

3 

- Open 1146

3 

- Open 1146

575 - 1301-1 - 1301-1

575 

-

582 MO-2075

562 1301-16 - 1301-16 - 1301-16

1301-1

1301-2

MO-2075

1301-17

1301-17

1301-17

562

562

562
PSAR/FSAR DATA (Continued)

Design

BWR

Plant Class

Brunswick-i

Bunswick-2

Cooper

Arnold

Enrico Fermi-2

Hatch-l

Hatch-2

Fitzpatuick

Limerick-i

Limenck-2

Peach Bottom-2

Peach Bottom-3

Service

RCIC steam

Supply

RCIC Steam

Supply

Rac Steam

Supply

Rac Steam

Supply

RCIC Steam

Supply

RCIC Steam

Supply

RCIC Steam

Supply

RCIC Steam

Supply

RC Steam

Supply

RCIC Steam

Supply

RaC Steam

Supply

RaC Steam

Pipe

Size 3334433333333

Type Sotzs - Open - Open

Gate Open - Open

Gate Open

Gate Open

Gate Open - Open - Open - Open

Gate Open

Gate Open

Inside

Valve

PSIG 'F ID

1500 560 P007

1500 560 F007 - - MO-15 - - MO-24

1250 575 E51FOO7

1250 575 F007

1250 575 P007

1250 575 MOV-1S - - MO-1S - - MO-1S

1120 

- MO-1S

1120 - MO-IS

Outside

Valve

ID

F008

F008

MO-16

MD-24

E51F008

F008

F008

MOV-16

MO-16

MO-16

MO-16

MO-16

Notes

Supply

Vermont Yankee 4 RCIC Steam

- Open 1250 575

a

Supply

a. Inside valve open/outside valve closed
PSAR/FSAR DATA (Continued)

Design

Inside Outside

Plant

BWR Pipe

Cass Service Size Type Status P_10

Valve Valve

OF ID ID Notes

La Salle Co.-i

La Salle Co.-i

La Salle Co.-2

La Salle Co.-2

WNP-2

Clinton-1

- Perry-1

Mllstone-i

PilgriimQuad Cities-i

Quad Clties-2

Browns Fer}y1

Browns Ferry-2

Browns Ferry-3

Brunswick-1

Bnmswick-2

5 RClC Steam

Supply

5 RCIC Steam

Supply

S RCIC Steam

Supply

5 RCIC Steam

Supply

5 RCIC Steam

Supply

6 RC3C Steam

Supply

6 RCIC Steam

Supply

10 Gate Open

I Globe aose

10 Gate Open

1 Globe Cose

3 Gate Open - - Open - Gate Cose 

I

-

- Open

6 Gate Open

- Gate Open - Gate Open

6 

- Open 

:

6 - Open

6 - Open:

6 

- Open

6 - Open I

575

575

575

575

E51F063

E51F076

E51F063

E51F076

D51F064

ab

E51F064

ab

-- P007-1

333344444RWCU Supply

RWCU Supply

RWCU Supply

RWCU Supply

RWCU Supply

RWCU Supply

RWCU Supply

RWCU Supply

RWCU Supply

575

575

575

575

575

575

564

564

F063

F063

1201-2

1201-2

1201-2

PFO1

F001

P008

F064

P064

1201-5

1201-5

1201-5

P004

P004

a. Inside valve open/outside valve dlosed.

b. Bypass line.
PSAR/FSAR DATA (Coin 

)

Desgm

BWR

Plant aawS Sere

Pine

Size Typ Swum PSIG 1OF

Inside Outside

Valve Valve

ID ID Notes

Cooper

Ain

Enrico Fermi-2

H1ch-1

Hatch-2

Limerick-i

Limerick-2

Peach Bottom-2

w Peach Bottom-3

Vermont Yanke

La Salle Co,-l

La Salle Co.-2

WNP-2

cuiton-i

Perty-I

4 RWCU Supply

4 RWCU Supply

4 RWCU Supply

4 RWCU Supply

4 RWCU Supply

4 RWCU Supply

4 RWCU Supply

4 RWCU Supply

4 RWCU Supply

4 RWCU Supply

4 RWCU Supply

5 RWCU Supply

5 RWCU Supply

5 RWCU Supply

6 RWCU Supply

6 RWCU Supply

6 Gate Open

6 - open

6 Gate open

6 Gate Open

6 Gate Open

6 - Open

3 - Open

3 - Open

6 Gate Open

6 Gate Open

4 - open

6 Gate open

6 Gate Open

6 Gate Open

_ - Open

- Gate -

- 575 - 564

1250 575

1250 575

1250 575

1250 575 - 564 - 564

1120 575

1120 575

MO-1S

MO-2700

G33F001

FOOl

FOOI

MOV-15

MO-iS

MO-lS

MO-1S

MO-1S

MO-16

MD-2701

G33F004

F004

F004

MOV-18

MO-18

MO-18

MO-16

MO-16

G33F004

G33F004

P004

F004

F004

1250

1250

1250

1250

1250

- G33F001 - G33F011 - P001

575 Fool

575 F001
NRC FORM 35 US. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER NRC t249 OM 8S UnRd Q Addi Vol.. Iup ..It" 112335u wbm tet"

3201.3202 BIBLIOGRAPHIC DATA SHEET * ""' .

0" ostrwrios and OVA}rj UREG/CR-5406

2. TITLE AND SUBTITLE EGG-2569

BWR Reactor Water Cleanup System Flexible Wedge Gate Isolation Vol. 3

Valve Qualification and High Energy Flow Interruption Test M OATEREPORTPUHLISHED

Review of Issues Associated with BWR-Containment October 1989 Isolation Closure Valve ~~~~~~~~~~~4. FIN OR GRANT NUMBER Isolation Valve Closure A6857 and B5529

5. AUTHORISI 6. TYPE OF REPORT

Kevin G. DeWall Technical

Robert Steele, Jr. 7. PERIOD COVERED om'C aehareas

8. PE RFORM ING ORGAN IZATION -NAM E AN D A0DR ESSfitf NRcw~ide OVhi, Ufaf of Region, Utteaap Bn t wlh on t1nsopt sarne andme Au addims

Idaho National Engineering Laboratory

EG&G Idaho, Inc.

P.O. Box 1625

Idaho Falls, ID 83415

9. SPONSORING ORGANIZATION - NAME AND ADDRESS ftINRC. er osebo -;0.nwtmfactor.pva NRC DivIIOn. Ofef or Rhobn. V.& NuewaRugu rooCmlaslon.

wnd mallng addretal

Division of Engineering and Division of Safety Issue Resolution

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT 12o0 wormr rste

This report presents the measured data and the analyses performed to date on the full-scale high-energy

qualification and flow interruption gate valve testing to develop technical insights for the United States Nuclear

Regulatory Commission (USNRC) effort regarding Generic Issue 87 (0187). The research was sponsored by the

USNRCt and conducted by researcheru from the daho National Engineering Laboratory. We tested two 6-i, 900lb

class valve assemblies, which epresent a significant percentage of the reactor water cleanup isolation valves installed in

plant applications. These valves were modified before testing by adding a high temperature load cell in the valve stms,

which allowed the direct measurement of valve stem thrust during both opening and closing valve cycles.

Instrumentation installed in the flow loop and on the valve assemblies measured the important valve and system test

responses. Additionallyduringthetestprogram, allofthecuntlypopularmotoroperatedvalvediagnosttestisystems

monitored the performance of the valves. Initially the valves were subjected to the hydraulic and leakage qualification

tests defined in ANSI B16Al and then to flow interruption and reopening valve tests at boiling water reactor primary

system water temperature and pressure conditions with downstream line break flows. For the two valves tested, results

show that (a) the disc factor used in current industry motor operator sizing equations underpredicts actual valve thrust

requirements at all high temperature loadings, and for one valve design the equations may require an additional term to

account fornonlinearperformance, (b) the thrusts required to close the valves were sensitive to the fluidtemperature, and

(c) the results of testing at lower pressures, temperatures, and flows cannot be extrapolated to design basis pressures,

temperatures, and flows for valve designs that have not exhibited linear performance behavior during design basis

27 rotoypTical testmn. "' h___ WV1VALAWLITYSTATEMENT II _ _so 

valve testing

BWR reactor water cleanup

Generic Issue 87 (GI-87)

Unlimited 14. SECURITY CLASSIFICATION

(T{aS ega

Unclassified

_mwffo

Unclassified

15. NUMBER OF PAGES

16. PRICE

-7 - _________________________

NRC FORM 335 (249)
UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555

SPECAL FOURTH-CLiASS RATE

POSTAGE b FEES PAID

USNRC

PERMIT No. 0G a

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE. $300

0'

0

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pe''

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