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88

Information Notice No. 93­88: Status of Motor­Operated Valve

Performance Prediction Program by The Electric Power Research

Institute

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555

November 30, 1993

NRC INFORMATION NOTICE 93‐88: STATUS OF MOTOR‐OPERATED VALVE PERFORMANCE

PREDICTION PROGRAM BY THE ELECTRIC POWER

RESEARCH INSTITUTE

Addressees

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information

notice (IN) to alert addressees to preliminary results of motor‐operated valve

(MOV) tests conducted by the Electric Power Research Institute (EPRI). It is

expected that recipients will review the information for applicability to

their facilities and consider actions, as appropriate, to avoid similar

problems. However, suggestions contained in this information notice do not

constitute NRC requirements; therefore, no specific action or written response

is required.

Background

On June 28, 1989, the NRC issued Generic Letter (GL) 89‐10, "Safety‐Related

Motor‐Operated Valve Testing and Surveillance," to request that nuclear power

plant licensees and construction permit holders verify the design‐basis

capability of their safety‐related MOVs. In GL 89‐10, the NRC staff requested

that licensees and permit holders test each MOV within the scope of the

generic letter under design‐basis differential pressure and flow conditions,

where practicable. The recommended schedule in GL 89‐10 would have licensees

and permit holders verify MOV design‐basis capability by June 28, 1994, or

three refueling outages after December 28, 1989 (whichever is later).

In response to concerns regarding MOV performance, EPRI and its utility

advisors established a research program to develop a methodology to predict

the performance of MOVs under design‐basis conditions. NUMARC coordinates the

interaction between EPRI, its utility Technical Advisory Group (TAG), and NRC

staff related to the EPRI program. The EPRI program includes detailed

analyses and testing of MOVs at test facilities and nuclear power plants. The

EPRI MOV Performance Prediction Methodology is intended to allow licensees to
demonstrate the design‐basis capability of MOVs based on analytical

predictions combined with diagnostic tests conducted under static conditions.

In August and October 1993, EPRI presented the status and preliminary results

from its Flow Loop Testing Program to the NRC staff. The flow loop results in

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the EPRI presentation have not received full quality assurance verification,

but the preliminary information may be helpful to licensees as they implement

their MOV programs.

In a letter on September 16, 1993, NUMARC provided responses from EPRI to NRC

staff questions on the EPRI MOV Performance Prediction Program. Among the

information provided in the enclosure to the letter, EPRI stated that its

program is expected to cover about 90 percent of gate valves (about half with

its computer code and half with empirically‐based data), essentially all globe

valves, and about 95 percent of butterfly valves. The globe and gate valves

are covered primarily by the computer code. EPRI also stated its method for

determination of operator output torque capability under degraded voltage

conditions is to apply standard methods as documented in EPRI NP‐6660‐D

(March 1990), "Application Guide for Motor‐Operated Valves in Nuclear Power

Plants."

Description of Circumstances

In conducting its MOV Performance Prediction Program, EPRI tested 28 gate,

4 globe, and 2 butterfly valves under a total of 62 test conditions. These

tests were performed at Wyle Laboratories and Siemens test facilities. EPRI

plans to obtain test data for an additional 35 valves being tested in nuclear

power plants. In addition, EPRI completed testing at Kalsi Engineering of

10 butterfly valve designs to assess flow and upstream piping configuration

effects. The results summarized below are based on the Wyle/Siemens MOV

tests.

1. Gate Valves

EPRI stated that all gate valves tested were initially preconditioned by

conducting a large number (50‐1000) of short (no flow) strokes in cold water

under differential pressure loading. Initial "sliding friction coefficients,"

prior to preconditioning, generally ranged from 0.2 to 0.4. EPRI indicated

that, after preconditioning, "apparent friction coefficients" ranged from 0.3

to 0.6 for all but four valves tested under cold water pumped‐flow conditions.

The "apparent friction coefficients" for the remaining four valves ranged from

0.66 to 1.93. EPRI results demonstrated "apparent friction coefficients"

ranging from 0.34 to 0.41 for hot water pumped‐flow conditions, 0.35 to 0.8

for hot water blowdown conditions, and 0.25 to 0.64 for steam blowdown

conditions. EPRI's "apparent friction coefficients" reflect all valve

internal phenomena and are not necessarily indicative of a "sliding friction

coefficient." The major difference between the "apparent friction

coefficient" used by EPRI and the "valve factor" used historically by valve

vendors in sizing motor operators is the consideration of the valve disc angle

in determining the EPRI "apparent friction coefficient."

Most valve vendors have used a "valve factor" of 0.3 for flexible wedge gate

valves and 0.2 for parallel disc gate valves in sizing motor operators.

Therefore, the EPRI test results indicate that the thrust required to operate
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gate valves could be significantly greater than the thrust predicted by the

valve vendors. The EPRI blowdown test results are generally consistent with

those obtained in the limited testing program conducted by the Idaho National

Engineering Laboratory (INEL) for the NRC Office of Nuclear Regulatory

Research in 1989.

EPRI reported that the valve sliding friction coefficient tends to decrease

with increasing differential pressure which lends support for linear

extrapolation of reduced differential pressure results when there is a low

potential for valve damage (for example, under nominal flow velocity

pumped‐flow conditions).

EPRI reported that several gate valves were damaged during hot water and steam

blowdown testing. These included a 6‐inch Anchor‐Darling valve (disk and seat

damage); a 6‐inch Crane valve (guide damage); a 10‐inch Velan valve (guide

damage); a 6‐inch Walworth valve (guide damage); and a 10‐inch Edward valve

(disk and seat damage). Two of the damaged valves exhibited "apparent

friction coefficients" exceeding 0.6.

Two gate valves were damaged under cold water pumped‐flow conditions. These

included a Velan 6‐inch valve (plastic bending of body guides at high flow

velocity greater than 30 feet per second) and an 18‐inch Anchor‐Darling valve

(valve disk forced through seating area resulting in leakage above disk).

EPRI test results revealed that it is generally not possible to determine

accurately the point of flow isolation prior to disk wedging based on the

thrust diagnostic trace alone.

EPRI stated that it had not observed differences in thrust requirements for

valve operation between valves installed in horizontal pipes with the stem

either vertical or horizontal. This finding differs from some operating

experiences in nuclear power plants.

2.Globe Valves

EPRI stated that, for incompressible flow conditions, globe valve thrusts are

consistent with industry calculational‐method predictions only if the

appropriate area is chosen for differential pressure application. The

appropriate area (disk mean seat area versus disk guide area) appears to be

unique to valve design. It was determined that use of disk mean seat area

rather than disk guide area can result in significant underestimation of

required thrust for some globe valve designs. Specifically, one globe valve

tested under cold water pumped‐flow conditions required approximately twice as

much thrust to close using disk mean seat area and a valve factor of 1.0.

A two‐inch Rockwell/Edward globe valve, tested under hot water blowdown

conditions, exhibited thrust requirements exceeding predictions based on disk

guide area by approximately 35 percent. This valve sustained damage to the

portion of the body bore that guides the disk.

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Current industry practice for determining the required thrust for globe valves

varies by manufacturer. Many manufacturers assume disk mean seat area

multiplied by a valve factor in the 1.0 to 1.1 range. Others use disk guide

area in making thrust predictions. Therefore, the EPRI results indicate that

actual thrust requirements may exceed those predicted using current industry

practice for some globe valve designs.

3.Butterfly Valves

EPRI stated that the Wyle flow loop testing revealed torque requirements to

operate Pratt butterfly valves which were bounded by the most current torque

predictions of the manufacturer. However, butterfly valves at some nuclear

power plants (for example, Catawba and Palo Verde) have demonstrated torque

requirements that exceed vendor predictions. EPRI is currently evaluating

data from testing conducted at Kalsi Engineering to assess butterfly valve

torque requirements for a wide range of butterfly valve designs.

4.Data Interpretation and Assessment

In July 1993, EPRI sent a Quarterly Status Report to all utilities

participating in the EPRI MOV Performance Prediction Program. This report

summarized preliminary flow loop test results. After the completion of

Wyle/Siemens quality assurance checks, EPRI plans to update this information

in its next Quarterly Status report scheduled for late 1993. Detailed test

reports documenting these results are scheduled for delivery to participating

utilities between October and December 1993. EPRI stated that, in

interpreting the EPRI flow loop test results, utilities need to understand the

assumptions and equations that were used by EPRI in presenting the data. For

example, the EPRI calculated "apparent friction coefficient" for gate valves

is based on the equation provided in EPRI Report NP‐6660‐D, "Application Guide

for Motor‐Operated Valves in Nuclear Power Plants." This equation is solved

for "apparent friction coefficient" using (1) the maximum measured stem thrust

which occurs prior to the initiation of wedging (for valve closing) or the

maximum thrust which occurs after cracking (for valve opening); (2) full

(valve closed) tested differential pressure; (3) mean seat area; (4) valve

disk angle; (5) full (valve closed) upstream tested pressure for stem

rejection thrust; and (6) measured values of packing load.

EPRI stated that valve design and test conditions, maintenance history, and

operating experience may be important in assessing the applicability of EPRI

test results to plant MOVs.

EPRI uses the greatest thrust requirement to overcome differential pressure

and flow to determine its "apparent friction coefficient." EPRI assumes the

highest differential pressure observed during the test regardless of the stem

position where the greatest differential pressure/flow required thrust occurs.

This results in a lower calculated friction coefficient than would be

determined if the actual differential pressure at the point of greatest thrust

was used in determining the friction coefficient.

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EPRI plans to submit sections of a topical report for NRC review as they are

completed between November 1993 and April 1994. EPRI intends to submit

supporting reports in advance of the final topical report to allow the staff

to raise questions with EPRI early in the review process.

EPRI stated that its final methodology is scheduled for delivery to utilities

in April 1994 as a tool that may be used to confirm many aspects of MOV

calculations and setup. Further, the EPRI flow loop test results provide
licensees with information which might be helpful in supplementing other "best

available" data in establishing MOV switch settings.

Discussion

Since EPRI initiated its MOV Performance Prediction Program, the NRC staff has

conducted public meetings with NUMARC and EPRI to discuss the goals of the

EPRI program, the development of the program activities to accomplish those

goals, the tests conducted in support of the program and the results of those

tests, and the completion schedule for the program. The staff has provided

questions and comments to NUMARC and EPRI on the EPRI MOV program as a result

of these meetings. For example, in a public meeting on October 6‐7, 1993, the

staff emphasized the need for EPRI to ensure that licensees clearly understand

the application of the EPRI test data and methodology. Also at this meeting,

contents of this notice were discussed and the comments from EPRI have been

considered. The staff expressed concern about the valves damaged during the

EPRI testing and the apparent lack of action by some valve manufacturers in

response to the valve damage. The staff also discussed the need for EPRI to

ensure that adequate peer review of the EPRI methodology is conducted.

Although some issues remain to be resolved, the EPRI testing program should

provide a significant amount of MOV test data that can assist nuclear power

plant licensees in demonstrating the design‐basis capability of MOVs that

cannot be tested under dynamic conditions as installed. The preliminary test

information provided in this notice is provided for licensee consideration in

implementing programs in response to GL 89‐10 . The staff plans to conduct

additional public meetings with NUMARC and EPRI to discuss the status of the

EPRI MOV program. The staff will consider the need for additional generic

communications to nuclear power plant licensees and construction permit

holders as additional information is obtained from the EPRI MOV program.

Related Generic Communications

NRC Information Notice 90‐40, "Results of NRC‐Sponsored Testing of

Motor‐Operated Valves."

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This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

the person listed below or the appropriate Office of Nuclear Reactor

Regulation (NRR) project manager.

/s/'d by BKGrimes

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contact: Thomas G. Scarbrough, NRR

(301) 504‐2794

Attachment:

List of Recently Issued NRC Information Notices

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