Information Notice 1990-40, Results of NRC-Sponsored Testing of Motor-Operated Valves
It9
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C.
20555
June 5, 1990
NRC INFORMATION NOTICE NO. 90-40:
RESULTS OF NRC-SPONSORED TESTING
OF MOTOR-OPERATED VALVES
Addressees
All holders of operating licenses or construction permits for nuclear power
reactors.
Purpose
This information notice is intended to provide addressees with specific infor- mation regarding the results of recent NRC-sponsored testing of motor-operated
valves (MOVs) which was discussed at a public meeting on April 18, 1990. It is
expected that recipients will review the information for applicability to their
facilities and consider actions, as appropriate, to avoid problems with
safety-related MOVs.
However,.suggestions contained in this information notice
do not constitute NRC requirements; therefore, no specific action or written
response is required.
Background
The NRC Office of Nuclear Regulatory Research (RES) has been sponsoring an MOV
testing program in support of the resolution of Generic Safety Issue 87 (GI-87),-"Failure of HPCI Steam Line Without Isolation." The initial scope of
GI-87 involved the evaluation of the capability of certain motor-operated
flexible wedge gate containment isolation valves to mitigate the loss of
reactor coolant inventory in the event of a pipe break outside of the contain- ment building at boiling-water-reactor (BWR) plants. The particular MOVs
involved in the GI-87 program were those in the turbine steam supply lines for
the high pressure coolant injection (HPCI)
and reactor core isolation cooling
(RCIC) systems, and in the supply line to the reactor water cleanup (RWCU)
system.
The MOV research is applicable to the programs established by licensees in
response to Generic Letter 89-10, "Safety-Related Motor-Operated Valve Testing
and Surveillance." In that generic letter, the staff recommended that
licensees and construction permit holders establish a program to provide for
the testing, inspection, and maintenance of safety-related MOVs and certain
other MOVs in safety-related systems.
The purpose of this program is to
provide assurance that the MOVs will function when subjected to design-basis
differential pressure and flow conditions.
As part of the generic letter
program, the staff recommended that licensees and permit holders test the MOVs
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9005290270 Z 9 E
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June 5, 1990 within the program in situ under design-basis conditions, where practicable.
The schedule in the generic letter requested that the description of the MNOV
program be available within about a year of issuance of the generic letter and
that the initial test program be completed in approximately five years.
As a
followup to the initial program, the staff recommended that the MOV switch
settings and, thus, operability of the MOVs be reverified periodically.
Although the generic letter has a five-year schedule for completing the initial
program, the staff indicated at the public workshops held to discuss the
generic letter that the NRC regulations require that licensees act to resolve
operability problems with specific MOVs when the problems are identified.
As
part of its review of the research results, the staff will consider the need to
accelerate a portion or all of the Generic Letter 89-10 program for particular
MOVs or systems.
In Generic Letter 89-10, the staff acknowledges that in situ testing of some
MOVs within the generic letter program under design-basis conditions will not
be practicable.
At the generic letter workshops, the staff discussed several
possible alternatives if such testing is not practicable, as well as potential
problems and limitations associated with those alternatives.
For instances in
which testing of an MOV in situ under design-basis conditions is not practica- ble and the licensee cannot currently justify the use of an alternative to
design-basis testing in situ, the staff has recommended the use of a "two- stage" approach:
the licensee would set the MOV operating switches by means of
the best data available and then would work to obtain applicable test data as
soon as possible.
The staff believes that applicable test data can be obtained
within the five-year schedule.
For the initial setting of the MOV switches
under the two-stage approach, the test results obtained through the.NRC
research may constitute some of the best data available for the tested valves
under a variety of fluid conditions.
Description of Circumstances
The MOV testing program for GI-87 has been conducted in two phases by the Idaho
National Engineering Laboratory (INEL).
Phase I was performed in 1988 at the
Wyle Laboratory facility in Huntsville, Alabama.
The most significant tests in
that phase consisted of opening and closing two 6-inch flexible wedge gate
valves (manufactured by Anchor/Darling and Velan) under high differential
pressure and high-temperature water conditions. The valves in Phase I of the
research program were considered typical of those used for containment isola- tion in the supply line to the RWCU system.
The results of the tested valves
were discussed at a public meeting on February 1, 1989, and are documented in
NUREG/CR-5406, "BWR Reactor Water Cleanup System Flexible Wedge Gate Isolation
Valve Qualification and High Energy Flow Interruption Test."
Phase II of the MOV test program was performed in 1989 at the Kraftwerk Union
facility in the Federal Republic of Germany.
This phase consisted of opening
and closing three 6-inch flexible wedge gate valves (Anchor/Darling, Velan, and
Walworth) and three 10-inch flexible wedge gate valves (Anchor/Darling, Powell, and Velan) against normal and blowdown (design-basis) flow conditions.
The
Phase II 6-inch and 10-inch valves were considered typical of those used for
containment isolation in the supply line to the RWCU system and the turbine
June 5, 1990 steam supply line of theAPCI systems, respectively. *'On
December 26, 1989, the
NRC staff issued Information Notice 89-88, "Recent NRC-Sponsored Testing of
Motor-Operated Valves," which alerted addressees to the tests and provided some
preliminary results.
On April 18, 1990, the NRC staff held a public meeting to
discuss-the results of Phase II of the MOV testing program.
The test data are
available in printed form in the NRC Public Document Room (Accession No.
9005170154).
Magnetic tapes of the test data are available through the INEL
Office of Technology Transfer.
The overall objectives of the MOV test program included the determination of
the force required to close the tested valves under various operating and
design-basis fluid conditions through the measurement of stem thrust.
Other
program objectives were the determination of opening thrust requirements for
the tested valves under different fluid conditions; evaluation of valve closure
force components (such as disc friction and packing drag); measurement of the
effects of temperature, pressure, and valve design on valve opening and closing
loads; and evaluation of the valve thrust equation commonly used in the
industry.-
The tests for each MOV included cold leakage, cold and hot cycling, opening and
closing under normal flow, closure under design-basis, and partial opening and
closing under high differential pressure and flow conditions.
Although the
tested valves were intended to be typical of those used for containment isola- tion in the HPCI and RWCU systems of BWR plants, the results of the tests
should be considered in terms of their applicability to all MOVs in nuclear
power plants. A detailed analysis of the test data should be available in July
1990.
Nevertheless, the NRC staff has begun to develop conclusions from the
test data as a result of its review of the data and the discussions at the
April 18, 1990, public meeting.
Several preliminary conclusions are discussed
below:
1.
Regardless of fluid conditions (i.e., steam, slightly subcooled water, or
cold water), the tested valves required more thrust for opening and
closing under various differential pressure and flow conditions than would
have been predicted from standard industry calculations and typical
friction factors.
Thus, a potential exists for the underestimation of
thrust requirements for valves in applications, and under fluid condi- tions, other than those of the valves involved in the NRC research.
For
the conduct of the tests, the motor operators for the valves were sized, and the torque switches were set,Ain an effort to ensure that each valve
would fully stroke without regard to the thrust requirements predicted by
the commonly used valve thrust equation.
(Despite this effort, one valve
failed to close completely during a'blowdown test.) To provide an indica- tion of the accuracy of the valve thrust equation, the thrust predicted by
that equation for valve friction factors of both 0.3 and 0.5 was calculat- ed during each test.
Table 1 provides a summary of the blowdown tests and
the minimum required thrust to close the tested valves.
The table also
indicates whether the valve thrust equation would have bounded the thrust
requirement if valve friction factors of 0.3 or 0.5 had been used.
June 5, 1990 2.
Some of the tested valves sustained considerable internal damage during
the blowdown tests.
The occurrence of internal damage can cause the
thrust required to operate a valve to exceed the thrust requirements
predicted by the valve thrust equation.
Such valves were referred to as
"unpredictable" in the test program and included the 6-inch Anchor/Darling
valve and the 10-inch Anchor/Darling, Powell, and Velan valves.
In some
instances, this increase in required thrust can be considerable and might
exceed the capability of the motor or operator.
Thrust requirements to
close unpredictable valves under design-basis loads cannot be accurately
determined without testing the valves (either individually or as proto- types) under those conditions.
3. The research program revealed that the testing of a valve under static or
low flow conditions cannot always be used to accurately predict the
behavior of the valve under design-basis conditions by extrapolation.
For
example, the valves that were damaged during blowdown tests operated
normally under less severe flow tests.
Thus, low-flow tests might not
identify a valve that requires significantly more thrust than predicted by
the valve thrust equation (i.e., a valve that is unpredictable).
4.
During opening of the valves, the maximum required thrust did not always
occur at unseating.
Rather, in certain instances, it occurred much later
during the valve stroke.
At nuclear plants, the staff has found that
torque switches for MOVs are sometimes bypassed only during the initial
portion of the opening stroke on the assumption that the thrust required
to unseat the valve would be the maximum thrust for the full stroke.
Thus, the research results raise a concern that the torque switches in
some MOVs at nuclear plants might not be bypassed for a sufficient period
of time during the opening stroke.
5.
For certain tests, the valve was closed from a partially open position.
This partial stroking of the valve failed to predict the thrust require- ments and to identify nonpredictable performance that were found during
closure of the valve from a full open position.
For example, during
certain blowdown tests, valve damage began to occur before the valve was
half closed.
The accumulated damage over the full stroke influences the
thrust required to close the valve.
6. The research program revealed that measurements of torque, thrust, and
motor operating data were needed to completely characterize MOV perfor- mance.
For example, the measurement of torque or thrust alone cannot
identify problems in the conversion of torque to thrust (i.e., abnormally
large stem factors).
Such problems can cause the thrust measured at
normal or static conditions to be misleading as compared to the thrust
that actually would be available under design-basis conditions.
The
measurement of motor operating characteristics allows the adequacy of the
motor to be determined.
7. The research program revealed that reliable information can be obtained
from diagnostic analysis of MOVs only when operating data are collected by
trained personnel using accurate and calibrated equipment.
The MOV data
must then be evaluated by individuals experienced in the performance of
MOV diagnostic analysis.
June 5, 1990 The staff is continuing its review of the results of the MOV research.
From
this review, the staff may prepare additional information notices that discuss
the staff's conclusions regarding the research.
If an immediate safety problem
is identified, the staff will initiate regulatory action to ensure the MOVs
will perform their safety functions.
This information notice requires
you have any questions about the
of the technical contacts listed
no specific action or written response. If
information in this notice, please contact one
below or the appropriate NRR project manager.
ar es .
s i, Vrco
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contacts:
Thomas G. Scarbrough, NRR
(301) 492-0916 la
Richard J. Kiessel, NRR
(301) 492-1154 Attachments:
1. Table 1 - GI-87 Research Results for Blowdown Tests
2. List of Recently Issued NRC Information Notices
.
Attachment 1
June 5, 1990 TABLE 1
GI-87 RESEARCH RESULTS FOR BLOWDOWN TESTS
Manufacturer
SIX-INCH VALVES
Anchor/Darling
(Phase 1)
D/P (psi)
990
T (OF)
524 Fluid
Hot water
Required
Thrust (lbs)
20,000
NOTES
(1)(2)
(1)(2)(3)
Anchor/Darling
(Phase 2)
900
520
Hot water
>23,000
Velan (Phase 1)
Velan (Phase 2)
990
950
1040
750
600
1000
1300
524
520
550
<100
540
470
520
Hot water
Hot water
Steam
Cold water
Hot water
Hot water
Hot water
15,000
14,000
14,000
13,000
9,000
14,000
16,000
(4)
(2)
(4)
(2)
(2)
(2)
(4)
Walworth
TEN-INCH VALVES
Anchor/Darling
Powell
920
1100
1300
750
800
1040
520
550
570
510
525
550
Hot water
Hot water
Hot water
Steam
Steam
Steam
9,000
12,000
15,000
29,000
28,000
29,000
(4)(5)(6)
(4)(5)(6)
(4)(5)(6)
(1)(2)
(1)(4)
(1)(4)
Velan
990
1400
1100
550
590
560
Steam
Steam
Steam
33,000
40,000
36,000
(1)(2)
(1)(2)
(1)(2)
NOTES:
1.
Valve damage during stroke could result in higher thrust requirements than
predicted by the valve thrust equation.
(These valves are referred to as
"unpredictable").
2.
The valve thrust equation with valve friction factors of either 0.3 or 0.5 did not bound the required thrust in the blowdown test.
3.
The torque switch tripped before full valve closure.
4.
The valve thrust equation with a valve friction factor of 0.3 did not
bound the required thrust in the blowdown test, but the equation did bound
the required thrust if a valve friction factor of 0.5 was used.
5. This valve had a removable guide which deformed during the blowdown test.
6.
In determining whether the MOV can accommodate the required thrust to
close the valve, the weak link among the motor, operator, and valve must
be identified.
For the Walworth valve, this is especially important
because stems with relatively small diameters are typically used in these
valves.
Attachment 2
June 5, 1990 LIST OF RECENTLY ISSUED
NRC INFORMATION NOTICES
Information
Date of
Notice No.
Subject
Issuance
Issued to
90-39
90-38
90-37
90-36
90-35 Recent Problems With Service 6/1/90
Water Systems
Requirements for Processing
5/29/90
Financial Assurance Submittals
for Decommissioning
Sheared Pinion Gear-to-Shaft 5/24/90
Keys in Limitorque Motor
Actuators
Apparent Falsification of
5/24/90
State of Connecticut
Weight Certificates
Transportation of Type A
5/24/90
Quantities of Non-Fissile
Radioactive Materials
Response to False Siren
5/10/90
Activations
Sources of Unexpected
5/9/90
Occupational Radiation
Exposures at Spent Fuel
Pools
Surface Crack and Subsurface 5/3/90
Indications in the Weld of
A Reactor Vessel Head
Update on Waste Form and
5/4/90
High Integrity Container
Topical Report Review
Status, Identification
of Problems with Cement
Solidification, and
Reporting of Waste Mishaps
All holders of OLs
or CPs for nuclear
power reactors.
All fuel facility
and materials
licensees.
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
or CPs for nuclear
power reactors, and
10 CFR 70 licensees.
All U.S. NRC licensees.
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
or CPs for nuclear
power reactors.
All holders of OLs
or CPs for nuclear
power reactors, fuel
cycle licenses, and
certain by-product
materials licenses.
90-34
90-33
90-32
90-31 OL = Operating License
CP = Construction Permit
June 5, 1990
Page 5 of S
The staff is continuing its review of the results of the NOV research.
From
this review, the staff may prepare additional information notices that discuss
the staff's conclusions regarding the research. If an Immediate safety problem
is identified, the staff will initiate regulatory action to ensure the MDVs
will perform their safety functions.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact one
of the technical contacts listed below or the appropriate NRR project manager.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contacts:
Thomas G. Scarbrough, NRR
(301) 492-0916
Richard J. Kiessel, NRR
(301) 492-1154 Attachments:
1. Table 1 - GI-87 Research Results for Blowdown Tests
2. List of Recently Issued NRC Information Notices
Document
- SEE PREVIOUS
- EMEB:DET:NRR
TGScarbrough
05/25/90
Name: IN 90-40
CONCURRENCES
- C/EMEB:DET:NRR*D/DET:NRR
LBMarsh
JERichardson
05/25/90
05/25/90
D/
- OG :DOEA:NRR
RJKiessel
05/25/90
- C/OGCB:DOEA:NRR
CHBerlinger
05/25/90
- RPB:ADM
TechEd
05/25/90
IN 90-xx
June xx, 1990
Page 5 of
The staff is continuing its review of the results of the MOV research. From
this review, the staff might prepare additional Information notices that
discuss the staff's conclusions regarding the research. If an immediate
safety problem is identified, the staff will initiate regulatory action to
ensure the MOYs will perform their safety functions.
This information notice requires no specific action or written response. If
you have any questions about the information in this notice, please contact one
of the technical contacts listed below or the appropriate NRR project manager.
Charles E. Rossi, Director
Division of Operational Events Assessment
Office of Nuclear Reactor Regulation
Technical Contacts: Thomas G. Scarbrough, NRR
(301) 492-0916
Richard J. Kiessel, NRR
(301) 492-1154 Attachments:
1. Table 1 - GI-87 Research Results for Blowdown Tests
2. List of Recently Issued NRC Information Notices
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