Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
October 25, 1990
TO: ALL LICENSEES OF OPERATING NUCLEAR POWER PLANTS AND HOLDERS OF
CONSTRUCTION PERMITS FOR NUCLEAR POWER PLANTS
SUBJECT: GENERIC LETTER 89-10, SUPPLEMENT 3, "CONSIDERATION OF THE RESULTS
OF NRC-SPONSORED TESTS OF MOTOR-OPERATED VALVES"
BACKGROUND
In Generic Letter 89-10 (June 28, 1989), "Safety-Related Motor-Operated
Valve Testing and Surveillance," the staff of the U.S. Nuclear Regulatory
Commission (NRC) requested holders of operating licenses and construction
permits to establish a program to provide for the testing, inspection, and
maintenance of safety-related motor-operated valves (MOVs) and certain other
MOVs in safety-related systems. Supplement 1 to Generic Letter 89-10 (June
13, 1990) provides the results of public workshops held to discuss the
generic letter and to answer questions on the staff positions regarding its
implementation. In Supplement 2 (August 3, 1990) the NRC staff stated that
inspections of program descriptions would not commence until January 1,
1991, and, thus, the program descriptions need not be available on site
until that date.
In parallel with the NRC staff's activities leading to Generic Letter 89-10,
the staff performed tests of MOVs as part of an ongoing research effort.
The tests were conducted on 6-inch and 10-inch gate valves typically used to
provide containment isolation in the steam supply lines of the High Pressure
Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems,
and in the supply line for the Reactor Water Cleanup (RWCU) system at
boiling water reactor (BWR) nuclear power plants. On June 5, 1990, the
staff issued Information Notice 90-40, "Results of NRC-Sponsored Testing of
Motor-Operated Valves."
As discussed in Information Notice 90-40, the NRC-sponsored tests revealed
that, regardless of fluid conditions, 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 using typical friction factors. Thus, although the
NRC-sponsored tests focused on the HPCI, RCIC and RWCU containment isolation
valves at BWR plants, the information obtained from those tests may be
applicable to valves used in other systems at BWR and pressurized water
reactor (PWR) plants. For example, calculations using low valve friction
factors may underestimate thrust requirements for opening and closing
valves.
9010220146
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In response to a staff request, the BWR Owners' Group obtained information
from the BWR licensees regarding the capability of MOVs used to provide
containment isolation in the steam lines of the HPCI and RCIC systems, and
in the supply line of the RWCU system. The staff's review of the
NRC-sponsored test results and the MOV data provided by the BWR Owners'
Group indicates that deficiencies might exist in those MOVs.
DISCUSSION
In Generic Letter 89-10, the NRC staff requested that licensees and
construction permit holders complete the programs established in response to
the generic letter (excluding the periodic verification of MOV switch
settings) by June 28, 1994, or within 3 refueling outages after December 28,
1989 (or operating license issuance for construction permit holders),
whichever is later. While recommending that licensees and permit holders
consider the safety significance of MOVs in developing their programs, the
staff did not have sufficient information at that time to recommend that
licensees and permit holders establish any particular priority for MOVs
within the generic letter program. The information recently obtained from
the NRC-sponsored tests, however, may affect the priorities being
established by licensees and permit holders for implementing their generic
letter programs. From its evaluation of the MOV data provided by the BWR
Owners' Group and the results of the NRC-sponsored tests, the staff has
determined that correction of any deficiencies in the HPCI, RCIC and RWCU
MOVs described herein need to be given high priority in the implementation
of generic letter programs. While such deficiencies may not need to be
corrected immediately, the staff has determined by means of a safety
assessment (Enclosure 1) that any MOV deficiencies should be corrected
within 18 months or by the end of the first refueling outage, following
issuance of this generic letter supplement, whichever is later. The staff's
review of a generic safety assessment performed by the BWR Owners' Group
(Enclosure 2) confirmed that this time period is acceptable for correcting
any deficiencies in those MOVs. If a BWR licensee believes that there are
MOVs with potential deficiencies at its facility that have greater safety
significance than the HPCI, RCIC, and RWCU MOVs described herein, the
licensee should determine the appropriate priority for completing the
generic letter program for those valves.
REQUESTED ACTIONS
BWR licensees are requested to assess the applicability of the data from the
NRC-sponsored MOV tests, to determine the "as-is" capability of the HPCI,
RCIC, and RWCU MOVs described herein, and to identify any deficiencies in
those MOVs. Where applicable, BWR licensees should also evaluate the MOVs
used for containment isolation in lines to the isolation condensers.
Elements that a BWR licensee may consider in determining whether the
NRC-sponsored test data are inapplicable to its HPCI, RCIC and RWCU MOVs
include valve size, type and manufacturer; disk type; design-basis
differential pressure and flow conditions; internal dimensions and
clearances; and disk and guide surface materials.
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BWR licensees are requested to perform a plant-specific safety assessment to
verify that the generic safety assessments performed by the NRC staff and
the BWR Owners' Group are applicable. In performing the plant-specific
safety assessment, BWR licensees should address factors such as
consideration of functional valve test results; operating procedures and
emergency operating procedures; the conduct of training; current torque
switch bypass settings including the potential for motor overload on a first
attempt to close the valve; leak detection capabilities; inspection programs
for erosion-corrosion and intergranular stress corrosion cracking (including
response to Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic
Stainless Steel Piping"); water-hammer prevention practices; the
environmental qualification of the MOVs and other nearby equipment;
radiological consequences both on and off the plant site that could result
from a pipe leak or break; and probabilistic risk considerations. Where
applicable, BWR licensees should include in their plant-specific safety
assessments MOVs used for containment isolation in lines to the isolation
condensers. If a BWR licensee believes that there are MOVs with potential
deficiencies at its facility that have greater safety significance than the
HPCI, RCIC and RWCU MOVs (and the MOVs in the isolation condenser lines)
described herein, the licensee should justify as part of its plant-specific
safety assessment the prioritization of its effort to identify and correct
MOV deficiencies.
BWR licensees should consider the implementation of short-term corrective
actions. For example, BWR licensees should evaluate the feasibility of
increasing torque switch settings where the motor, actuator, and valve are
designed to accommodate such an increase. BWR licensees should develop
procedures and provide training for plant personnel to respond to a pipe
leak or break in a line containing a deficient MOV, particularly if the
deficiency cannot be corrected in the short term.
BWR licensees may accomplish these recommendations as part of an accelerated
response to Generic Letter 89-10 for the applicable MOVs. For example, BWR
licensees could complete the design-basis reviews for those MOVs and could
establish torque switch settings as described in Recommended Actions a and b
of the generic letter, respectively. Recommended Action c of the generic
letter requests that the MOVs be tested in situ under design-basis
differential pressure and flow conditions, where practicable. For those
instances where design-basis testing in situ is not practicable and an
alternative to such testing cannot be justified at this time, the staff
recommends that the BWR licensee use the "two-stage" approach discussed in
Generic Letter 89-10 and Supplement 1. Following that approach, the BWR
licensee would determine the operating requirements of the MOV using the
best data currently available and then obtain applicable data as soon as
possible.
While the reporting requirements below are addressed to BWR licensees, all
licensees and construction permit holders should consider the applicability
of the information obtained from the MOV tests and the staff evaluation of
the
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test results to other MOVs within the scope of Generic Letter 89-10. In
addition, all licensees and permit holders should consider this information
in the development of priorities for implementing the generic letter
program.
REPORTING REQUIREMENTS
In order for the NRC to determine whether any BWR operating licenses should
be modified, suspended or revoked, BWR licensees shall provide written
information, signed under oath or affirmation pursuant to Section 182 of the
Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f), as follows:
1. Within 30 days of the receipt of this letter, BWR licensees shall
notify the NRC staff that a plant-specific safety assessment report
addressing, as a minimum, the factors described herein, is available on
site for staff review. BWR licensees shall also notify the NRC staff
whether they believe that there are MOVs with deficiencies of greater
safety significance than the MOVs used to provide containment isolation
in the steam supply lines of the HPCI and RCIC systems, in the supply
line of the RWCU system, and in the line to the isolation condenser.
2. Within 120 days of the receipt of this letter, BWR licensees shall
provide to the NRC staff the following:
a. Criteria, reflecting operating experience and the latest test data,
that were applied in determining whether deficiencies exist in the
HPCI, RCIC and RWCU MOVs described herein, in the MOVs in isolation
condenser lines, and in any MOVs considered to be more safety
significant, as applicable;
b. The identification of any MOVs found to have deficiencies; and
c. A schedule for any necessary corrective action.
3. Subsequent to the determination of necessary corrective actions or the
establishment of the schedule for completion of those actions, BWR
licensees shall inform the NRC staff of any changes to the planned
actions or schedule.
As noted above, based on the generic safety assessments prepared by the NRC
staff and the BWR Owners' Group, the staff believes that justification
exists for individual plants to which those safety assessments are
applicable to take 18 months or to the end of the first refueling outage,
following issuance of this generic letter supplement, whichever is later, to
resolve any deficiencies in the HPCI, RCIC and RWCU MOVs described herein.
However, if a BWR licensee determines that a more limited time is mandated
by its plant-specific safety assessment, the licensee should utilize the
more restrictive time. If additional time is needed to complete the
corrective actions, BWR licensees should submit the plant-specific safety
assessment and obtain staff approval for the corrective action schedule.
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BACKFIT DISCUSSION
Based on operating experience and research results, the staff determined
several years ago that MOV tests beyond those previously acceptable are
necessary to satisfy the NRC regulations. As that determination constituted
a backfit, the staff prepared Generic Letter 89-10 in accordance with NRC
procedures for the issuance of staff guidance containing backfit provisions.
Supplement 3 represents a further backfit in that the staff is requesting
BWR licensees to advance the schedule for Generic Letter 89-10 with respect
to specific MOVs at BWR plants. This limited advancement of the Generic
Letter 89-10 schedule is the result of the information obtained from
NRC-sponsored MOV tests indicating that deficiencies might exist in certain
MOVs installed to perform containment isolation functions at BWR plants.
The staff has determined that the issuance of Supplement 3 to Generic Letter
89-10 is necessary to provide confidence that BWR facilities are in
compliance with their safety analyses and NRC regulations such as described
in 10 CFR Part 50, Appendix A, Criteria 54 and 55. More specifically,
because deficiencies might exist in the MOVs described herein, the staff
does not have adequate confidence that (1) as required by Criterion 54, the
applicable piping systems which penetrate containment have been provided
with leak detection, isolation, and containment capabilities having
redundancy, reliability, and performance capabilities which reflect the
importance to safety of isolating these piping systems, or have been
designed with the capability to test periodically the operability of the
isolation valves and associated apparatus or (2) as required by Criterion
55, appropriate requirements, such as higher quality in design, fabrication,
and testing, to minimize the probability or consequences of an accidental
rupture of lines which are part of the reactor coolant pressure boundary and
penetrate reactor containment have been provided as necessary to assure
adequate safety. Therefore, the staff has determined that the backfit
provisions of this generic letter supplement are justified under 10 CFR
50.109 (a)(4)(i). Based on its safety assessment, the staff determined that
no immediate corrective actions are needed and that BWR licensees may
proceed to resolve any deficiencies in the MOVs described herein as
recommended in this letter.
This request is covered by Office of Management and Budget Clearance Number
3150-0011, which expires December 31, 1991. The estimated average burden
hours are 150 person-hours per licensee response, including assessment of
the new recommendations, searching data sources, gathering and analyzing the
data, performing data evaluations, and preparing the required letters.
(These estimated average burden hours pertain only to the identified
response-related matters and do not include the time for actual
implementation of the requested action.) Send comments regarding this
burden estimate or any other aspect of this collection of information,
including suggestions for reducing
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this burden, to the Information and Records Management Branch, Division of
Information Support Services, Office of Information Management, U.S. Nuclear
Regulatory Commission, Washington, D.C., 20555; and to the Paperwork
Reduction Project (3150-0011), Office of Management and Budget, Washington,
D.C. 20503.
James G. Partlow
Associate Director for Projects
Office of Nuclear Reactor Regulation
Enclosures: As stated
TECHNICAL CONTACT:
Thomas G. Scarbrough (301) 492-0794
LEAD PROJECT MANAGER:
Anthony T. Gody, Jr. (301) 492-1387
