Operability Testing of Relief and Safety Relief Valves (Generic Letter 79-27)
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D. C. 20555
July 16, 1979
ALL BOILING WATER REACTOR LICENSEES
(Except Big Rock Point and Humboldt Bay)
In December 1977, we sent letters to the majority of licensees who operate
Boiling Water Reactors (BWR) regarding the relief and safety-relief that are
installed in the reactor coolant system and/or the automatic
depressurization system. The letters requested licensees to propose
Technical Specification changes to incorporate additional Surveillance
Requirements for these valves. Model Technical Specifications were included
for guidance in preparing plant specific requirements. The principal feature
of the new requirements was a variable frequency test schedule for
operability testing of relief and safety-relief valves.
Some licensees objected to this feature on the basis that increased testing
of the Target Rock safety-relief valves could significantly degrade valve
reliability because such testing could aggravate pilot valve leakage thereby
increasing the likelihood of future malfunctions. This objection was not
voiced by all licensees and was not shared by the NRC staff at the time.
However, we did believe that further consideration of this view, within the
context of overall reactor safety, was warranted.
We have since made an independent study of BWR pressure relief system
failures. The results of the study have been published in NUREG-0462,
"Technical Report on Operating Experience with BWR Pressure Relief Systems",
dated July 1978. A copy is enclosed for your convenience. Based on the
findings of this report and further information obtained from General
Electric, in our meeting of March 30, 1979, we have concluded that
implementation of a requirement for increased surveillance testing would not
be the most effective way of assuring safety-relief valve reliability.
Consequently, unless you supply information to the contrary, we do not plan
to act on any proposed Technical Specification changes you may have
submitted in response to our December 1977 request. However, due to the
potential effects of safety-relief valve malfunctions, the NRC staff
continues to believe that licensees should make all reasonable efforts to
increase the reliability of these valves and to reduce the frequency of
inadvertent actuation and subsequent failure to reseat properly. This
general matter is further discussed in the NUREG-0560 staff report
concerning the Three Mile Island Unit 2 accident, wherein the pressurizer
power operated relief valve failed to close during a feedwater transient and
resulted in a small break LOCA. Staff review of other operating events
indicates a significant frequency of such valve failures leading to small
break LOCA events. Accordingly, reliability goals are currently being
developed by the staff for safety and relief valves which are part of the
reactor, coolant pressure boundary, consistent with the recommendations of
- 2 - July 16, 1979
General Electric has made recommendations to licensees that we believe would
substantially reduce the likelihood of future failures of BWR safety-relief
valves. General Electric's recommendations consist of a short term and a
long term program. Basically, the short term program consists of an
intensified maintenance program and minor modifications to the valve
assembly which will enable the simmer margin of the valve to be increased to
about 120 psi. General Electric has provided operating data that indicate
the malfunction of valves having a simmer margin of about 100 psi is
appreciably less than those with smaller simmer margins. The long term
program consists of replacing the original three stage pilot operated
actuator with a redesigned two stage pilot operated actuator. It is our
understanding that the newly designed actuator has, by tests, demonstrated
improved reliability due to the elimination of the bellows and its reduced
sensitivity to pilot valve leakage.
The NRC plans to continue to monitor the performance of safety/relief
valves, and the status and effectiveness of actions to improve their
reliability over the long term. To apprise us of the current situation at
your plant(s), as well as your plans for future actions, we request that you
provide responses to the items identified in the enclosure within 60 days.
If your plant design does not utilize Target Rock safety/relief valves, so
indicate within 60 days; and disregard the enclosure.
If you have any questions, or care to discuss this matter, please contact
This request for generic information was approved by GAO under clearance
number B-180225 (S79014); this clearance expires June 30, 1980
Brian K. Grimes, Acting Assistant Director
for Systems Engineering
Division of Operating Reactors
1. Request for Information
2. NUREG-0462, dated July 1978
REQUEST FOR INFORMATION
TARGET ROCK SAFETY/RELIEF VALVES
1. What is the status of each of the Target Rock safety/relief valves at
your plant(s); i.e.:
a. Are they in their original design configuration?
b. What is the existing simmer margin?
c. What modifications have you implemented to improve reliability?
d. On what date were these modifications made?
2. What maintenance and testing do you routinely perform on these valves
and how often is it performed?
3. What additional modifications and/or maintenance do you plan to
implement in the future?
4. On what date will the modification(s) and/or maintenance in item 3 be
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