ML19319D506
| ML19319D506 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/01/1976 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Rodgers J FLORIDA POWER CORP. |
| References | |
| NUDOCS 8003170632 | |
| Download: ML19319D506 (4) | |
Text
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Docket' File 3 %
W Local PDR J. Miller LWR 1 File R. C. DeYoung F. J. Williams J. Stolz L. Engle Docket iio. 50-302 E. Hylton bec:
J. R. Buenanan, flSIC R. Heineman T. B. Abernathy, TIC D. Ross Florida Power Corporation J. Knight,' SS ATIN:
Mr. J. T. Rodgers R. Tedesco Assistant Vice President and H. Denton Nuclear Project Manager V. A. Moore P. O. Box 14042 R. H. Vollmer St. Petersburg, Florida 33733 M. L. Ernst W. P. Gammill GentJemen:
ELD VmIFICATION FOR COMPLIAtiCE WIDI APPENDIX G PPESSURE-TEMPERATURE LIMITS DURING STAR 3JP NiD ShVIIXLN An unexpectedly large nuceer of reported instances of reactor vessel overpressurization in Pressurized Water Rcactor (PWR) facilities have occurred in which tne Technical Specifications implementing 10 CFR Part 50 Acpendix G limitations have been exceeded. 'Ihe majority of cases nave occurred during cold snutdown in wnich the primary system has been in water solid conditions. Tnese overpressurization events have been initiated by a variety of causes, including the following:
(1) Isolation of FliR systeWietdown system wnile charging to a water solid primary system,
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(2) Thermal expansion following tne starting of a primary coolant pump due to stored thermal energy in steam generators, (3) Inadvertent actuation of safety injection accumulators, and (4) Initiation of operation of a reactor coolant pump or a high pressure safety injection pump.
In essentially all of tne events reported, a single personnel error, equicoent calfunction or procedural deficiency has been sufficient to cause tne event.
In view of the potential seriousness of exceeding Ascendix G limitations, we believe that appropriate steps snould be taken prcmptly oy all PiR licensees to minimize tne likelihoco of aoditional occurrences of reactor. vessel overpressurizatica. To that end we recently completed a series of meetings with sevaral P6a licensees ard W
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I '97a Florida Power Corporation 2-NSSS suppliers in wnich we discussed the recorted overpressurization events aM assessed tne measures that aro currently being employed to either avoid or reduce the probability of similst occurrences, or to control the pressure transient to less than Apndix G liraits.
Exscples of those measures identified oy the various licensees are as follows:
(1) Complete avoidance of water solic conditions by eitner maintaining a pressurizer steam bubole or by providing a low pressure nitrogen blanket in the pressurizer wnen a steam bubole cannot be maintained,.
(2) Disabling flign Pressure Injection _and_ Safety _InjectionJ;upps.by_
disconnecting electrical powr supplies when at low primary system temperatures, (3) Installation of.dusi setpoint pressurl=er power relief valve (s) to provide protection against exceed $ng Appendix G limits while at icw primary system temperatures, (4) Minimization of time at water solid conditions and upgrading plant procedures to include appropriate we nings and cautions wnen suen operations are necessary, and (5) Installation of relief valves in charging pump discharge lines with a setpoint to provide protection against exceeding Appendix G limits.
It was noted in our discussions with the PWR licensees tnat, for the majority of those plants involved, not all potential overpressuri:ation events would be prevented oy tne measures tney had identified and that sone of the remaining measures may have undesiracle effects on reactor safety.
assed on tne information gathered to date, we have concluded tnat all PWR licensees should evaluate tneir system designs to determine the vulnerability to overpressurization events. Specifically, you snould proviae tne following:
(1) An analysis of tne Reactor Coolant System (IC) response to pressure transients that can occur during startup and shutdown. Any design mooifications determined to be necessary u.
. der to preclude exceeding Appendix G liraits are to be incorporated in 'this analysis.
'Ihe analysis should include a plot of pressure as a function of ti:re until termination of the event. ':he analysis should assume the h[
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'most limiting initial conditions (e.g., one PRR train operating or available for letdown, other ccxnponents in normal operation when the system is water solid suEh as pressurizer heaters and enarging pumps, and one or more reactor coolant pumps in operation) with the worst single failure or operator error as the initiating event.
Justification should ce provided for the enoice of limiting conditions and worst single failure or operator error assumed in the analysis.
(2) A description of those design modifications determined to be necessary, including equipment performance specifications and systea operatione.1 sequences. 'ihe oesign basis used in the choice of equipnent snculd be included, and (3) A schedule for the prompt imolementation of the proposed' design modifications.
The basic criteria to be applied in determining the adequacy of over-pressurization protection are tnat no single equipaent failure or single operator error will result in Agendix G limitations being exceaded.
For those situations in which the necessary design cnanges identified i
cannot be implemented within the next few months, you should identify short-term measures to reduce the likelihood tnat overpressurization events will not occur in the interim period until the permanent design changes can be made. Short term measures enould be identified separately for irmediate implementation subject to the terms and conditions of your license. Short term measures might consider ccce ccooination of, but would not be limited to, the following suggestions:
(1) Procedural changes to minimize the time in whicn the primary system is in a water solid condition, (2) Upgra: ling existing plant procedures and administrative controls to assure that appropriate warnings and cautions are included to alert the operator wnenever the potential for primary system overpressurization exists, (3) Provide alarms and/or indications to alert the operator wnenever primary system pressure increases toward Appendix G limits, (4) Introducing temporary plant modifications for pressure relief, and (5) Assignment of additional personnel to monitor plant operations wnen water solid, oenc = >
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Florida wwer Corporation f You snould be aware that the design modifications required to preclude or miniraize the probability of reactor vessel overpressurization events, are plant dependent and that the examples given may or may not be adaptable to yoGF spiidific ~ system design. In addition, proper consideration must be given to tne potential effects of both the snort term and long term measures you consider to insure that other aspects of nuclear safety are not comprmised.
To verify compliance with Appendix G pressure-temperature limits during startup and shutcbwn, you snould assure that the appropriate instrumen-tation is installed to provide a continuous permanent record over tne full range of both pressure and temperature. Itiis instrumentation should-ce in service during long periods of cold snutdown as well_as_during startup and 5Edt' M~ operations. Reliance upori the plant computer to reconstruct a preseure transient is not considered sufficient oecause of tne likelihood of cceputer cbwntime especially during plant shutdown conditions.
We request that within 20 days after receipt of this letter you notify us that you will provide all the information requested within 60 days or explain why you cannot meet this schedule and provide the schedule tnat you will meet.
This request for generic information was approved by GAD under a blanket cicarance number B-180225 (R0072); this clearance expires July 31, 1977.
Sincerely, Original Siene.1 by' John F. Stolz John P. Stolz, Chief Light Water Reactors Brancn tio.1
, Division of Project Margement cc: Mr. S. A. Brandimore Vice President and General Counsel P. O. Dox 14042 St., Petersourg, Florida 33733 I
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