ML19322B873
| ML19322B873 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/11/1976 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 7912060704 | |
| Download: ML19322B873 (5) | |
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- and 50-237 Duke Po,er Cor:pany ATTN: Er. William O. Parker, Jr.
Vice President Stcan Production Post Of fica Box 217d 422 South Church Street Charlotte, Morth Carolina 28242 Centiccen:
RE: Cconee f.uclear Station (! nits Nos.1, 2, and 3 A nmeer of repcrted instances of reactor vessel overpressurization in Pressurized Water Reactor (PWR) facilities have occurred in wnich the Tecnnical Specifications implementing 10 CFR Part 50 Appenoix G limitations have been exceeced. The majority of cases have occurred during cold shutcown in which the pricary systen nas been in water solic ccnditiens. These overpressurization ovents nave Deen initiated by a variety of causes, including the following:
(1)
Isolation of Irn7 systesa/letacwn system while charging to a water solid pritaary systea, (2) Tneraal expansion following the starting cf a primary ccolant pm.p due to stored thermi energy in staac generators, (3) Inauvertent actuatton of safety injection accunulators, and (4)
Initiation of oceration of a reactor coolant pump or a high pressure safety indection pump.
In essentially all of the events reported, a single personnel error, equirent nalfunction or procecural ceficiency has been sufficient to cause tr.e event.
' e believe that appropriate steps should be taken pronptly by all M licensees to minimize tne likelthood of additional occurrences or reacter vessel overpressurization. To enat end we recently ctgleted a series of meetings with several PWt licensees and GS succliers in whicn we discussed the repcrted overpressur-1:st1on events snd assessac the wasures tnat are currently being
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,,,,,, i l icits. Ex eples of tnose neasures identified Dy tne rarious licensees ~nre U TotioWI.
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Dure Power Coopany AUG 1 1 1976 (1)' Cor41ete avoicance of water solid conditiens by either eaintaining a pressurizar steam bubole or cy providing a low pressure nitrogen olantet in the pressurizer when a stea-cuoole cannot be maintair.ec, (2) Disan11ng High Pressure Injection and Safety Injection pumps ey discenr.ecting electrical power supplies when at icw priaary system temperatures, (3)
Installation of dual setpoint pressurizer power relief valve (s) to provide protection against exceeding Appencix G listits while at low primary system toeperatures, (4) Minimization cf tioe at water solid conditions and upgrading plant procedures to include aporopriate warnings and cautions when such operations are necessary, and (5) Installation of relief valves in charging puep discharge lines l
with a setpoint to provido protecticn against exceeoing Appennin 6 licits.
It e.as noted in our discussiona with the PhR licensees tnet, for the uajority of those plants involved, not all potential overpressurization events would be prevented by the exasures tney had icentified and that some of the remaining messures may have uncestrable effects en reacter safety.
Based on the inforntion gathered to date, we have ccccluced that all ph2 licensees should evaluate their systeri cesigns to cetemine susceptibility to overpressurization events. Specifically, you should provice ste following:
(l) An analysis of tne iieactor Coolant Syste: (;tC5) resoonse to pressure transients that can occur durin; startup ar.d snutccwn. Ariy design sodifications cetermined to be necessary to preciade exceecing Appencix G 11oits are to be incorporated in this analysis. The analysis s.tould incluca a plot of pressure as a function of time until termination of the event. The analysis should assze the most liatting initial conditiens (e.g., one RMt train Operating or availaole for letcown, other coexonents in r.ormal o;:eration wnen the syste= is water solid such as pressurizer heaters and charging puwes, and one or more reactor coolant pu=ps,in operatien) with the worst single failure or operator error as the initiating event.
Justification should be provided for the choice of 11: siting concitions and worst single failure or operator error assumed in tne a ulvsis.
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o 3-Duke Power Cocoany (2) A description of those design modifications detemined to be necessary, including equipment perfornance specifications and systen operational secuences. The design basis used in the choice of equipment should oe included, and (3) A schedule for the prompt implementation of the proposed design nodifications.
The casic criteria to be applied in detemining the adecuacy of over-pressurization protection are that no single equipment failure or single operator error will result in Appendix G limitations being exceeded.
For those situations in which the necessary design changes identified cannot be i.mplemented within the next few months, you should identify short-term measures to reduce the Itkelihood that overpressurization events will occur in the interim period until the pemanent design changes can be made. Short tenn neasures snould be identified separately for imediato implementation within the terms and conditions of your license. Short tem ceasures might include some ccubination of, but would not be limited to, the following suggestions:
(1) Procedural changes to minimize the time in which the prinary system is in a water solid concition, (2) Upgrading existing plant procedures and administrative controls to assure that appropriate warnings and cautions are included to alert the operator whenever the ::ctential for primary system overpressurization exists, (3) Provide alarns and/or indications to alert tro operator whenever 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.
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Duke Power Company AUG 1 1 B76 Modifications to preclude or ainiaize the probability of reactor vessel overpressurf ution events are plant dependent and the examples given may or nay not be adaptable to your specific system design. Consideration must also be given to tto potential effects of both the short term and lonq tern measures you consider to assure that other aspects of nuclear safety are not compromised.
To verify cocpliefnce with Appendix G pressure-teoperature limits during startup and shutdown, you snould assure that the appropriate instrumen-tation is installed to provide a continuous pemanent record over the full range of ooth pressure and tecoerature. This instrumentation should be in service during long periods of cold shutdcwn as well as during startup and shutcown operations. Reliance upon the plant computer to reconstruct a pressure transient is not considered sufficient because of the likelihood of computer downtime especially during plant shutdown conditions.
lie request that you notify us within 20 days after receipt of this letter that you will previde all the infomation requested within 60 days or explain why you cannot acet this schedule and provide the schedule that you will meet.
This reauest for generic infomation was approved by MO under a blanket clearance number B-180225 (R0072); this clearance expires July 31, 1977.
Sincerely, Original signed by A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors cc: Mr. ',lilliam L. Porter FOR DISTRIBUTION SEE PAGE 5 Duke Power Company P. O. Box 217d 422 South Church Stree't Charlotte, North Carolina 28242 Mr. Troy B. Conner Conner & Knotts 1747 Pennsylvania Avenue, N. W.
Washington, 0. C.
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Duke Power Company AUS 1 1 1976 cci Oconee Public Library 201 South Spring Street Walhalla, South Carolina 2 % 91 DISTRIBUTION NRC PDR Local PDR Docket (3)
ORB #1 Reading KRGoller TJCarter OELD OI&E(3)
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