ML13330A434

Forwards Request for Addl Info Re Reactor Vessel Overpressure Protection.Response Requested by 820215

ML13330A434
Person / Time
Site:	San Onofre
Issue date:	12/28/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Dietch R Southern California Edison Co
References
TAC-06836 LSO5-81-12-079, LSO5-81-12-79, NUDOCS 8201190781
Download: ML13330A434 (5)

Text

2 DISTRIBUTION Docket NRC PDR December 28, 1981 Local PDR ORB Reading

/0 D Crutchfield Docket No. 50-206 WPaulson LS05-81 =12-079 ELD 4,

CRS (10)

Mr. R. Dietch Vice President Nuclear Engineering and Operations Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, -California 91770

SUBJECT:

REACTOR VESSEL OVERPRESSURE PROTECTION -SAN ONOFRE UNIT 1 We are continuing our review of the multi-plant issue of reactor vessel overpressure protection for San Onofre Unit 1 and find that additional infommation described in the enclosure to this letter is needed. We request your response by February 15, 1982.

The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, 0MB clearance is not required under P.L.96-511.

Sincerely, Original signed by Walter A. Paulson for/

Dennis M. Crutchfield, Chief Operating Reactors Branch #5 Division of Licensing

Enclosure:

Request for Additional Information cc w/enclosure:

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NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD CO PY USGPO: 1981-335-960

-Mr. December 28, 1981 cc Charles R. Kocher, Assistant General Counsel James Beoletto, Esquire Southern California Edison Company Post Office Box 800 Rosemead, California 91770 David.R. Pigott Orrick, Herrington & Sutcliffe 600 Montgomery Street San Francisco, California 94111 Harry B. Stoehr San Diego Gas & Electric Company P. 0. Box 1831 San Diego, California 92112 Resident Inspector/San Onofre NPS c/o U. S. NRC P. 0. Box 4329 San Clemente, California 92672 Mission Viejo Brahch Library 24851 Chrisanta Drive Mission Viejo, CalifoFnia 92676 Mayor City of San Clemente San Clemente, California 92672 Chairman Board of Supervisors County of San Diego San Diego, California 92101 California Department of Health ATTN: Chief, Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814 U. S. Environmental Protection Agency Region IX Office ATTN:

Regional Radiation Representative 215 Freemont Street San Francisco, California 94111

.REQUEST FOR ADD ONAL INFORMATION -

SAN ONOFRE *T

1. The Branch Technical Position,RSB '5-2,requires the overpressure mitigation systems (OMS) to meet single active failure analysis when the initiating cause of the event is not considered as tlte single active failure. In your analysis submitted in your October 12, 1977 letter, you cover many single failures of system active components.

Address the following scenario for the San Onofre OMS design.

Consider as the initiating event the failure of a vital bus which results in the isolation of letdown flow (i.e., fails c~hased PCV-1105 or TCV-1105 and LCV-1100A) and also fails closed one of the pressurizer PORVs. A postulated single failure (failing closed) of the other PORV would fail mitigating systems for the event. Discuss yo=ur plant's provisions to mitigate such a transient.,

2. The branch position requires an alarm to alert the operator to enable the OMS at the correct plant condition during cooldown.

You rely on a pressure actuated alarm to perform this function. How do you ensure that the Reactor Coolant System temperature does not fall below the allowable temperature corresponding to the above alarm pressure setpoint, thus violating limits specified in Appendix G to 10 CFR Part,35.

3. You state that the required functional test will be perfo-rmed prior to returning to a water-solid condition following a cold shutdown.

Will this test also be performed prior to placing the OMS inr service during plant cooldown from a long period of operation? If not,1 give your basis for not testing the OMS at this time.

4. The Branch Technical Position requires the OMS to functikon during an Operating Basis Earthquake. You stated that portions of your OMS meet Seismic Category A requirements and that the rest crf the system meets Category B.

a. Identify all sys components that are not desipled to function during an Operating Basis Earthquake, and give the basis for these components not meeting this requirement.

b. Analyze the situation if the components that you identify in 'a' above do not function.

c.

Are the PORV operators qualified to operate through an OBE?

5. Additional information is requested on the following items concerning RCS and RHR relief and safety valves.

a. Is the relief capacity of RHR safety valve RV-206 sufficient to mitigate the overpressure transients that were analyzed for the OMS?

b. Does the valve position indication of the PORVs and the PORV block valves indicate a control circuit condition or the acutal position of the valves?

6. In your May 2, 1977, you provided some information on the training that you conducted on the overpressurization incidents; provide the following additional training information.

a. What overpressure training have you performed since 1977?

b. How do you ensure that a continued emphasis is placed on possible overpressurization situations in your licensing and retraining programs?

c. How is this training and LER review documented?

7. List all the administrative procedures and controls used to minimize the probabilit of an overpressure transient (i.e., solid-water operation limited to certain RCS pressure, pressurizer heater and HPI pumps disabledat certain RCS pressure and temperature, etc.)

Indicate which of the adminstrative procedures and controls is in the plant's Technical Specifications.

8. *What is the present s atus of the San Onofre-1 OMS?

a. Have all permanent OMS installations and modifications been completed?

b. Have warnings and caution notes been included in all affected procedures?

c. Have all necessary administrative documentation changes'been made?

9. If the Westinghouse generic analysis applies to your plant, reactor coolant pumps should not be started when water-solid and with a temperature difference of >50 0 F between the RCS and the steam generator secondary side.

What means are available at San Onofre-l to determine the respresentative temperature difference between the RCS and the steam generators?

10.

During the 'no-flowltest' portion of the cold operation test of the safety injection system while the RCS is water-solidis the OMS lined up to provide RCS protection?

If the answer above is-no, please discuss the following secenarios:

If an operator error resulted in a feedwater pump breaker not being racked out; would starting a feedwater pump, with only the water in the piping as a source (suction valves remain closed),

result in violating the Appendix G limits?

11. Have you continued the practice of assigning a reactor operator to monitor reactor coolant system pressure during all water-solid operations?

12.

What is the reactor pressure vessel age in effective full power years(EFPY) for which your Appendix G limits are calculated?