05000206/LER-1981-014, Forwards LER 81-014/03L-0.Detailed Event Analysis Submitted

Forwards LER 81-014/03L-0.Detailed Event Analysis Submitted

ML20009G844
Person / Time
Site:	San Onofre
Issue date:	07/27/1981
From:	Haynes J SOUTHERN CALIFORNIA EDISON CO.
To:	Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
Shared Package
ML20009G845	List: 05000206/LER-1981-014, Forwards LER 81-014/03L-0.Detailed Event Analysis Submitted 05000206/LER-1981-014-03, /03L-0:on 810630,erroneous Feedwater Flow Indications Produced Indicated Power Levels Below Actual Levels.Caused by Feedwater Flow Orifice Installed Backwards on One Line & Excessive Vibration on Other
References
NUDOCS 8108050142
Download: ML20009G844 (2)

LER-1981-014, Forwards LER 81-014/03L-0.Detailed Event Analysis Submitted

Event date:
Report date:
2061981014R00 - NRC Website

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Southern California Edison Company

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Attention: Mr. R. H. Engelken, Director A

cp DOCKET No. 50-206 SAN ON0FRE - UNIT 1

Dear Sir:

This letter describes a reportable occurrence involving the operation of Unit 1 at a power level higher than indicated by secondary heat balance. This submittal is in accordance with the reporting requirement of Section 6.9.2.b(1) of Appendix A to the Provisional Operating License DPR-13.

The established method of calculating reactor thermal power is a secondary plant heat balance which involves the use of feedwater flow and temperature and steam pressure. On June 30, 1981 it was noted that erroneous feedwater flow signals were producing indicated power levels below those of actual power.

Several calculations, including a primary plant heat balance, indicated that actual reactor power was as much as 10% higher than the indicated level of 84%. However, the plant was not operated above 405 MWe since its return to service on June 16, 1981, which was well below the established 100% reactor power capacity of 432 MWe. Operation continued with the power range meters set to 100% until the reactor tripped from another cause on July 2,1981. Subsequent analysis showed that in the event of a power excursion during this period the reactor would have tripped at or below 118% power which is the value assumed in the FSA.

Investigation revealed that the flow orifice in loop C was installed backwards during the last outage. The orifice was removed, inspected, and properly installed on July 3, 1981.

Excessive vibration in feedwater loop A caused by chatter in the control valve caused the differential pressure meter on loop A to read low. Placing the A feedwater valve control into MANUAL corrected the low meter reading. Subsequently the A feedwater bypass valve was opened and the feedwater control returned to AUTO. Operation in this mode resolves the problem and will be used until permanent repairs to the A feedwater control valve vibratien dampener are completed.

8108050142 810727 PDR ADOCK 05000206

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U. S. NUCLEAR REGUIA'IORY ComISSION MR. R. H. ENGEMEN, DIRECTOR LICENSEE EVENT REPORT No.81-014 PNE 2 If you should have any questions please contact me.

Sincerely N

J. G. Haynes Manager of Nuclear Operatice N:akp Enclosure: Licensee Event Report 81-014 cc:

U. S. Nuclear Regulatory Ccruission Office of Inspection and Enforcanent U. S. Nuclear Regulatory Comnission 7

Office of Manacrement Information & Program Control y

Nuclear Safety Analysis Center L. F. Miller (USNBC Resident Inspector) 1 hals '