05000219/LER-1981-057, Forwards LER 81-057/03L-0.Detailed Event Analysis Encl

Forwards LER 81-057/03L-0.Detailed Event Analysis Encl

ML20038B595
Person / Time
Site:	Oyster Creek
Issue date:	11/30/1981
From:	Carroll J JERSEY CENTRAL POWER & LIGHT CO.
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20038B597	List: 05000219/LER-1981-057, Forwards LER 81-057/03L-0.Detailed Event Analysis Encl 05000219/LER-1981-057-03, /03L-0:on 811031,electronic Relief Valve High Pressure Switch IA-83B Found to Trip at Value Above Tech Spec Limit.Caused by Instrument Repeatability.Switch Reset
References
NUDOCS 8112080421
Download: ML20038B595 (3)

LER-1981-057, Forwards LER 81-057/03L-0.Detailed Event Analysis Encl

Event date:
Report date:
2191981057R00 - NRC Website

text

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OYSTER CREEK K

NUCLEAR GENERATING STATION

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(609) 693-1951 P.O. BOX 388

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m No aber 30, 1981 Mr. Ronald Haynes, Director Office of Inspection and Enforcemnt Region I United States Nuclear Regulatory Ccnmission 631 Park Avenue King of Prussia, Pennsylvania 19406

Dear Mr. Haynes:

SUBJECI': Oyster Creek Nuclear Cenerating Station Docket No. 50-219 Licensee Event Report Reportable Occurrence No. 50-219/81-57/3L This letter forwards three copies of a Licensee Event Report to report Reportable Occurrence No. 50-219/81-57/3L in compliance with paragraph 6.9.2.b.t of the Technical Specifications.

Very truly yours, l

W J. T. Carroll, Jr.

Acting Director ter Creek JIC:dh l

Enclosures cc: Director (40 copies)

Office of Inspection and Enforcenent United States Nuclear Regulatory Cannission l

Washington, D.C.

20555 i

Director (3)

Office of Managemnt Infonnation and Program Control United States Nuclear Regulatory Ca mission Waslungton, D. C. 20555 NRC Resident Inspector (1)

Oyster Creek Nuclear Generating Station Forked River, N. J.

8112080421 811130 b ).b PDR ADOCK 05000219 l j//

/

OYSIER CPEEK NUCLEAR GENEPATING STATION Forked River, New Jersey 08731

- Licensee Event Report Reportable Occurrence No. 50-219/81-57/3L Report Date Novenber 30, 1981 Occurrence Date October 31, 1981 Identification of Occurrence During surveillance testing, reartor high pressure switch IA-83B, for the "B" Electrcxnatic Relief Valve (DW), was found to trip at a value higher than that required by the Technical Specifications, section 2.3.4.

This event is considered to be a reportable occurrence as defincd in the Technical Specifications, paragraph 6.9.2.b.l.

Conditions Prior to Occurrence The plant was in the cold shutdown condition with reactor coolant less than 0

212 F.

. Description of 'tcurrence During performance of the "Electrmatic Relief Valve Pressure Sensor Test and Calibration", reactor high pressure switch IA-83B tripped at 1090 psig, which is greater than the corrected Technical Specifications limit of 1084.5 psig. The test was perforned due to the plant shutdown on the previous day (these switches are tested any time reactor coolant is brought below 212 F).

Testing performed on October 31 yielded the following data:

Corrected Corrected Corrected Tech.

As Found As left Switch Spec. Limit (psig)

Trip (psig)

Value (psig)

IA-83A

<1079.15 1059 1059 IA-83B 71084.5 1090 1083 IA-83C 71076.8 1074 1074 IA-83D 71082.2 1060 1060 IA-83E 71082.2 1080 1080 Apparent Cause of Occurrence

'Ihe cause of the occurrence was attributed to instrument repeatabil.ity. The switch was last reset at 1084 psig and tripped at 1090 psig. This 6 psig change is less than 1% of full range, which is well within the range of instrunst repeatability.

Reportable Occurrence Page 2 Report No. 50-219/81-57/3L Analysis of Occurrence The Electranatic Relief Valves are provided to renove sufficient energy fran the primary systen in order to prevent the Safety Valves fran lifting, even during the most severe transients. An analysis of a turbine trip fran full power with a failure of the bypass valves-has shown that with an EMRV setpoint of 1125 psig the D4RV's are capable of limiting the peak pressure below 1200 psig, which is below the setpoint of any of the safety valves. 'Ihis is also well below the fuel design pressure limits and reactor coolant systen pressure safety limit.

The Technical Specification limit of 1070 psig (with the head correction,1084.5 psig for "B") was chosen to maintain an adequate margin between peak pressure and the setpoint of the first group of safety valves. Since all DIRV's were functional and the "B" EMRV would have lifted only 6 psig.over the limit ("B" EMRV is also on a 1.5 second time delay), the safety significance is considered minimal.

Corrective Action

The switch was reset to trip within the required value (note the "As Left" value in the Description of Occurrence). These switches are also scheduled to be-replaced with an improved model during the upacming refueling outage.

Failure Data Manufacturer - Barksdale Model - B2SH12SS Range 1200 psig