05000219/LER-1980-047, Forwards LER 80-047/03L-0

Forwards LER 80-047/03L-0

ML19345B808
Person / Time
Site:	Oyster Creek
Issue date:	11/25/1980
From:	Finfrock I JERSEY CENTRAL POWER & LIGHT CO.
To:	Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML19345B809	List: 05000219/LER-1980-047, Forwards LER 80-047/03L-0 05000219/LER-1980-047-03, /03L-0:on 801101,while Performing Core Spray Sys Instrument Channel Calibr & Test,High Drywell Pressure Switches Tripped Above Tech Spec Limit.Cause Attributed to Instrument Repeatability.Switches Adjusted
References
NUDOCS 8012020552
Download: ML19345B808 (3)

LER-1980-047, Forwards LER 80-047/03L-0

Event date:
Report date:
2191980047R00 - NRC Website

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OYSTER CREEK NUCLEAR GENERATING STATION

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• NEW JERSEY

• o8731 u% som, November 25, 1980 Mr. Boyce H. Grier, Director Office of Inspection and Enforcement Region I United States Nuclear Regulatory Commission 641 Park Avenue King of Prussia, Pennsylvania 19406

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Dear Dr. Grier:

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E p Ji SUBJECT: Oyster Creek Nuclear Generating Station

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Docket No. 50-219

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Licensee Event Report Reportable Occurrence No. 50-219/80-47/3L C

This letter forwards three copies of a Licensee Event Report to h

report Reportable Occurrence No. 50-219/80-47/3L in compliance with paragraph 6.9.2.b.(1) of the Technical Specifications.

Very truly yours,

'N' van R. Finfroc Jr.

p Vice President Generation IRF:dh Enclosures cc: Mr. John G. Davis, Acting Director (40 copies)

Office of Inspection and Enforcement United States Nuclear Regulatory Conmission Washington, D.C.

20555 Mr. William G. Mcdonald, Director (3 copies)

Office of Management Information and Program Control United States Nuclear Regulatory Commission U shington, D.C.

20555 i

8012020 552--

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• 4 OYSTER CREEK NUCLEAR GENERATING STATION

.i Forked River, New Jersey 08731 Licensee Event Report

~ Reportable Occurrence No. 50-219/80-47/3L

'i' Report Date R,

November 25,1980 ~

t'l Occurrence Date November 1, 1980.

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' Identification of Occurrence Core Spray System high drywell pressure sw' itches tripped at a value greater than

_ that specified in the Technical Specifications, Table 3.1.1 item D.2.

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- This event is considered to be a reportable occurrence as defined in the Technical

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_ Specifications,. paragraph 6.9.2.b.(1).

Conditions Prior to Occurrence The plant was operating at steady state power.

Plant parameters at the time of occurrence were:

Pcwer:

Core 1873 MWt

' Electrical 633 MWe.

6.94x10lgpm Flow:

Recirculation 15.1 x 10 Feedwater lb/hr Description of Occurrence On Saturday, November -1,1980, while performing the " Core Spray System Instrument 1

Channel Calibration and Test", the setpoints for the high drywell pressure twitches, RV-46A, RV-46B, RV-46C and RV-46D, were found to be less conservative than that sp'ecified in the Technical Specifications.

Surveillance testing revealed the following data:

Fr-essure' Switch Desired Setpoint As Found As Left Designation 4

RV-46A 1 psig 2.21 1.98 2

2 psig.

2.20 1.95 RV-468 2

RV-46C 12 psig 2.25 2.0 RV-46D 1 psig 2.05 1.94

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l Reportable _0ccurrence Page 2 Report No. 50-219/80-47/3L i-Apparen't Cause of Occurrence i.

The cause of the occurrence was g tributed to instrument repeatability. The 1

' instruments is approximately 2-3% (1 psig and the long tenn repeatability of the i

switches are set to trip at~2.0 0

.2.3 psig) of full range, Therefore, although j~

L the instruments ~ wil.1 be op'erating within design accuracy.. the technical s9ecifica-t tion limit _ of 2.0 psig can be exceeded during surveillance testing.

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.Analy' sis of Occurrence I

The' Core Spray Sy' stem is a' low pressure' standby; core cooling sys. tem which provides

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an alternate supply-of cooling water:to' the reactor following a postulated. pipe f(

break accident insthe reactor primary system;.The system consists of two 1

independent loops;. both of. which initiate upon receipt of a trip signal. from any

.one (1) of four (4) drywell pressure sensors or four (4) reactor low level sensors. :When the reactor pressure decreases to 285 psig isolation valves open to allow flow to the reactor. The surveillance data indicated that the core i.

spray system would have been initiated by high drywell pressure switch RV-46D at a pressure of 2.05 psig. The delay in initiating the core spray system is

.neglibible considering that the pressure in the.drywell is well above 2.0 psig within 0.1 second after the start of the accident.

Corrective Action

High drywell pressure sensors RV-46A, RV-46B, RV-46C and RV-46D were reset to trip within the Technical Specification limit of <2.0 psig.

It is recognized that there is a drift problem associated with the new snap-action switches which were installed during the past outage' to upgrade their seismic qualifications. An engineering study is in progress regarding the feasibility of replacing the existing sensors with a solid state system.

Failure Data ITT Barton Differential Pressure Indicating Switch Switch Model #288A 0.5 - 9.5 psig adjustable range

Previous Similar Events

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