05000219/LER-1981-021, Forwards LER 81-021/03L-0.Detailed Event Analysis Encl

Forwards LER 81-021/03L-0.Detailed Event Analysis Encl

ML20009A540
Person / Time
Site:	Oyster Creek
Issue date:	07/01/1981
From:	Finfrock I JERSEY CENTRAL POWER & LIGHT CO.
To:	Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20009A541	List: 05000219/LER-1981-021, Forwards LER 81-021/03L-0.Detailed Event Analysis Encl 05000219/LER-1981-021-03, /03L-0:on 810603,high Pressure Sensor Switches RE03B, C & D Were Found Tripping within 8 Lb of Setpoint. Caused by Sensor Repeatability.Switches Reset.Switches Will Be Replaced During Next Refueling Outage
References
NUDOCS 8107130293
Download: ML20009A540 (3)

LER-1981-021, Forwards LER 81-021/03L-0.Detailed Event Analysis Encl

Event date:
Report date:
2191981021R00 - NRC Website

text

e

.s

~

OYSTER CREEK NUCLEAR GENERATING STATION

-: wa (JLS*Tf lf,70'$T.'

(609)693-6000 P O. BOX 388

• FORKED RIVER

• NEW JERSEY

• 08731 ce e<hc u..m s, stem July 1, 1981 E

Office of Inspection and Enforement 2Y

' 7/

Mr. Boyce H. Grier, Director N

I p " 4.L.i n ;9 3 7 Region I United States Nuclear Regulatory Ccmnission p Q u = 2 4., j N

631 Park Avenue M;.l V

, css /

King of Prussia, Pennsylvania 19406

'(

.. _,;.,p V

Dear Mr. Grier:

~ P SUBJECT: Oyster Creex I?uclear Generating Station Docket No. 50-219 Licensee Event Report Reportable Occurrence No. 50-219/81-21/3L This letter forwards three copies of a Licensee Event Report to report Reportable Occurrence No. 50-219/81-21/3L in cmpliance with paragraph 6.9.2.b.1 of the Technical Specifications.

Very truly yours, j'

Ivan R. Finfrock, r Vice President w i&L Director - Oyster Creek IRF:dh Enclosures cc: Director (40 ccpies)

Office of Inspection and Enforement United States Nuclear Pegulatory Ccmnission Washington, D.C.

20555 Director (3)

Office of Management Infornution and Program Control United States Nuclear Peculatory Ccmnission Washington, D. C. 20555 NRC Resident Inspector (1)

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

[//

8107130293 810701 PDR ADOCK 05000219 PDR g

OYSTER CREEK NEXIERR GENERATING STATION Forked River, New Jersey 08731 Licensee Event Report Reportable Occurrence No. 50-219/81 21/3L Report Date July 1, 1981 Occurrence Date June 3, 1981 Identification of Occurrence Reactor High Pressure Sensors RE03B, C and D trip settings were discovered to be greater than 1060 PSI.

This event is considered to be a reportable occurrence as defined in the Technical Specifications, paragraph 6.9.2.b.1.

Conditions Prior to Occurrence The plant was operating at Steady State Conditions.

Power:

Reactor 1518 FMt Generator 488 MWe 4

Flow:

Recirculation 14.1 x 10 gpn Feedwater 5.55 x 106 lb/hr Description of Occurrence On June 3,1981 at approximately 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br /> while perfonning surveillance testing, Reactor High Pressure Sensors RE03B, C and D were found to be less conservative than that specified in Technical Specifications. Tests on all reactor high pressure switches revealed the following data:

Press. Switch Desired Designation Setpcint As Found As Left Systen 1 RE03A 1068 1068 1068 RE03C 1066 1070 1065 Systen 2 RE03B 1068 1074 1068 RE03D 1066 1074 1064 Apparent Cause of Occurrence The cause of switches RE03B, C and D tripping within 8 PSI of setpoint is attributed to sensor repeatability.

~

Reportable Occurren Page 2 Report No. 50-219/81-21/3L Analysis of Occurrence The setting on the reactor high pressure scram, anticipating scram, reactor coolant systs relief valves and isolation condanser have been established to assure never reaching the reactor coolant syst s pressure safety limit as well as assuring the systs pressure does not excaed the range of the fuel clam 1ng integrity safety limit. In addition, the APR! neutron flux scram and the turbine bypass syst s also provide protection for these safety limits e.g. turbine trip and loss of electrical load transients.

In aMition to preventing power operation above 1060 PSIG, the pressure scram backs up the other scrams for these transients and other steam line isolation type transients. Analysis of worst case turbine trip, without bypass transient indicates that the relief valves limit the peak pressure to 1188 PSIG well below the 1250 PSIG range of applicability of the fuel clam 1ng integrity safety limit and the 1375 PSIG reactor coolant systs pressure safety limit. The reactor coolant syst s safety valves offer yet another protective feature for the reactor molant syst s pressure safety limit since these valves are sized assuming no credit for other pressure relief devices.

The safety valves are sized according to the code for a condition of turbine stop valve closure, without bypass, while operating at 1930 Wt followed by (1) a delay of all scrams (2) failure of the turbine bypass valves to open and (3) failure of the isolation condensers and relief valves to operate. Therefore, the safety significance of this event is considered minimal because each switch would have operated within 8 psig above setpoint limit.

Corrective Action

Innw11 ate corrective action consisted of resetting Reactor High Pressure Sensors RE03B, C and D to trip withm the prescribed safety limit. An engineering study concerning related probles with ITT Barton Switches indicated scme contribution of calibration techniques with instrument repeatability. Certain reccmnendations of the study, including independent verification of as-found values, if out of l

tolerance adjusting the test variable slowly to avoid misroarling, and verification of %vitch actuation value after the switch lock is tightened will be incorporated into our calibration procedures for these switches. The long term solution will be the replac ment of the RE03 pressure switches with an analog trip syst s during our upccming refueling outage.

Failure Data Barksdale Pressure Actuate-1 Switch (3)

Switch B2T-Al2SS Proof 1800 PSI Adjustable Range 50-1200 PSIG l

i m

p,

,-w.-

m