ML20012A017

LER 90-002-00:on 900128,reactor Manually Scrammed Due to Leak of Electrohydraulic Control Sys Fluid.Caused by Lock Nut on Interlock Dump Valve Setting Adjustment Bolt Becoming Unsecured Due to Sys Vibration.Leak stopped.W/900223 Ltr

ML20012A017
Person / Time
Site:	Peach Bottom
Issue date:	02/23/1990
From:	Cribbe T, Danni Smith PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-002-02, LER-90-2-2, NUDOCS 9003080132
Download: ML20012A017 (4)

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PHILADELPHIA ELECTRIC COMPANY PEAC}l BOTTOM ATOMIC POWI'R STATION R. D.1, Box 2o8 -

Delta. Itnnsyh2nia .17314 rwu montm .Tue rosse os excrut.wr _ (717) 456-7014 3

February 23, 1990 Docket No. 50-278 l

Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Licensee Event Report Peach Bottom Atomic Power Station - Unit 3 This LER concerns Engineered Safeguards Feature (ESF) actuations due to-reactor vessel level fluctuatior.t following a manual scram.

Reference:

Docket No. 50-278 Report Number: 3-90-002 Revision Number: 00 Event Date: 01/28/90 Report Date: 02/23/90

-Facility: Peach Bottom Atomic Power Station RD 1, Bcx 208, Delta, PA 17314 This LER is being submitted pursuant to the requirements of 10 CFR

-50.73(a)(2)(iv).

Sincerely, 4 n h/V ,4 E cc:_

J. J. Lyash, USNRC Senior Resident Inspector W. T. Russell, USHRC, Region I

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, . LICENSEE EVENT REPORT (UC DOCElf NUMetR 421 P' sACILITV NAME til Peach Bottom Atomic Power Station - Unit 3 0 l 5 l 0 l 0 l 01217 l 8 1 loFl o l3

' " Various Engineered Safety Feature System Actuations Due To Reactor Vessel Level Pluctuations After Manual Scram iVINT DATE tll L4R NUMBER tal REPOR T DATE (71 OTH(R I ACILITIES INVOLVED 451 YiAM [*f, MONTH DAY YEAR 9 AceLivv NAuss DOCKET NUMBEntsi MON 1H - OA V VEAR 5"U n' '

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[ vss ter env co.v.oiew axorcito sunwssion OA rts } NO 9l9 3l3 9 lq Assin Aci want w reso uwe, ie , amena ew, r.rwen e,no>e unce uvaew~rwn 1,nea no, On January 28, 1990 the Unit 3 reactor was manually scrammed due to a leak of Electrohydraulic Control System (EHC) fluid at the number one main turbine control' valve.. Reactor coolant level fluctuations following the scram resulted in three Group II and III Primary Containment Isolation System (PCIS) actuatte r The cause of the EHC. fluid leak was failure of an 0-ring on the fluid in1 6 p M t the servo valve which controls the number one main turbine control valvt. A n t.; tor level fluctuations were aggravated by the need for rapid depressurizaticn vit W main turbine bypass valves, inability to restart the 'C' reactor feedwater pump (RFP) and the subsequent batch feeding of reactor coolant via the High Pressure Coolant Injection (HPCl} and Reactor Core Isolation Cooling (RCIC) systems.- Inability to reset the 'C' RFr was due to a RFP turbine hydraulic control problem. The cause'of this problem was the lock nut on the RFP interlock dump valve setting adjustment bolt becoming unsecured due to system vibration. No actual safety consequences occurred as.a result of this event. The EHC fluid leak was stopped and the leaking servo valve was replaced. The failed 0-ring is undergoing analysis to determine the failure mode. The RFP interlock dump valve lock nuts on both units will be mechanically secured. No previous similar LERs were identified.

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' Unit-Status at Time of Event

. Unit 3 was in the Run Mode at 100 percent power.

Description of the Event

- On January 28,1990 at 0854 hours0.00988 days <br />0.237 hours <br />0.00141 weeks <br />3.24947e-4 months <br /> the Unit 3 reactor power was lowered to 50 percent and nanually scrammed due to a leak-of Electrohydraulic Control System (EHC) fluid at the' number one' main turbine control valve. Reactor water level then decreased to the

minus-eight; inch level resulting in an anticipated Group 11 and III Primary

~ Containment Isolation System (PCIS) actuation. The main turbine was m nually tripp1., Reactor-level quickly recovered and at 0858 hours0.00993 days <br />0.238 hours <br />0.00142 weeks <br />3.26469e-4 months <br /> reached the plus forty five inch. level resulting'in' automatic trip of the 'C' reactor feedwater pump (RFP).

The PCIS actuation was reset. At 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> the turbine bypass valves were manually opened approximately 25 percent to provide depressurization of the reactor. The 'C' RFP_could not be restarted so the Reactor Core Isolation Cooling System (RCIC) was manually started to control reactor level. However by this time 0909-hours, reactor level dropped to the minus three inch level resulting in a second Group 11 and III PCIS: actuation. The High Pressure Coolant Injection System (HPCI)'was then utilized intermittently to assist in controlling reactor level. The PCIS actuation wu again

-reset.: At 0933 hours0.0108 days <br />0.259 hours <br />0.00154 weeks <br />3.550065e-4 months <br /> raactor level decreased to the minus 2 inch level resulting in a third Group II and III PCIS actuation. By 0936 hours0.0108 days <br />0.26 hours <br />0.00155 weeks <br />3.56148e-4 months <br /> the Main Steam Isolation Valves were manually closed, reactor level had recovered, HPCI and the EHC pump were-secured.

Cause of the Event

'The EHC fluid leak was. caused by failure of an 0-ring-on the fluid inlet port to the servo valve for hydraulically operated valve H0-5275A which controls the number one

, main turbine control valve. The failed 0-ring is being analyzed to determine the cause of failure.

The. initial-Group 11 and III PCIS actuation was anticipated on low reactor level as a

,.* result of the manual scram. The second PCIS actuation occurred on low reactor level due to~the failure of the 'C' RFP to restart. This was caused by drift of the setting of the interlock dump valve which did not allow sufficient oil pressure to be

maintained to open-the RFP turbine control valve. The root cause of the setting drift of the interlock dump valve is attributed to the lock nut on the setpoint adjustment bolt becoming unsecured due to system vibration.

The third . low level PCIS actuation was due to the rapid level fluctuations created by

.the combination of attempting rapid depressurization via the turbine bypass valves

.whi.le batch feeding reactor coolant inventory with HPCI. Rapid depressurization was necessary due to the potential for the turbine bypass valves to fail closed on loss of EHC fluid.

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. Analysis of the Event No-actual safety consequences occurred as a result of this event, tA Group 11-and 111 PCIS actuation signal occurs when reactor coolant reaches the zero-inch level (172 inches above the top of active fuel). The initial low level PCIS actuation immediately followino the scram is not unexpected and is the result of void collapse upon insertion of the control rods. During this. event reactor level fluctuated within a range of plus fifty five Inches to minus 8 inches. A minus 8 inch level corresponds to 164 inches of reactor coolant above the tep of active fuel.

Closure of the main turbine bypass valves due to loss of EHC fluid would have been bounded by the Final Safety Analysis Report abnormal. transient analyses for a turbine trip without bypass or isolation of all main steam lines. Therefore, the fuel was not in danger of becoming uncovered at any time during this event.

Corrective Actions The EHC fluid leak was stopped and the leaking EHC servo valve w replaced. The failed servo valve 0-ring is undergoing analysis to determine the failure mode. The results of this analysis and associated corrective actions will be detailed in a revision to this report. A walkdown of the EHC fluid piping was performed and no other leaks were identified.

The A, B and C RFP interlock dump valves lock nuts were inspected in Unit 2 and 3.

The Unit 3 'B' RFP interlock dump valve lock nut was also found to be unsecured. The Unit 3 RFP interlock' dump valves were readjusteti and tested to verify proper setting.

The Unit 2 RFP interlock dump valve lock nuts were found to be properly secured. The RFP interlock dump valve lock nuts on both units will be mechanically secured to

prevent inadvertant movement of the adjustment bolt.

Previous Similar Events No previous similar Licensee Event Reports have been identified.

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