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January 19, 1990 .I

Docket No. 50-277;

Document Control Desk i

U. S. Nuclear Regulatory CommissionWashington, DC 20555'

SUBJECT: Licensee Event ReportPeach Bottom Atomic Power Station - Unit 2

This LER concerns a Unit 2 Scram due to personnel error during testing iactivities. '

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Reference: Docket No. 50-277Report Number: 2-89-033Revision Number: 00Event Date: 12/20/89Report Date: 1/19/90Facility: Peach Bottom Atomic Power Station

RD 1, Box 208, Delta, PA 17314

i This LER is being submitted pursuant to the requirements of 10 CFR| 50.73(a)(2)(iv).

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' W. T. Russell, USNRC, Region I '

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On 12/20/89 at 1752 hours with Unit 2 operating at 100% power, a full scram signalwas received when a technician performing a surveillance on Average Power RangeMonitor (APRM) "0" inadvertently operated a switch on APRM "A" (APRM "0" actuates a"B" channel half scram signal while APRM "A" actuates an "A" channel half scramsignal). The reactor feedwater pumps (RFP) tripped following the scram. The HighPressure Coolant Injection (HPCI) and the Reactor Core Isolation Cooling (RCIC)systems actuated as designed to maintain reactor water level, other safety systemsoperated as designed. Control Rod 38-39 settled to position 02 shortly following thescram and was later re-inserted. The root cause of the event has been attributed toprocedural deficiencies and inattention to detail by the technician performing thesurveillance. The technician involved in this event was counseled and disciplined.APRM Surveillance procedures which test APRM channels adjacent to other APRM channelshave been revised to provide physical barriers between APRM channels during testingand to instruct operators to bypass adjacent APRM channels when permissible. Currentpanel labeling will be evaluated and improved as appropriate. There was one previoussimilar event,

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Requirements for the Report

| This report is required pursuant to 10 CFR 50.73(a)(2)(iv) because an inadvertent ;

equipment manipulation resulted in an engineered safeguard feature actuation (i.e.,Reactor Protection System (EIIS:SC)).

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Unit Status at Time of Event i

Unit 2 was operating at 100% power. A surveillance was in progress on Average PowerRange Monitor (APRM) "0" (EIIS: MON).

Description of Eved !

On 12/20/89 at 1730 hours with Unit 2 operating at 100% power, a surveillance wasbeing performed on Average Power Range Monitor (APRM) "D". At 17:52:54 hours, with a"B" Reactor Protection System (RPS) channel half scram signal inserted per the ~,

ongoing surveillance, an "A" RPS channel half scram signal was received when one ofthe technicians (non-utility, non-licensed) performing the surveillance inadvertentlyoperated a switch on APRM "A" resulting in a full reactor scram.

Five seconds following the scram signal, indicated reactor level dropped to 0 inches !(approximately 172 inches above top of active fuel) resulting in a Pr.imaryContainment Isolation System (PCIS) (EIIS:JM Group II and III isolation and a

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Standby Gas Treatment (SBGT) System (EIIS:BH initiation. Ten seconds following the *

scram (17:53:04 hours) the High Pressure Coo ant Injection (HPCI) System (EIIS:BJ),theReactorCoreIsolationCooling(RCIC) System (EIIS:BN),andtheAlternateRodInjection (ARI) System auto initiated as designed when indicated reactor water level idropped to less than -48 inches.

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AT 17:53:07 hours, the A, B and C Reactor Feed Pumps (RFP) (EIIS:SJ) unexpectedly ;tripped. At 17:53:34, reactor level returned to an indicated level of 0" and the '

HPCI and RCIC Systems were manually shutdown. '

AT 1801 hours Control Rod (EIIS: ROD) 38-39 was observed to have settled to position02 (position 48 is fully withdrawn). At 18:15 hours reactor water level wasstabilized to an indicated +24 inches, the PCIS Group II and III Isolation was reset,and the SBGT System was removed from service. At 1825 hours, the RPS and ARI signalswere reset. On 12/21/89 at 0250 hours. Control Rod 38-39 was inserted from position02 to 00. -

Cause of the Event

The proximate cause of the reactor scram was the inadvertent operation of the modeswitch on APRM "A" while APRM "D" was in test. The root cause of the inadvertentoperation of the mode switch on APRM "A" appears to be a combination of the followingtwo conditions:

a. Inattention to detail by the technician performing the surveillance in that he *

did not realize that he was operating the wrong switch, although he hadpreviously operated the correct switch during the surveillance.

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b. Procedural deficiencies in that the procedure should have specified that theadjacent APRM channel be bypassed prior to performing the test. ;

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The cause of the RFP trip appears to be associated with the low suction pressure trip jcircuitry on each pump. The current setpoint for the low suction pressure trip is300 psig with (at the time of the event) a common time delay of seven seconds foreach pump. It is believed that the sudden increase in Feedwater flow caused suctionpressure to drop below 300 psig for a period of time longer than seven seconds.

The settling of Control Rod 38-39 to position 02 appears to have been caused by a |reverse flow condition above the drive piston on the Control Rod Drive Mechanism 1

(EIIS:AA). This reverse flow condition is caused by broken or badly worn stop pistonseals (Ells: SEAL). Post scram testing of Control Rod 38-39 yielded results whichsupport this theory.

Analysis of Event

No actual safety consequences occurred as a result of this event.

The Reactor Protection Systein operated properly throughout this event. Other safety ijsystems operated as designed. j

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The unexpected trip of the Reactor feedwater Pumps did not cause a significantchallenge to plant safety. HPCI and RCIC actuated prior to the RFP trips and wereable to maintain reactor water level as designed. Operations personnel were able tore-establish feedwater flow shortly after the scram.

The Technical Specification required shutdown margin was maintained even thoughControl Rod 38 39 settled to position 02. The maximum subtritical banked withdrawalposition for Unit 2 for the current fuel cycle (cycle 8) is notch position 02 (185Control Rods can be at position 02 and the required shutdown margin maintained).

General Electric Company technical information concerning Control Rod settling(Services Information Letter 52) indicates that in other than severe cases, ControlRods will settle to between positions 00-02 following a scram when broken or wornstop piston seals exist. Since post scram testing of Control Rod 38-39 yielded scramtimes within Technical Specification limits, it is felt that this Control Rod will!

not drift out further than position 02 if another scram occurs.| Corrective Actions

The following corrective actions have been taken:

1. The technician involved in this event was counselled on the importance ofattention to detail when operating sensitive equipment and disciplined as deemed *

appropriate by his supervision.

2. APRM surveillance procedures (including SI-2N-60A-APRM-DICW) with similar testingconditions have been revised to add cautions about the consequences of operatingthe wrong switch, to provide instructions to physically cover the switch on the

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adjacent APRM channel, and to instruct the unit operator to bypass the other APRM,channel on the panel in test when permissible.

3. The time delays on the low suction pressure trips on the reactor feed pumps havebeen changed from a common time of seven seconds to seven, twelve and fifteenseconds for the A, B, and C RFP's respectively.

The following corrective actions will be taken:

1. A team of technicians and engineers will review this event and providerecommendations for permanent labeling improvements on the APRM panels, permanentor semi-permanent physical barriers to prevent inappropriate switch operation,and changes to the surveillance procedure to minimize the length of time a half-scram signal is present because of the test.

2. The current RFP low suction pressure trip point of 300 psig will be evaluated todetermine if it should be lowered to reduce the possibility of pump trips duringtransient conditions.

3. Control Rod 38-39 normal testing will be monitored during the current operatingcycle to identify any further degradation in the Rod's performance. ;

4. During the next refueling outage on Unit 2, the Control Rod Drive Mechanism foriRod 38-39 will be repaired as required.

Previous Similar Events

One previous similar event reported in LER 2-88-02 was identified. This eventinvolved the operation of the wrong control switch by a unit operator which resulted

! in a full scram signal while the unit was shutdown. The corrective actions taken asa result of LER 2-88-02 involved the stressing of attention to detail by operationspersonnel only and, therefore, would not have prevented this event (the technicianinvolved in this event is part of the Maintenance / Instruments and Controls group atthe site.)

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