ML19332E759

LER 89-008-00:on 891027,primary Containment Isolation Sys Group II & III Isolations Occurred When Spurious Reactor Low Level Signal Sensed by Instruments.Possibly Caused by Air Bubble in Sensing Lines.Line backfilled.W/891127 Ltr

ML19332E759
Person / Time
Site:	Peach Bottom
Issue date:	11/27/1989
From:	Cribbe T, Danni Smith PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-008-03, LER-89-8-3, NUDOCS 8912120089
Download: ML19332E759 (4)

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November 27, 1989 Docket No. 50-278

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' Document Control Desk U. S. Nuclear Regulatory Commission

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SUBJECT:

Licensee Event Report (LER)

Peach Bottom Atomic Power Station - Unit 3

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This LER concerns a Engineered Safeguards Feature (ESF) actuation due to a spurious low level signal. 4

Reference:

Docket No. 50-278 Report Number: 3-89-008 Revision Number: 00 Event Date: 10/27/89 Report Date: 11/27/89 Facility: Peach Bottom Atomic Power Station ,

RD 1, Box 208 Delta, PA 17314 e

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

Sincerely,

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cc: J. J. Lyash, USNRC Senior Resident inspector W. T. Russell, USNRC, Region I T

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1 l At 0346 on 10/27/89 with Unit 3 in cold shutdown, a Primary Containment Isolation i~

System (PCIS) Group II and III Isolation occurred when a spurious reactor low level l signal was sensed by one group of reactor level instruments. Other reactor level

( instrumentation indicated no change in reactor level at the time of this-event. The Residual Heat Removal (RHR) System isolated as a result of the PCIS Isolation. Other

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L safety systems operated as designed. There were no actual safety consequences as a l result of this event. The exact cause of this event-is unknown. It is believed that an air bubble was induced into the affected instrumentation sensing lines during excess flow check valve (EFCV) testing activities causing a brief low level signal to be generated by the instruments on that sensing line. The RHR System was returned to service at 0440. The affected instrument sensing line was backfilled on 11/13/89.

l The EFCV testing procedure will be revised to reduce the possibility of air intrusion L into instrument sensing lines. There was one previous similar event.

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UCENSEE EVENT REP 3RT (LER) TEXT CONTINUATION AreRovie ove No. sino-oio4 -

,, EXPIRES. 8731/OB P ACILtVY NAML (1) DOchti hum 8t h (21 ggR WUMetR ($1 Pact (31 Peach Bottom Atomic Power Station "'" " W A O.U Unit 3 0 l5 l0 l0 l0 l 2l7 l8 8l9 -

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0l0 0l2 oF 0l 3 TLXT IJ neore apoco e revuwed was samonet NRC Form JAM W 11h Requirements for the Report This report is required pursuant to 10CFR50.73(a)(2)(iv) due to an Engineering Safeguard Feature actuation (i.e., Primary Containment Isolation System (ElIS:JM).

Unit Status at Time of Event

-Unit 3 was in cold shutdown with the Residual Heat Removal (RHR) System in the shutdown cooling mode. Reactor water level was at +21" (0" reactor level equals approximately 172" above top of active fuel). The Reactor Water Cleanup (RWCU)

System was in service. A full SCRAM signal had previously been inserted.

Description of the Event <

At 0346 hours0.004 days <br />0.0961 hours <br />5.720899e-4 weeks <br />1.31653e-4 months <br /> on 10/27/89 with Unit 3 in cold shutdown, a reactor water level 0" alarm was received along with a Group II and III Primary Containment Isolation System (PCIS) isolations. As a result of the isolations, the Residual Heat Removal (RHR)

(EIIS:BO) tripped as designed. A Reactor Protection System (RPS) (EIIS:JC) actuation did not occur because a full SCRAM signal had previously been inserted.

At 0435 hours0.00503 days <br />0.121 hours <br />7.19246e-4 weeks <br />1.655175e-4 months <br />, the Group II and Group III PCIS isolations were reset. At 0440 hours0.00509 days <br />0.122 hours <br />7.275132e-4 weeks <br />1.6742e-4 months <br />, the "A" loop of the RHR System was returned to service.

A review of the process computer printout following this incident indicated that both an "A" and "B" reactor low water level trip signals (generated by level transmitters (EIIS:LT)_LT-101a and b) occurred simultaneously at 03:46:06 and cleared 7 to 9 hundredths of a second later. The "A" and "B" trip signals are gencrated from two '

instruments which share a common sensing line. Reactor water level recorder LR-96 (EIIS:LR) (input signal generated by LT-6-52a which shares a common sensing line with LT-101a and b)' trace from the time of the incident indicated a spike down to 0" (or below) reactor water level and a return to +29" reactor water level. Other reactor water level instrumentation in the Control Room indicated reactor level remained constant at the time the isolations occurred. The other reactor low water level trip signals ("C" and "D") were not received.

Immediately following the event, the Shift Technical Advisor (STA) made a visual inspection of the instruments associated with the isolation (LT-101a, LT-10lb, LT l 52a, and LT-6-52c) signal. No unusual conditions were found. Additionally, there was no work being performed on or around these instruments.

Cause of the Event The exact cause of the isolation signal is unknown. The downscale spike in reactor water level was exhibited on the level instrumentation on a common sensing line.

Instrumentation on other vessel sensing lines did not indicate a change in reactor level. It is postulated that Excess Flow Check Valve (EFCV) testing performed 10/22/89 on the low side sensing line for the affected level instrumentation may have introduced air into the line. It is believed that this air bubble progressed through the sensing line causing a lower than actual pressure signal to be sensed on the low side of the level instrumentation on that line, thus causing the output of the instrumentation to decrease (lower level indication). l

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The EFCV testing procedure, as it is currently written, establishes a flow path for lines to be tested by disconnecting certain fittings on the-instrument racks. This method of testing creates the potential for air to be introduced into the sensing

' lines during EFCV testing. Backfilling of instrument sensing lines is normally i performed near the end of each refueling outage after the majority of testing (including EFCV testing) and maintenance activities have been completed.

Backfilling of the sensing lines associated with the transmitters involved in this event had not yet been completed previous to the event. It is believed that air was introduced into the common low side sensing line for LT-101a, LT-10lb, LT-6-52a, and

'T-6-52c and was able to progress through the sensing line causing the transmitters  ;

to spike downscale before backfilling was completed.

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

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A Group 11 and III Isolation occurred. As a result of this isolation, the RHR System was out of service for approximately 50 minutes following the event. Reactor water temperature did not increase significantly as a result of the loss of RHR.

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I Had this instrumentation spike occurred at full power, a full reactor scram and similar PCIS isolation would have occurred as designed.

!- Corrective Actions i

The Group II and III Isolations were restored at 0435 hours0.00503 days <br />0.121 hours <br />7.19246e-4 weeks <br />1.655175e-4 months <br /> on 10/27/89. The RHR l . System was returned to service at 0440 hours0.00509 days <br />0.122 hours <br />7.275132e-4 weeks <br />1.6742e-4 months <br />.

The sensing line associated with this event was backfilled on 11/13/89.

The excess flow check valve testing procedure will be revised to require that the instrument sensing lines be opened via the existing drain lines on the instrument racks. This method of draining the sensing lines is easier to control, does not

, require the disconnection of any fittings on the instrument sensing lines, and l reduces the potential for air entering the system.

l Previous Similar Events  :

One previous similar event was identified which occurred on 10/20/89 and was reported in LER 3-89-005. This event involved a spurious Reactor Protection System actuation due to an instrumentation spike. The exact cause of the event described in LER 3 l 005 is unknown but is believed to be associated with excess flow check valve testing.

l Because the event reported in LER 3-89-005 is not believed to be attributed to air

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intrusion into instrument sensing lines, the corrective actions taken to prevent recurrence could not have been expected to prevent this event.

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