05000352/LER-1990-025-01, :on 901110,spurious LOCA Signal Resulted in ESF Actuations

:on 901110,spurious LOCA Signal Resulted in ESF Actuations

ML20024F736
Person / Time
Site:	Limerick
Issue date:	12/07/1990
From:	Madsen G, Mccormick M PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-025-01, LER-90-25-1, NUDOCS 9012120294
Download: ML20024F736 (6)

LER-1990-025, on 901110,spurious LOCA Signal Resulted in ESF Actuations

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv), System Actuation
3521990025R01 - NRC Website

text

_ - - _ - - - _ ______-__. _ _ _.

10 CPR 50.73 PHILADELPHIA ELECTRIC COMPANY LIMERICK GENER ATING ST ATION P. O.DO X A SAN ATOG A, PENNSYLV ANI A 19464 Docket Nos. 50-352 (216) 3271200 ext. 2000 50-353 u. t ucconu;c q.ja., (c.

License Nos. NPF-39 NPF-85

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

SUBJECT:

Licensee Event Report Limerick Generating Station - Units 1 and 2 I

This LER reports various Engineered Safety Feature actuations resulting from a spurious Division 1 Loss of Coolant Accident (LOCA) signal.

The spurious LOCA signal occurred while an Instrumentation and Controls technician was performing an incorrect surveillance test procedure.

Reference:

Docket Nos. 50-352 50-353 Report Number:

1-90-025 Revision Number:

0 Event Date:

November 10, 1990 Report Date:

December 7, 1990 Facility:

Limerick Generating Station P.O.

Box A, Sanatoga, PA 19464 This LER is being submitted pursuant to the requirements of 10 CPR 50.73(a)(2)(iv).

Very truly yours, y

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T. Martin, Administrator, Region I, USNRC T.

J. Kenny, USNRC Senior Resident Inspector, LGS h

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ExPints r3i n 8 ACILITY N AME fil DOLElf NUMBE R 431 PAut 6 h Limerick Generating Statton, Unit 1 0 [5101010 l3 l512 1 lOFl 0l 5

'nrL in Spurious LOCA Signal Causes Engineered Safety Feature Actuations

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.On November 10, 1990, with Unit 1 in a refueling outage, an Instrumentation and Controls (I&C) technician was performing Surveillance-Test (ST) procedure ST-2-036-680-1, "Drywell Pressure Transmitter Sensing Line Blowback Procedure."

A

- spurious Division 1 Loss of Coolant Accident (LOCA) signal occurred when he opened the LO side isolation valve to restore differential pressure switch PDS-59-106A.

Trapped pressurized air was then released down the instrument line when the LO side and two drywell pressure transmitters sensed the high pressure and initiated the LOCA signal.

The spurious LOCA signal caused automatic actuations of.a number of Engineered Safety Features which functioned as designed.

The actuations included a start of the

'lA' Core Spray (CS) pump with injection, opening of the lA' Low Pressure Coolant. Injection valve, a start of the Dll Emergency Diesel Generator, and tripping of the

'0A' Residual Heat Removal Service Water pump.

The consequences of this event were minimal.

Reactor' vessel inventory increased approximately 8 inchee due to the CS injection and reactor coolant temperature increased approximately 3 degrees due to termination of the shutdown cooling function The cause of the event was an incorrect procedure.

The I&C technician correctly performed the procedure as written.. The procedure will be revised to prevent similar future occurrencese A warning label will also be added to PDS-59-106A, PDS-59-106B, PDS-59-206A, and PDS-59-206B to alert I&C technicians of their unusual valving sequence, y,C,,,

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Unit 1 Operating Condition was 5 (Refueling) at 0% power level with the Reactor Pressure Vessel (RPV) pressure at 0 psig and the reactor cavity water level at 202 inches or 363.-Inches above the top of" active fuel.

Unit 2 Operating Condition was'l (Power Operation) at 100% power

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level.

The

'lA' Residual Heat Removal (RHR) system (EIIS BO) was in service in the shutdown cooling mode of operation.

The common

'OA' Residual' Heat Removal Service Water (RHRSW) system (EIIS:BI) pump was also in service providing shutdown-cooling to Unit 1.

There were,no structures, systems or components out of service which contributed to this event.

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' Description of the Event:

On-November 10,-1990, a: contractor employed Instrumentation and Controls (I&C) technician was performing Unit 1 Surveillance Test

-(ST) procedure ST-2-036-680-1,-"Drywell Pressure l Transmitter Sensing Line Blowback Procedure."

At 2232 hours0.0258 days <br />0.62 hours <br />0.00369 weeks <br />8.49276e-4 months <br /> he opened the LO side isolation valve to restore differential pressure switch-

'PDS-59-106A (see Figure 1) which resulted in a spurious Division 1 foss of Coolant-Accident _(LOCA)-signal.

The spurious LOCA r

sisnalcoccurred.when greater than-1.68 psig was-momentarily sensed by pressure transmitters PT-42-1N094A and PT-42-lN094E along -with the-RPV pressure below :455 psig.

A half scram signal was also generated by pressure, transmitter PT-42-lN050A due-to

- the spurious high drywell-pressure condition. _Various Engineered Safety Feature (ESP) actuations were initiated.in addition to

- control room' annunciation.

The Division'l LOCA-signal caused the following-automatic actions, as designed:

'lA' Core Spray (CS) (EIIS:BM) pump started =and injected suppression pool water into'the RPV,-

'lA' Low Pressure Coolant _ Injection (LPCI) injection value opened, D11/ Emergency Diesel Generator (EDG) (EIIS:EK)Estarted and the 1 associated Division 1 AC Safeguard power bus load shed occurred, and-

' Common '0A' Emergency. Service Water (ESW) '(EIIS:BI) pump started.

The load shed' caused the

'0A' RHRSW pump to trip.

Operators verified that the_ Division 1 LOCA signal was spurious, secured the

'1A' CS pump to terminate the RPV injection, closed the

'1A' LPCI--injection valve, and realigned the

'1A' RHR system to the

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The D11 EDG and

'OA' ESW pump were then secured.

Operators also reset the LOCA signal.

At 2248 hours0.026 days <br />0.624 hours <br />0.00372 weeks <br />8.55364e-4 months <br /> operators started the 'OC' RHRSW pump in the

'A' loop of the RHRSW system which re-established shutdown cooling on Unit 1.

Unit 2 operation was unaffected by the activities involving the common systems.

A four (4) hour notification was made to.the NRC at 0117 hours0.00135 days <br />0.0325 hours <br />1.934524e-4 weeks <br />4.45185e-5 months <br /> on November 11, 1990, in accordance with 10CPR50.72(b)(2)(ii) because the event resulted in the automatic actuations of various ESP.

This report is being submitted in accordance with the requirements of 10 CPR 50.73(a)(2)(iv).

Analysis of the Event

The consequences of this event were minimal.

Procedure ST-2-036-680-1 is only performed during refueling operations so that this event could not have occurred under normal power operation.

There was no release of radioactive material to the environment as a result of this event.

RPV inventory increased approximately 8 inches from 202 inches to 210 inches while the

'1A' CS system injected for approximately one minute.

During the 15 minutes that shutdown cooling was not in service, reactor coolant temperature increased approximately 3 degrees from 114 degrees to 117 degrees.

The maximum reactor coolant temperature allowed by Technical Specifications (TS) is 140 degrees while in Operating Condition 5.

Operations personnt i had ample time to restore shutdown cooling prior to exceeding the TS limit since it would have taken over two hours to reach the temperature limit.

The redundancy in the common ESW and RHRSW systems made the consequences of this event minimal to Unit 2.

The portions of the ESW and RHRSW systems unaffected by this event were available to satisfy all Unit 2 requirements.

Cause of the Event

The cause of the event was an incorrect procedure.

Procedure ST-2-036-680-1 is performed during a refueling outage to ensure that instrument sensing lines which penetrate the primary containment are clean and unobstructed.

The pressure transmitters are isolated during the blowback operation.

The I&C technician correctly followed ST-2-036-680-1 in restoring the isolated pressure transmitters.

PT-42-lN094A and PT-42-lN094E were restored satisfactorily (see Figure 1).

PDS-59-106A was the last pressure instrument restored.

Just prior to restoration, the HI and LO side isolation valves were closed and the equalizing valve was open.

The HI side line to PDS-59-106A is connected to the Primary Containment Instrument Gas system which was in service operating at 100 psig pressure.

Per the ST

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..c, mm v em procedure, the I&C technician opened the HI side isolation valve to PDS-59-106A pressurizing the instrument line to 100 psi up to L

the LO side' isolation valve.

The equalizing valve was then closed.

Finally, the LO side isolation valve was opened which released the pressurized air trapped between the equalizing valve and LO side-isolation valve down the instrument'line such that,

PT-42-lN094A and PT-42-lN094E momentarily sensed the high

pressure-and-initiated a_-Division 1 LOCA-signal.

Corrective Actions

Procedure ST-2-036-680-1 will be revised to restore PDS-59-106A-and PDS-59-106B prior to restoration of other pressure transmitters.

This revision prevents the other pressure transmitters-from sensing momentary high pressure in the instrument-line and-actuating various ESF.

In additlon, the-i valving sequence to= restore the pressure _ switches will be revised

- to'first open the LO-side _ isolation valve.. The equalizing valve

isLthen: closed.

Finally, _the HI side isolation valve is opened.

This new valving sequence wi'll prevent-trapping pressurized air on the-LO side of the' pressure switch and releasing lt down the in'strument.line. :An additional-caution will be added to the procedure..to-alert the_I&C technician of the possibility of generating a LOCA signal.

Unit 2 procedure ST-2-036-680-2 will receive-similar revisions. 'The other procedures which isolate andJrestore_.these pressure-. switches _were reviewed and contained the correct valving sequence...'.A warning label will also be added

'to PDS-59-106A,-PDS-59-106B,'PDS-59-206A, and PDS-59-206B to alert I&C-technicians-of their unusual valving _ sequence.

This-

will-aid personnel performing work on these pressure instruments for-reasons other than_-sensing line blowback procedures,-such as'

- maintenance-activities.

.A review was performed and did not identify any similar configurations where a deviation from the normal pressure instrument valving sequence was required'to avoid-any' adverse conditions'.

Previous Similar Occurrences:

'LER 1-85-037',-1-85-040, 1

1 87-019, and'l-87-042 reported ESP

_actuations_.resulting_from spurious LOCA signals.

Each occurred

' as a result of a personnel error involving failure to correctly follow procedures.,The corrective actions for these previous occurrences could not:have prevented-the-spurious LOCA signal and ESP'actuationsfreported in this LER.

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