ML17262A217

LER 90-012-00:on 900926,turbine Trip Relay Actuation Due to Dropped Flashlight in Relay rack.W/901026 Ltr

ML17262A217
Person / Time
Site:	Ginna
Issue date:	10/26/1990
From:	Backus W, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-012, LER-90-12, NUDOCS 9011060419
Download: ML17262A217 (13)

Text

ACCELERATED DIS RIBUTION DEMONST TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9011060419 . DOC.DATE: 90/10/26 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME .AUTHOR AFFILIATION BACKUS,W.H. Rochester Gas 6 Electric Corp.

MECREDY,R.C. Rochester Gas & Electric Corp.

RECIP.NAME . RECIPIENT AFFILIATION

SUBJECT:

LER 90-012-00:on 900926,turbine dropped flashlight in relay rack.

trip relay actuation due to D

DISTRIBUTION CODE: 1E22T COPIES RECEIVED:LTR TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

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NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 A

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slee t / t Tears i.~fr sta'st ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER N. Y. 14649-0001 ROBERT C MECREOY TELEPHONE Vice President AAEACODEVte 546 2700 Clnna Nuclear Production

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October 26, 1990 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 90-012, Turbine Trip Relay Actuation Due to Dropped Flashlight in Relay Rack (Personnel Error),

Causes a Reactor Trip R.E. Ginna Nuclear Power Plant Docket No. 50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(iv), which requires a report of, "any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF) including the Reactor Protec-tion System (RPS)", the attached Licensee Event Report LER 90-012 is hereby submitted.

This event has in no way affected the public's health and safety.

Ve trul y rs, Robert C. Me redy XC ~ U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406.

Ginna USNRC Senior Resident Inspector 90 f ] 0!~04 i 9 90$ 02I PDR ATiOCV 0" t.l00264 PDC

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',"~'" Turbine Trip Relay Actuation Due to Dropped F1ashlight in Relay Rack (Personnel Error Causes a Reactor Tri

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SIAN 4l IlI lsl NTSNIOIIHI N.T SW Ollal LICSNSSS CONTACT tOA THH LS1 IISI NAIIS TSLStHONS NUTSO Wesley H. Backus AIISA COOS Technical Assistant to the ations Man 31552 -4446

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CAUSS SYSTSII SSANUt AC SIOATASLS SYSTSII COQIONSNT asANU ACI StOATAS TUAIA TO NtAOS TUN SA TO IHAOS 4?Svs&. s ss SUttLSISSNTAL AltOAT SÃtSCTSO IISI NONTH CAY YSAA SXtSCTSO

~ Us as I NI01 OATS Illl Y SS III tss. sssHNN SAASCTIO SUSasOSICUI OATSI assTAAcr IIJvvs  % lsoI NssN. IA, sAAsssssvNst INNse ~ vssvAHs ltrreettHs Aseat llsl On September 26, 1990, at 1100 EDST with the reactor at approximately 974 full power, a reactor trip occurred from an opening of the "A" Reactor Trip Breaker, followed in approximately seven (7) seconds by ' low pressurizer pressure reactor trip signal and the opening of the "B" Reactor Trip Breaker.

The '" Reactor Trip Breaker opening was caused by the inadvertent dropping of a flashlight on two of three turbine autostop trip relays. The low pressurizer pressure reactor trip was caused by the reactor coolant system cooldown due to the reactor being tripped with the turbine still on the line.

F Immediate corrective action was to stabilize the plant in hot shutdown. Corrective action to prevent recurrence will be based upon the recommendations of a Human Performance Enhancement System (HPES) evaluation of the dropped flashlight event. Corrective action for subsequent hardware malfunctions will also be taken.

UA. NUCLNAh RIOULATOhT C~l&KW UCENSEE EVENT REPORT {LER) TEXT CONTINUATION AttlKWEDOUO HO. 31iO&IOI 0 htlh N4."i/all&

tACILITYNets I I I COCKET ~lh QI Llh MLSlh IO tAOC Iil eTA1 QiQMCNTIAL )(p OIOII 1

R.E. Ginna Nuclear Power Plant o 5 o o o2 449 0 0 12 00 0 20F TlXT Idssae Nece 4 lese'ee eeeweeel AOC tete ~'tl IITI I ~ PRE= -P CONDITIONS The reactor was at approximately 974 steady state full power with no ma)or activities in progress. Electrical Control Configuration Drawing (ECCD) - personnel were performing electrical wire verifications in the RAl Relay Protection Rack.

DESCRIPTION OF EVENT A. Dates and approximate times of major occurrences:

.'I o September 26, 1990, 1100 EDST: Event date and time.

0 September 26, 1990, 1100 EDST: Discovery date and time.

0 September 26, 1990, 1100 EDST: Control Room operators verify both reactor and all control and shutdown tiip breakers rods inserted.

open September 26, 1990, 1105 EDST: Control Room operators closed both Main Steam Isolation Valves (MSIVs) to terminate plant cooldown.

0 September 26, 1990, 1115 EDST: Plant stabilized at hot, shutdown.

B. EVENT:

On September 26, 1990,:at 1100 EDST, with the reactor at approximately 974 full power, the Control Room received several annunciator alarms., Most notable of these alarms was the red first out annunciator alarm, K-2 (Rx Trip Breakers Open).

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M$. hUCLEAh hEOULATOhT C~IKIOH UCEN8EE EVENT REPORT ILER) TEXT CONTINUATION ASSAOVED OISE HO. 31ED&IOE EKSIKEE: EIBIIEE SACILITY NAIIE III DOCKET NLNNEh IEI LEh WANIh IN ~ AOE (N YEAA EEOUENTIAL SION 1 1 TEXT I/mar e nakeC ~ ~

R.E. Ginna Nuclear gower. Plant

~ IIIICSsrw ~WJ IITI 0 5 0-0 0 2 4 9 0 0 1 2 0 0 03 o~1 1 The Control Room operators immediately checked the reactor trip breaker position indicators and observed-that the >>A>> Reactor Trip Breaker was open and the >>B>>

Reactor Trip Breaker .was, still closed. The Reactor Trip Annunciator Panel was also checked at this time for the cause of the trip. There were no annunciators The Control Room operators immediately began verifying the immediate actions of procedure E-0 (Reactor Trip or Safety Injection) from memory as follows:

o At least one train of reactor trip breakers open;

- The >>A>> Reactor Trip Breaker was open.

I o Neutron Flux decreasing o MRPI indicates all control and shutdown rods on the bottom At. this time (i.e. approximately seven (7) seconds into the 'event), the Reactor Protection System received a reactor trip signal from low pressurizer pressure. The >>B>> Reactor Trip Breaker opened and the turbine was verified to be tripped.

It should be noted here that for the approximately seven (7) seconds, between the >>A>> Reactor Trip Breaker opening until the low pressurizer pressure reactor approximately trip, 80%

the power.

turbine was still This was due to the Turbine online at Emergency Trip (ET) solenoid valve not functioning

.properly when the ET relay received a trip signal

. from the >>A>> Reactor Tr'ip Breaker >>open>> contact. The

- 'low pressurizer pressure reactor trip and the opening of the >>B>> Reactor Trip Breaker subsecpxently tripped the turbine.

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~ OCENSEE EVENT REPORT {LER) TEXT CONTINUATlON UA. NUCLEAh hEOULATOhY COMEOENOH AtttKIVEO OME HO, g I EO~I OE ElltlhEE; E/SIIEE I'ACI LITT NAME I I I DOCKET ~Eh tlI LEh NVMEEh IEI ~ AOE IS TEAK g, ~ EQVENTIAL M  ??

1IOM M 1 R.E.'Ginna Nuclear Power Plant TEKT ~

IS~ M?tE??Eth ow ttEEAMtllthCPena ~'el IITI o so o o2 44 9 0012 .0 0 ov 11 With the turbine operating at approximately 804 power and the reactor shutdown (i.e. all shutdown and control rods inserted), a Reactor Coolant System (RCS) cooldown occurred, due to the heat removal imbalance. The level of both steam generators decreased below 164 narrow range level indication for, a short period of time. Level recovery was achieved through the operation of the turbine-driven and motor-driven auxiliary feedwater pumps. The RCS cooldown from the seven (7) second heat imbalance caused RCS temperature to decrease to approximately 535 F, which is about 10 F less than the temperature at stable hot shutdown conditions. The operators, believing a cooldown was still in progress five (5) minutes after the reactor shutdown, followed optional; procedural guidance, and closed both Steam Generator MSIVs. The closed'SIVs, coupled with the turbine trip, mitigated any remaining mechanisms for RCS cooldown, and the plant was stabilized in hot shutdown. This cooldown was also partially due to cooler water being fed to the steam generators by the Auxiliary Feedwater System.

Other equipment problems that occurred during the event were:

0 The "A" Steam Generator MSIV failed to fully close after receipt of an actuation signal. The valve external indicator 'revealed approximately one-quarter of an inch lack of travel from being fully closed. 'he valve subsequently closed approximately five minutes after signal receipt.

0 The Intermediate Range Nuclear Instrumentation, Channel N-35, after tracking -identical to Channel N-36, down to approximately 10 amps had its indication rapidly drop below 10 1~

amps. The N-35 channel returned to normal approximately ten hours following the trip.

The Control Room operators notified higher supervision and 'he Nuclear Regulatory Commission (NRC) of the event.

MAC fOAM EEEA lE 431

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~ \ UW NUCLSAII lllOULATOIIYCOMMNNON LICENSEE EVENT REPORT ILERI TEXT CONTINUATION AftAOVSO OMS rNO 3150W104 S)ttllISS SISI/SS FACILITYNANO III OOCIIKT NUMCKII ISI LSII NUiNSN ISI ~ AOS ISI SSOUSNTIAL '. >

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R.E. Ginna Nuclear Power Plant N OJ45asV NIIC AnM ~SI I Ill o 5 o o o 2 4 4 0 0 1 2 000 5 OF C~ INOPERABLE STRUCTURES ~ COMPONENTS / OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

None D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None E~ METHOD OF DISCOVERY:

The event was immediately apparent due to alarms and indications in the Control Room.

F. OPERATOR ACTION:

After the reactor trip, the Control Room operators performed the actions of Emergency Operating Procedures E-O, (Reactor Trip or Saf ety Zng ection) and ES-O. l, (Reactor Trip Response) and stabilized the plant.

The MSZVs were closed approximately five (5) minutes

'after the trip to prevent further plant cooldown.

G. SAFETY SYSTEM RESPONSE:

None IIZ. CAUSE OF EVENT A. IMMEDIATE CAUSE:

The'eactor trip occurred due to the "A" Reactor Trip Breaker opening.

B. INTERMEDIATE CAUSE:

The "A" Reactor Trip Breaker opening was due to the bumping, of two turbine autostop trip (AST) relays in the RAl Relay Rack. This bumping occurred from a smail flashlight (powered by "AA" batteries) that was 1 NAC AOAM SSSA I$ 4$ l

NRC eerro 9$ 9A \~ US. NUCLEAR REOULATORY COMMIEEION (9491 LICENSEE EVENT REPORT ILERI TEXT CONTINUATION 'eehovEo oME'No. $ 1$ C~ICI EX91RE$ "9191/9$

eACILITY NAME I'l OOCKET NUMOER QI LER NINER IEI YEAR I9 QU 9 N'Fl A L 4 M O OEVrorOH 4 M H.E. Ginna Nuclear Power Plant TExT II1 more RMoe o terhreerE

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o s o o o 2 4 4 9 0 1 2 00 06 OF oee Iroerenor Hrrlc Fdic@ 99EAOI I ITI being used by an ECCD person.. While performing electrical wire verifications, the flashlight was accidently dropped and fell onto the relays. A normal actuation of the two relays would result in

,

actuation of both trains of logic '(i.e. opening of both reactor trip breakers and subsequent turbine trip) as each relay has an "A" and "B" train contact.

The postulated cause for actuation of only one .train of logic, ("A" Train Reactor Trip Breaker) is that the bumping effect was such that localized chatter of only the "A" train contacts occurred. The logic for.,

reactor trip is the actuation of at least two out of the three AST relays. Later testing confirmed that,:-

if normally actuated, the "B" train trip signal AST.

circuit would operate properly.

C. ROOT CAUSE:

The accidental dropping of the flashlight that bumped the relays was a personnel error.

IV ANALYSIS OF EVENT This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(iv), which requires reporting of "any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System (RPS)," in that the "A" Reactor Trip Breaker opening and subsequent Low Pressurizer Pressure Reactor -Trip was an automatic actuation of the RPS.

An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:

o When initiated by an input signal, the two reactor trip breakers opened as required.

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROYfO OM9 NO. 3194M10e f RPIRf 9: 9l9 I I99 I'ACILITYNAMf 111 OOCKfT NUM9 fR Ill Lfh NUMhfh IQ PA4f 191 IIQUANNTIAL YNIOS H N Ginna Nuclear .Power Plant 244 90

~ R.E.(SsNN NNee N Nfsleh eM eAWosel WIC Pens ~'II11Tl o s o o o 1 2 00 07 o~ 1 1 o All control and shutdown rods inserted to shut the.

reactor down as designed.

o The plant was quickly stabilized in hot shutdown.

This transient was compared to 'ncrease in decay heat removal transients in the Ginna Updated Final Safety Analysis (UFSAR). None of the assumptions of the UFSAR were violated during the event. The resultant cooldown from the reactor trip -and the turbine remaining on line caused pressurizer pressure to decrease rapidly. The rate sensitive low pressurizer pressure trip caused' reactor trip, opening the "B" Reactor Trip Breaker. The opening-:

of the "B" Reactor Trip Breaker caused the turbine to trip.

The following factors led to the cooldown, as compared with the cooldown expected following a normal reactor trip:

o Failure of the 20 ET solenoid to trip the turbine when the "A" Reactor Trip Breaker opened. Reactor trip without turbine trip caused the RCS to cooldown to approximately 535 F.

o . Lo Lo level on both steam generators caused the turbine driven auxiliary feedwater pump to start.

Steam extracted from the header by the turbine driven auxiliary feedwater pump contributed to the cooldown.

o Total auxiliary feedwater flow of greater than 470 gallons per minute per steam generator-also contributed to the cooldown.

Due to the above circumstances, the cooldown of the RCS to approximately 535 F is not unexpected. This cooldown is bounded by plant accident analysis and does not exceed the technical specification limit of 100oF per hour. Additional protection is provided by closure of the MSZVs.

Based on the above, it can be concluded that the public health and safety was assured at all times.

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19491

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L'ICENSEE EVENT REPORT (LERI TEXT CONTINUATION AttAOVSD OSIS NO JISOWIOS tlItIASS: SQI ISS tACILITYAASSS Ill OOCKST IIU~SA ITI LtA IIUSSSSA Itl ~ Aot ISI

~ SOVSNTIAL VISION N V II V N R.E. Ginna Nuclear%?ower Plant TECT II'INSIS DISCS ~ ISSVSSA SASSSASI lYAC ANN ~'Sl IITI 0 5 0 0 0 2 4 4 9 0. 0 1 2 0009 OF SSS valve shaft due 'o the valve packing.

MSZV's were stroked several times to ensure Both aperability and adequate closure capability.

Results of these tests, support the conclusian that failure of the "A" -MSZV to fully seat during the plant trip was not due to internal valve distortion and bending, but was the result of a lack of flow across the valve disc.

Failure to close is attributed to .the closure operation occurring in a quiescent environment.

Valve closure was dependent solely upon two factors: the moment developed by the weight of the valve disc, and the spring pravided to,-

assist in,valve closure. Without the additional assistance of steam flow across the valve disc; the valve was not capable af completing its closing operation. Based upon valve seat area, a one (1) psi differential across the valve seat would develop a moment oi approximately 450 ft-lbs. This moment is comparable to the moment developed by the weight of'he valve disc in its closed position. For all design basis accidents where MSIV closure is required, the accident transient would a large enough differen-tial pressure todevelop obtain complete valve closure.

We are evaluating various packing materials which have a low friction coefficient and can perform the sealing function.

0 As the Intermediate Range NZS Channel N-35 tracked NIS Channel N-36 for its normal operating range and returned to normal'pproximately ten (10) hours after the trip, no .immediate .action was deemed necessary. Specific logarithmic amplifier idle current adjustments internal to the PC card are being considered to correct NIS Channel N-35 response below 10 10 amps. Westing-house personnel were contacted, and they confirmed potential impact af the amplifier idle 'current on channel autput, and that this output does not affect the safety function of the Intermediate NAC SOAIS SSSA 19491

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