ML17285A601

LER 89-023-00:on 890531,ESF Isolations & Actuations Occurred Due to Loss of Reactor Protection Sys Bus During Testing. Caused by Personnel Error.Logic Sys Functional Test Procedure Revised.Test Engineer counseled.W/890630 Ltr

ML17285A601
Person / Time
Site:	Columbia
Issue date:	06/30/1989
From:	Arbuckle J, Powers C WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-023, LER-89-23, NUDOCS 8907100048
Download: ML17285A601 (8)

Text

gc CP~RATED ACCESSION NSR:8907100048 D1 <i~BUTION

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DEMO?i STRATI ON REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

DOC.DATE: 89/06/30 NOTARIZED: NO

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SYSTEM DOCKET. 4 FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION.

ARBUCKLE,J.D'. Washington Public Power Supply System

'" Washington Public Power Supply I

POWERS,C.M. System RECIP.NAME RECIPIENT AFFILIATION ~

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SUBJECT:

LER 89-023-00:on 890531,ESF isolations & actuations due to

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loss of reactor protection sys bus during testing.

W/8 DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SI Zg:

TITLE: 50;73/50.9 Licensee'vegt 'Report,(LER), Incident, Rpt, .'etc;.; '

r NOTES RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D PD5 LA 1 1 PD5 PD 1 1 SAMWORTH,R 1 1 I INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 '

ACRS WYLIE 1 1 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 DEDRO 1 1 IRM/DCTS/DAB 1 1 NRR/DEST/ADE SH 1 1 NRR/DEST/ADS 7E 1 0 NRR/DEST/CEB SH 1 1 NRR/DEST/ESB SD 1 1 NRR/DEST/ICSB 7 1 1 NRR/DEST/MEB 9H 1 1 NRR/DEST/MTB 9H 1 1 NRR/DEST/PSB 1 1

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1. 1 NRR/DLPQ/HFB 10 1 1 NRR/DLPQ/PEB 10 1 1 NRR/DOEA/EAB 11 1 1 2 2 NUDOCS-ABSTRACT 1 1 G,FILE OM 1. 1 RES/DSIR/EIB 1 1 R 8TDSR/PRAM 1 1 RGN5 FILE 01 1 1 EXTERNAL: EG&G WILLIAMS,S 4 4 FORD BLDG HOY,A 1 1 I L ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC MAYS,G 1 1 5 NSIC MURPHY,G.A 1 1

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WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968 ~ 3000 George Washington Way ~ Richland, Washington 99352 Docket No. 50-397 June 30,.1989 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

NUCLEAR PLANT NO. 2 LICENSEE EVENT REPORT NO.89-023

Dear Sir:

Transmitted herewith is Licensee Event Report No.89-023 for the WNP-2 Plant.

This report is submitted in response to the report requirements of 10CFR50.73 and discusses the items of reportability, corrective action taken, and action taken to preclude recurrence.

Very truly yours, C.M. o rs (M/D ant Manager 927M)'NP-2 Cf<P: 1 g

Enclosure:

Licensee Event Report No.89-023 cc: Mr. John B. Martin, NRC - Region V Mr. C.J . Bosted, NRC Site (tl/D 901A)

INPO Records Center - Atlanta, GA Ms. Dottie Sherman,'NI Mr. D.L. Williams, BPA (ff/D 399) 8<07100a48q 8906='0 7 POP /DOCK 8c,O003". pDC C;

NAC Form 388 U.S. NUCLEAR REGULATORY COMMISSION (84)3)

APPAOVEO OMB NO. 3(504))04 LICENSEE EVENT REPORT {LER) EXPIRES; 8/31/88 FACILITY NAME (II DOCKET NUMBER (2) PAGE 3 Washin ton Nuclear Plant - Unit 2 0 5 0 0 03 97 1 OFO 6

"'"("Engineered Sa ety eature E so atsons an Actuat>ons ue to Loss of Reactor Protection System (RPS) Bus During Testing - Personnel Error/Procedural Inadequacy EVENT DATE (SI LER NUMBER (5) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH OAY YEAR YEAR rrCgr. SEOUKNTIAL vyx RSYlslore MONTH OAY YEAR FACILITYNAMES DOCKET NUMBERIS)

NUMeER ,'I;'9 IIUMeeR 0 5 0 0 0 0 5 3 1 8 9 9 '023 0 06 308 9 0 5 0 0 0 OPERATING THIS REPORT IS SUBMITTED PURSUANT T 0 THE REOUIREMENTS OF 10 CF R (): (Cnrec One Or'mOrr Oi tnr iOiiOwinpi (Ill MDDE (8) 20A02(el 20.405(c) 50,7 3(e) (2) (iv) 73.71(e)

POWER 20.405( ~ ) (I I (0 50.38(cHII 50.73(e) (2 l(v) 73.71(cl LEYEL 0 0 0 20 405(e) (I I(ill 50.35(cH2) 50.73(e) I 2)(viiI OTHER ISpeciiy in Aettrect Oeiow end in Fret, Ni(C FOrm Ceps(: 6' 20.405( ~ l(1 l(nil 50.73(el(2HII 50,73(e l(2)(vill) (A) 36IFAI 143.'

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MANVFAC COMPONENT MANVFAC REPORTABLE CAVSE SYSTEM COMPONENT TURER TO NPRDS CAUSE SYSTEM TURER c~rr@I:)y':(y;rg)g(rior TO NPRDS

')y~gpj@~44p'UPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED SUBMISSION DATE I15l YES iiiyeL comoirtr EXPECTED SVBeerSSION DATEI NO ABSTRACT (Limit to 1400 rpecrr. i,r., epprommetrry Aiteen tinpot corer typewrinrn immi (18)

On May 31, 1989 at 1406 hours0.0163 days <br />0.391 hours <br />0.00232 weeks <br />5.34983e-4 months <br /> an Electrical Protection Assembly (EPA) Breaker (RPS-EPA-3E) tripped causing a loss of power to Reactor Protection System (RPS) Bus B.

Loss of power to RPS Bus B caused a half-scram in RPS Division B and multiple primary and secondary containment isolations and Engineered Safety Feature (ESF) actuations of ventilation systems. At the time of the event the Plant was shutdown for the annual maintenance and refueling outage.

The loss of RPS B power causes Nuclear Steam Supply Shutoff System (NSSSS) Containment Inboard and Outboard Isolations for Groups 1,2,5,6= and 7; and a Reactor Building Exhaust Plenum Radiation Monitor HZ" signal [a non-NSSSS ESF signal] which initiates several ESF actuations including the Standby Gas Treatment (SGT) System, the Control Room Emergency Filtration System, and a Reactor Building Ventilation System isolation. Plant Operators responded by restoring all systems, -including Residual Heat Removal (RHR) Shutdown Cooling, to pre-event lineup status by 1430 hours0.0166 days <br />0.397 hours <br />0.00236 weeks <br />5.44115e-4 months <br />.

The causes of this event are 1) personnel error in that a Plant Test Engineer and Plant Operators did not adequately review the consequences of starting a Reactor Recirculation Pump (RRC-P-lA) while the plant was aligned for the performance of Logic System Functional Testing (LSFT) of the ATWS Recirculation Pump HAA Trip System, and 2) inadequate procedure in that the LSFT procedure did not specifically caution against starting an RRC pump during test performance. The LSFT alignment requires lifting a lead in the logic circuitry for the 60 Hz power supply to RRC-P-1A. The lifted lead defeated tt)e normal pump starting sequence (60 cycle start power automatically shifting to a 15 NRC Form 358 le 83 I

NRC Form 388A U.S. NUCLEAR REOULATORY COMMISSION (ww LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPROVED OMS NO. 3(SOWlnd EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (8) PACE (3)

YEAR SEOUENTIAL 'N~ REVISION NUMSEII '.SC'UMBER Washin ton Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 B OF TEXT ///moro Enoco /F ror/uinod, u>> oddi5nrMI NRC Fonrr SSSAS/ (IT)

Abstract (continued) cycle source). With the lead lifted, the down shift sequence could not occur and the pump remained at 60 Hz until plant operators tripped the pump from the control room. The large load starting and remaining on 60 Hz power longer than normal caused under voltage transients on the power supply of sufficient magnitude to trip the EPA breaker and cause the actuations and isolations.

Corrective actions consist of counselling the individual involved, revising the LSFT procedure to indicate the consequences of lifted leads, and performing an evaluation of those ESF actuations which have occurred during the recent maintenance and refueling outage.

There is no safety significance associated with this event. No actual plant conditions requiring the Engineered Safety Feature isolations and actuations existed, and all isolations and actuations occurred as designed.

Plant Conditions a) Power Level - OX b) Plant Mode - 5 (Refueling)

Even't Descri tion On May 31, 1989 at 1406 hours0.0163 days <br />0.391 hours <br />0.00232 weeks <br />5.34983e-4 months <br /> an Electrical Protection Assembly (EPA) breaker (RPS-EPA-3E) tripped causing a loss of power to Reactor Protection System (RPS) Bus B. The loss of power on RPS Bus B caused a half-scram in RPS Division B and multiple primary and secondary containment isolations and Engineered Safety Feature (ESF) actuations of ventilation systems. At the time of the event the Plant was in a shutdown condition for the annual maintenance and refueling outage.

The loss of RPS Bus B power causes Nuclear Steam Supply Shutoff System (NSSSS)

Containment Inboard and Outboard Isolations for Groups 1 (Main Steam Line Drain Valves only), Group 2 (Reactor Mater Sample Valves), Group 5 LResidual Heat Removal (RHR) and Traversing In-Core Probe (TIP) Systemsj, Group 6 (RHR Shutdown Cooling),

and Group 7 [Reactor Water Cleanup (RWCU) System3. At the time of the event, both the TIP and RMCU Systems were already out of service for maintenance.

In addition, the loss of RPS B power causes an NSSSS Group 3 (Primary and Secondary Containment Ventilation and Purge System) and partial Group 4 [Miscellaneous Balance of Plant (6-Valves)j isolation. These isolations occurred because RPS Bus B is the power supply for Reactor Building Exhaust Plenum Radiation Monitors (Channels B and D). Loss of RPS B power de-energizes these monitors, causing a "Z" signal - a non-NSSSS ESF trip signal. All required Group 3 and 4 actions occurred as designed, including the automatic start of the Standby Gas Treatment (SGT) System and the Control Room Emergency Filtration System', and a Reactor Building HVAC Isolation.

NIIC FORM nddA ;c.s. Cpni Ipsd-s2n-sdp nnn1n 1983)

NRC Form 38SA (9431 U3L NUCLEAR REOULATORY COMMISSION LICENSEE EV REPORT ILER) TEXT CONTINUATIO APPROVEO OMS NO. 3150M(04 EXPIRES: 8/31/88 FACILITY NAME (1) OOCKET NUMBER (3(

I.ER NUMBER (Bl PAGE (3) 81 SEOUENTIAL >Ps'o REVISION NUM ER 'r/Sr NUMSER Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 8 9 0 2 3 0 0 0 4 OF 0 6 TEXT ///more s/rsce /s rer/Ir/red, rrse sd //I/oos/HRC FINm 3//I/A'/ (17}

Further Evaluation and Corrective Action A. Further Evaluation

1. This event 's reportable under 10CFR 50.73(a)(2)(iv) as "an event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System (RPS).U

2. Had RPS-t%-1B been available and powering RPS Bus B, flywheel energy in the HG set would have carried the bus through the voltage transient and precluded the event. Alternate power was being supplied through Station Normal Transformer TR-N2 and Startup Transformer TR-S was out of service for maintenance. There were no other structures, components, or systems that were inoperable at the start of the event that contributed to the event.

3. The root causes of this event are as follows:

a) Personnel Error: Both the Test Engineer and Control Room Operators failed to fully investigate and understand the impact of starting RRC-P-lA while the plant was aligned for the LSFT. Had the implications'f lifting the lead been recognized the plant operators would not have attempted to start the pump.

b) Procedural RRC-P-lA Inadequacy: When asked by the Control Room Operators could be started, the Test Engineer reviewed the LSFT if procedure for precautions applicable to the situation and found no explicit precautions. The procedure did state that the "system being tested shall be out of service during the performance of that portion of the test" but the Test Engineer believed that this referred only to the trip of RRC-P-lA on an ATWS signal and did not consider mean a loss of high speed trip function on starting the pump.

it to As a result, the Test Engineer determined that a pump start was not prohibited and advised the Control Room Operators to proceed with the RRC pump start.

B. Further Corrective Action

1. 'he Test Engineer invol ved was counseled on the importance r

of fully understanding the implications of test line-ups prior to providing guidance on the advisability of plant operations.

2. Logic System Functional Test procedures are being revised to indicate the consequences of lifted leads and installed jumpers. In the case of the ATWS Recirculation Pump Trip System LSFT, this change has been made.

3. Guidance has been provided to Plant Operations from Plant Technical on restricting the manipulations of large electrical loads while on RPS alternate power.

NRC FORM ESSA ~ U.S. CPOr l9SS-530-509i00010 (9 831

NRO Form 366A U.S. NUCLEAR REOULATORY COMMISSION (943)

LICENSEE EVE REPORT ILER) TEXT CONTINUATION APPROVED OMS NO. 3(SOW(04 EXPIRES: 8/31/68 FACILITY NAME (1) DOCKET NUMBER (21 LER NUMBER (6) PACE (3)

SFOVSNTIAL rgyP REVISION NUMBER NVMSSR Washin ton Nuclear Plant - Unit 2 o s o 0 0 3 9 7 OF TEXT /// more eeeee /e Ter/Ir/red. Iree edd/0'or>>///RC Form 3//54'e/ (17)

4. Applicable Plant operating procedures will be reviewed to consider the

'ecessity to provide precautions against starting large loads whenever the RPS Alternate Power Supply is being utilized.

5. An overall evaluation is currently being performed for those ESF actuations which have occurred during the recent maintenance and refueling outage; As part of this evaluation, changes to the design of the alternate. power supply will be considered to make the power supply less susceptible to voltage or frequency transients.

=

Safety Si nificance There is no safety signficance associated with this event because no Plant condition requiring the ESF isolations and actuations existed, and all =ESF acutations occurred as designed.

In adaition, at tne time of the event reactor water level was greater than 22 feet above the reactor vessel flange with the fuel pool gate removed which provided a large heat sink for core cooling. Plant Operators responded by restoring all systems,,including RHR Shutdown Cooling, to pre-event lineup status within 24 minutes.

Accordingly, this event posed no threat to the health and safety of either the public or Plant personnel.

Similar Events None EIIS Information Text Reference EIIS Reference Sys tern Component EPA Breaker (RPS-EPA-3E) JC BKR Reactor Protection System (RPS) JC RPS-Bus-B JC Nuclear Steam Supply Shutoff System (NSSSS) BD Reactor Building Exhaust Plenum Radiation Monitor IL MON Standby Gas Treatment (SGT) System BH Control Room Emergency Filtration System VH Reactor Building HVAC VA Residual Heat Removal (RHR) System BD Reactor Recirculation Pump (RRC-P-lA ) AD p Main Steam Line Drain Valves SN LOV NRC FORM SSSA ~ V.S. OPOI )966-520 589r00010 (943)

NRC Forrrr 3$ 6A U.S, NUCI.EAR REOULATORY COMMISSION 1943)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMB NO. 3(60W104 EXPIRES: 6/31/BB FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PACE (3)

YEAR SEQUENTIAL REVISION NUM ell ..Ysr NUMBER hin ton Nuclear Plant - Unit 2 0 6 0 0 0 9 7 8 9 0 2 3 0 0 0 6 OF 0 TEXT /I/more Fosse /s rer)vtred, vse sdr//I/ons///RC Form 366A's/ ((7)

EIIS Information Text Reference EIIS Reference System Component Reactor Water Sample Valves AD I SV Traversing In-Core Probe (TIP) System IG Reactor Water Cleanup (RWCU) System CE Containment Ventilation and Purge System VH RPS NG Set (RPS-MG-1B) JC NG Breaker RRC-CB-PlA/RPT3 (RPT-3A) AD BKR Breaker RPT-2A AD BKR Transformer TR-N2 EA XFt/IR Transformer TR-S, EA XFMR NRC FORM 35OA <<0 ~ 6 ~ CPOI 1906 620 569r00010 (943)