ML17285A559

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LER 89-016-00:on 890514,reactor Operator Inadvertently Tripped Div 1 & 2 Offsite Power Supply Feeders Resulting in Loss of Power to Buses SM-7 & SM-4.Caused by Inadvertent Removal of Transformer Fuses.Fuses reinserted.W/890613 Ltr
ML17285A559
Person / Time
Site: Columbia Energy Northwest icon.png
Issue date: 06/13/1989
From: Arbuckle J, Powers C
WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-016, LER-89-16, NUDOCS 8906200200
Download: ML17285A559 (9)
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Text

AC CRAP>TED Dl Bt 'Tl ON DEMO~ > TV '104 SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8906200200 DOC.DATE: 89/06/13 NOTARIZED: NO DOCKET I FACIL!50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION ARBUCKLE,J.D. Washington Public Power Supply System POWERS,C.M. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 89-016-00:on 890514,ESF sys actuation caused by inadvertent fuse removal & loss of power due to IFL/IT/PE.

W/8 1tr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD5 LA 1 1 PD5 PD 1 1 SAMWORTH,R 1 1 INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 ,D 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 8H 1 1 NRR/DEST/ADS 7E 1 0 NRR/DEST/CEB 8H 1 1 NRR/DEST/ESB 8D 1 1 NRR/DEST/ICSB 7 1 1 NRR/DEST/MEB 9H 1 1 NRR/DEST/MTB 9H 1 1 NRR/DEST/PSB 8D 1 .1 NRR/DEST/RSB 8E 1 1 NRR/DEST/SGB 8D 1 1 NRR/DLPQ/HFB 10 1 1 NRR/DLPQ/PEB 10 1 1 NRR/DOEA/EAB 11 1 1 NRR/J)REPJRPB 10 2 2 NUDOCS-ABSTRACT 1 1 1 1 RES/DSIR/EIB 1 1 RES/DSR PRA 1 1 RGN5 FILE 01 1 1 EXTERNAL: EG&G WILLIAMS,S 4 4 FORD BLDG HOY I A 1 1 R L ST LOBBY WARD 1 1 LPDR 1 1 I

NRC PDR 1 1 NSIC MAYS,G 1 1 NSIC MURPHY,G.A 1 1 NIXIE 'IO ALL "RIDS" RECIPIENIS PIZASE HELP US IO RHXJCE WASTE! CONTACT 'IHE DOCUNEPZ CGNZRDL DESK, KXN Pl-37 (EXT. 20079) 'XO KZiZMZNATR RÃ3R NAME FBCH DISTEUBVTIGN LISTS H)R DOCUKWZS YOU DGNIT NEEDt TOTAL NUMBER OF COPIES REQUIRED: LTTR 43 ENCL 42

a~i WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968 ~ 3000 George Washington Way ~ Richland, Washington 99352 Docket No. 50-397 June 13, 1989 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

NUCLEAR PLANT NO. 2 LICENSEE EVENT REPORT NO.89-016

Dear Sir:

Transmitted herewith is Licensee Event Report No.89-016 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,

. Powers (M/D 927M)

WNP-2 Plant Manager CMP:lg

Enclosure:

Licensee Event Report No.89-016 cc: Mr. John B. Martin, NRC - Region V Mr. C.J . Bosted, NRC Site (M/D 901A)

INPO Records Center - Atlanta, GA Ms. Dottie Sherman, ANI Mr. D.L. Williams, BPA (M/D 399) 8906200200 890613 PDR ADOCK 05000397 8 PDC

NRC Form 385 V.S. NUCLEAR AEGULATORY COMMISSION (94)3)

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""'("Engineered Safety Feature System Actuation Caused By Inadvertent Fuse Removal and 050003971OF06 Loss of Power Due to Inade uate Fuse Labelin /Inade uate Trainin /Personnel'Error EVENT DATE (5) LER NUMBER 18) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEGVENTrAL RLvrsrON OAY eACILITY NAMES DOCKET NUMBERISI NUMBER NUMBER MONTH YEAR 0 5 0 0 0 0 514 8 9 8 9 01 6 0 0 0 6 1 3 8 9 0 5 0 0 0 OPERATING THIS AEPOATIll IS SUBMITTED PURSUANT T 0 THE REGUIREMENTs oF 10 cFR 4)I (cere>> one or morr ol mr lorlowinpt (11 MODE (0) 20.402(bl 20.405(c) X 50,73(ol(2)(ivl 73,7)(rr)

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NAME TELEPHONE NUMBER AREA CODE J .D. Arbuckle Com liance En ineer COMPLETE ONE LINE FOA EACH COMPONENT FAILURE OESCAIBEO IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFAC TVRER AEPOATABLE TO NPRDS CAUSE SYSTEM COMPONENT MANVFAC TURER EPOA  :~ ',-~'IYj,';.

TABLE:yR,jg<~(~g<~

SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED SUBMISSION IIIyrr, complrtr EXPECTED SVBMISSIOrY OA TEI DATE I15)

YES NO ABSTAAcT ILimir to 1400 rpecel. I e., rpprooimrrrly RINrn rinplr rpece typrwnrtrn Ii nHI (18)

On May 14, 1989 at 0912 hours0.0106 days <br />0.253 hours <br />0.00151 weeks <br />3.47016e-4 months <br />, a Plant Reactor Operator inadvertently tripped the Division 1 and 2 offsite power supply feeders which caused a loss of power to safety related power Buses SM-7 (Division 1) and SM-4 (Division 3). The loss of power to bus SM-7 caused the Division 1 Diesel Generator to start and load bus SM-7. In addition, the loss of Division power caused the loss of power to Reactor Protection System (RPS) Bus 1

A which caused multiple Engineered Safety Feature (ESF) isolations and actuations. At the time of the event the plant was shutdown for the annual refueling and maintenance outage. Preventive maintenance was scheduled for Division 2 of the AC Distribution System. (including safety related Bus SM-8) and event. The Backup Power Supply TR-B (which powers safety related buses SM-7 and SM-8) it was inoperable prior to and during the was de-energized and unavailable.

The cause of the event was the inadervertent removal of potential transformer fuses that sense TR-S (offsite startup) line voltage and provide trip and lockout signals to breakers S-1 and S-2 (through which buses SM-1 and SM-2 are powered). As a result, off-site power was lost or unavailable to all safety related buses, and Diesel Generator 1 (DG-1) started and was relied upon to restore safety related bus SM-7 to power. The following systems were lost due to the loss of power: Residual Heat Removal (RHR)

Shutdown Cooling, Plant Service Water (TSW), Fuel Pool Cooling (FPC), Tower Makeup Water (TMU), Control and Service Air (CAS) and Control Rod Drive (CRD). The loss of power to Bus SM-7 de-engerized RPS Bus A causing an RPS half scram. The loss of RPS Bus A also causes an Outboard Nuclear Steam Supply Shutoff System (NSSSS) isolation of Groups 1 (Main Steam Line Drains only), 2, 5, 6, and 7. NSSSS Group 5 isolates RHR Shutdown Cooling. In addition, the loss of RPS A power causes some NSSSS Group 3 (Primary and Secondary Containment Ventilation and Purge Systems) and Group 4 (Miscellaneous Balance of Plant) isolations and actuations.

NRC Form 358 (9.83)

I NRC Form 3SSA (94)3) U.S, NUCLEAR REOULATORY COMMISSION LICENSEE NT REPORT (LER) TEXT CONTINUA ON APPROVEO OMB NO. 3180-O(04 EXPIRES: 8/31/88 FACILITY NAME (I) OOCKET NUMBER (2)

LER NUMBER (8) PAOE I3)

YEAR @(< SEQUF NTIAL g~R'I REVISION NUMBER NUMSSR Washington Nuclear Plant - Unit 2 0 s 0 0 03 97 8 9 0 1 0 0 0 2 0 6 TEXT ///mors spssois rsr/Irirrd, Irss sddio'ons/HRC Fonrr 3884'4/(IT)

Abstract (continued)

Plant operators restarted RPS-MG-lA and reset the RPS half scram. The operators restored power to Division 1 SM-7 through SM-1 by 0941 hours0.0109 days <br />0.261 hours <br />0.00156 weeks <br />3.580505e-4 months <br />. The Division 3 (High Pressure Core Spray) bus SM-4 power was restored by 1002 hours0.0116 days <br />0.278 hours <br />0.00166 weeks <br />3.81261e-4 months <br />. By 1017 hours0.0118 days <br />0.283 hours <br />0.00168 weeks <br />3.869685e-4 months <br /> RHR Shutdown Cooling, Plant Service Water and Control and Service Air were recovered and, at 1426 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.42593e-4 months <br />, Fuel Pool Cooling was restored. The TMU System was not restarted due to the planned SM-8 outage which de-energized the discharge valve for pump TMU-P-lA.

The root causes of this event are 1) inadequate labeling of Potential Transformer Cabinet doors, 2) infrequent training of Operations personnel on pulling PT fuses, and 3) personnel error in not following procedures. Corrective actions include 1) improving the labeling of the Potential Transformer cabinet doors, 2) providing training on pulling PT fuses, and 3) counselling the individual involved.

There is no safety significance associated with this event. The Division 1 Diesel, Generator started and restored power to safety related bus SM-7 as designed, and RHR Shutdown Cooling was restored within the allowable Technical Specification time limit.

Plant Conditions a) Power Level - 0%

b) Plant Mode - 5 (Refueling)

Event Descri tion On May 14, 1989 at 0912 hour0.0106 days <br />0.253 hours <br />0.00151 weeks <br />3.47016e-4 months <br /> s, a Plant Reactor Operator inadvertently tripped the Division and 1 2 offsite power supply feeders which caused a loss of power to safety related power Buses SM-7 (Division 1) and SM-4 [Division 3: High Pressure Core Spray (HPCS)j. The loss of power to bus SM-7 caused the Division 1 Diesel Generator to start and load bus SM-7. In addition, the loss of Division 1 power caused the loss of power to Reactor Protection System (RPS) Bus A which caused multiple Engineered Safety Feature (ESF) isolations and actuations.

At the time of the event the Plant was in operational condition 5 (Refueling) with the reactor head removed, the reactor cavity flooded up and the fuel pool gates removed. In addition, buses SM-7 and SM-4 were powered from the Startup Power Supply TR-S through buses SM-1 and SM-2 and breakers Sl and S2. The Division'2 .bus SM-3 and associated safety related bus SM-8 were de-energized in maintenance. The Backup Power Supply TR-B (offsite power supply to safety related buses SM-7 and SM-8) was de-energized to prevent backfeed into the breakers scheduled for maintenance, and the Division 2 and 3 diesel generators were out of service for planned maintenance activities.

NRC FORM SddA ~ I/.S. CPOr (SSS-Ssn-SSS OOO)n (9 83)

NRC Form 38dA (94)3) U.S, NUCLEAR REGULATORY COMMISSION LICENSEE E NT REPORT (LER) TEXT CONTINUATION APPROVED OMS NO. 3150&104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (1)

LER NUMBER (8) PAGE (3)

SEOVENTIAL REVISION NVMSSR NUMss/I Washin ton Nuclear Plant - Unit 2 o3 97 89 0 1 6 000 3 OF TEXT /llmors 4/rssois rsrlrrlrsd, oss sddr(r'oos/HAC Form JSSA'4/ (IT)

The event occurred when the reactor operator implemented the last action on a Danger Tag Clearance Order which was described as UE-PT-SM3/1 Fuses (3) Auxiliary Cubicle". On examining the SM-3 Auxiliary Cubicle the reactor operator was confronted with two internal cabinets, one over the other, reading "Bus Potential Transformers" on the upper cabinet and "Line Potential Transformers" on the lower.

The PT fuses are located inside the cabinet doors and opening the doors removes the associated PT fuses from the circuit. The designations UBusR and "Line" Potential Transformers were the only labels in the cubicle. In an attempt to relate the Clearance Order EPN to the auxiliary cubicle labels, the operator consulted with other reactor operators and the Work Control Center Group for determining the appropriate equipment on the clearance order. The operator also reviewed the appropriate Electrical Wiring Diagram (EWD) to determine the correct fuses to be pulled. The operator felt that he had determined from the drawing that the two sets of fuses in the cubicle were in series and were the appropriate fuses to be pulled.

He then removed the UBus Potential Fuses" from the circuit. Although these were the correct fuses, he was expecting to see two fuses but found three and questioned which fuses should be pulled. During this process the operator failed to involve Operations supervision as required by procedure.

As a result of further review of the EWD and subsequent discussions with other individuals, the operator felt he had now identified the correct fuses. The operator then proceeded to the auxiliary cubicle and also pulled the "Line Potential Fuses," causing the loss of power to SM-1 and SM-2. This action was beyond the scope of that provided in the Clearance Order.

The loss of power to Class lE Bus SM-7 caused the Division 1 Diesel Generator to start and load Bus SM-7. The loss of power caused the following systems to be temporarily lost: RHR (Shutdown Cooling), Plant Service Water (TSW), Fuel Pool Cooling (FPC), Control Rod Drive (CRD), Tower Makeup Water (TMU) and Control and Service Air (CAS).

In addition, the loss of RPS A caused an Outboard Nuclear Steam Supply Shutoff System (NSSSS) isolation. The outboard isolations occurred for NSSSS Group 1 (Main Steam Line Drains only), Group 2 (Reactor Water Sample valves), Group 5 [Residual Heat Removal (RHR) and Traversing In-Core Probe (TIP) Systems], Group 6 (RHR Shutdown Cooling), and Group 7 (Reactor Water Cleanup System). In addition, loss of RPS A power causes some NSSSS Group 3 (Primary and Secondary Containment Ventilation and Purge System) and NSSSS Group 4 (Miscellaneous Balance of Plant) isolations and actuations. The Standby Gas Treatment (SGT) System also started on loss of power to RPS Bus A.

At the time of the event, RPS-Bus-B had previously been transferred to alternate power and was not affected by the event. In addition, the HPCS System had previously been isolated for maintenance.

NRC FORM 388A ~ V.S. CPOr )r/88-81O-SS&/OOO)O (84)3)

NRC Form 368A (94)3) U.S. NUCLEAR REGULATORY COMMISSION LICENSEE E (=NT REPORT ILER) TEXT CONTINUATION APPROVED OMB NO, 3150W104 EXPIRES: 8/3)/88 FACILITY NAME (I) OOCKET NUMBER (2)

LER NUMBER (6) PACE (3)

~+'O SEQUENT/Al YEAR NUMBBA 'O NUMBBA Mashin ton Nuclear Plant - Unit TEXT llfmoss space ls seooked. use addio'ooa/l/RC Form 355A'sl (IT) 2 05000397 89 01 0 0 4 OF Immediate Corrective Action Plant operators restarted RPS-MG-lA and reset the RPS half scram. By 0941 hours0.0109 days <br />0.261 hours <br />0.00156 weeks <br />3.580505e-4 months <br /> the "Line Potential Transformer" fuses had been reinserted and TR-S power returned to SM-7 with DG-1 in idle. SM-4 power was restored by 1002 hours0.0116 days <br />0.278 hours <br />0.00166 weeks <br />3.81261e-4 months <br />. By 1017 hours0.0118 days <br />0.283 hours <br />0.00168 weeks <br />3.869685e-4 months <br /> RHR Shutdown Cooling had been returned to service as had the rest of the systems listed above, with the exception of Fuel Pool Cooling which was returned to service at 1426 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.42593e-4 months <br />. The TMU System was not restarted due to the planned SM-8 outage which de-energized the discharge valve for pump TMU-P-1A.

Further Evaluation and Corrective Action A. Further Evaluation

1. The starting of the Division 1 Diesel Generator and the Nuclear Steam Supply Shutoff System isolations and actuations are Engineered Safety Features and; therefore, this event is reportable per 10CFR50.73(a)(2)(iv).
2. Had the backup power supply bus TR-B been available, relay logic would have allowed Breaker B-7 to energize SM-7 and return power to the RPS-MG-1A before the NSSSS isolation actuations occurred. Flywheel energy in the RPS-MG-lA would have carried transient. In addition, Backup Power Supply TR-B would have re-powered it through the loss of power SM-8 if it had been operable. The Division 2 and Division 3 Diesel Generators would have also started if they had been operable.

no other structures, components, or systems inoperable prior to the event There were that contributed to the event.

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

~ Inadequate labeling of the RBusR and "Line" Potential Transformer cabinet doors. Had the appropriate Equipment Part Number (EPN) label been affixed to the cabinet doors, the operator would have been able to determine and pull the correct PT fuses.

~ Infrequent PT fuse training of Operations personnel. Pulling PT fuses is not frequently performed by operators and, as a result, training as to designation, significance, and orientation is normally not covered in refresher training.

~ Personnel error. The reactor operator failed to involve Operations supervision as required by procedure when potential problem existed.

it was determined that a In addition, the operator expanded the scope of the Clearance Order without authorization to do so. The operator had pulled five fuses instead of three as designated by the Clearance Order. Plant procedures were not a cause of this event.

NAC FOAM 38BA ~ I/.S. CFOG (988-820-889 000)0 (9 831

NRC Form 36&A (9413) U.S. NUCLEAR AEGULATOAYCOMMISSION LICENSEE E NT REPORT (LER) TEXT CONTINUATION APPAOVEO OMS NO, 3150WI0&

EXPIRES: 8/31/88 FACILITY NAME (11 OOCKET NUMBER (31 LER NUMBER (6I PAGE (3)

YEAR h SEOVENTIAL REVISION NVMS&R gpss N VMS&A Washin ton Nuclear Plant - Unit TEXT lifmore epeee ie rerfofred, Fee eddf(lone/ NRC Form 36EA'ef (171 2 0 5 0 0 0 3 9 7 0 1 O 5'"O B. Further Corrective Actions

1. SM-3 Auxiliary Cabinet and similar cabinet labeling will be improved to eliminate the possibility for inaccurate identification of Potential Transformer fuses.
2. The Operations department will provide training to operators as to designation, significance, and orientation of the "Line" and "BusU Potential Transformers in the SM-3 Auxiliary Cabinet and similar cabinets.
3. The reactor operator involved was counselled on 1) the importance of informing Operations supervision when a problem is discovered during the clearance order process, and 2) the importance of following procedures.
4. This LER will be required reading for all Plant Operators (Licensed and Non-Licensed).
5. A double verification program for clearance order implementation is currently required prior to removal and reinstallation of fuses in the control room. The Operations department will evaluate expanding this program beyond the control room.

Safety Si nificance There is no safety significance associated with this event. The Division 1 Diesel Generator started and repowered Bus SM-7 as designed. At the time of the event the reactor water level was greater than 22 feet above the reactor vessel flange which provides a large heat sink for core cooling and adequate time to restore RHR Shutdown Cooling. Residual Heat Removal Shutdown Cooling was restored within the allowable Technical Specification time limit. Additionally, no Plant condition requiring the ESF isolations and actuations existed and all ESF isolations and actuations occurred as designed. Accordingly, this event posed not threat to the health and safety of the public or Plant personnel.

Similar Events88-021 EIIS Information Text Reference EIIS Reference System Component Bus SM-7 (Div. 1) and SM-4 (Div. 3) EB SWGR Diesel Generator DG-1 EK DG Reactor Protection System (RPS) JC RPS-Bus-A JC BU Bus SM-8 (Div. 2) EB SWGR NRC FORM 3&&A AVAST CPOs 1888-530 58$ r00010 (8831

NRC Form 388A U.S. NUCLEAR REGULATORY COMMISSION (9 83)

LICENSEE -NT REPORT (LER) TEXT CONTINUA N APPROVEO OMB NO, 3(50&)04 EXPIRES: Bl3(l88 FACILITY NAME (I ) OOCKET NUMBER (31 LER NUMBER (6) PAGE (3)

YEAR SEGVENTIAL os% REVISION NUMB'EA V R NVMBEA Washin ton Nuclear Plant - Unit TEXT fllmoro opoooi4 roqolrod, 044 oddidonoi HRC Form 388A 3) (17) 2 o s << o 3 9 7 8 9 0 1 6 00 06 OF 0 6 EIIS Information Text Reference EIIS Reference System Component TR-B and TR-S EA XFMR Breakers S-1 and S-2 EA BKR Bus SM-1 and SM-2 EB BU Residual Heat Removal (RHR) System (Shutdown Cooling) BD Plant Service Water (TSW) System KG Fuel Pool Cooling (FPC) System DA Tower Makeup Water (TMU) System KI Control and Service Air (CAS) System LD Control Rod Drive (CRD) System AA Nuclear Steam Supply Shutoff System (NSSSS) BD Primary and Secondary Containment and Purge Systems VH RPS-MG-lA JC MG TMU-P-lA KI P Diesel Generator DG-2 and DG-3 EK DG Fuses E-PT-SM3/1 EA FU Main Steam Line Drain Valves SN LOV Reactor Water Sample Valves AD ISV Traversing In-Core Probe (TIP) System IG Reactor Water Cleanup (RWCU) System CE Standby Gas Treatment (SGT) System. BH RPS-Bus-B JC BU High Pressure Core Spray (HPCS) System BG NRC FORM 3@BA

~ V.ST QPOa (986-530-589(00070 (94)3)

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