ML17285A584

LER 89-021-00:on 890527,electrical Protection Assembly Breaker Tripped Causing Loss of Power to Reactor Protection Sys Bus B.Cause Unknown.Upgraded Breaker Equipment Will Be installed.W/890623 Ltr

ML17285A584
Person / Time
Site:	Columbia
Issue date:	06/23/1989
From:	Arbuckle J, Powers C WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-021, LER-89-21, NUDOCS 8906300101
Download: ML17285A584 (6)

Text

DISTKBUTION DEMON STATION SYFI'ZM p ~ ~

ck REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8906300101 DOC.DATE: 89/06/23 NOTARIZED: NO DOCKET N FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFIL1ATION ARBUCKLE,J.D. Washington Public Power Supply System POWERS,C.C. Washington Public Power Supply System RECIP.NAME . ~

RECIPIENT AFFILIATION R I

89-021-00:on 890527,ESF isolations actuations due to I

SUBJECT:

LER &

EPA breaker trip cause unknown.

RPS ltr.

DISTRIBUTION CODE TITLE: 50.,73/50.9 IE22T COPIES RECEIVED:LTR J ENCL ( SIZE:

Licensee.

W/8 Event Report (LER), Incident Rpt, etc.

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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 23, 1989 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

NUCLEAR PLANT NO. 2 LICENSEE EVENT REPORT NO.89-021

Dear Sir:

Transmitted herewith is Licensee Event Report No.89-021 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. Powers (M/D 927M)

WNP-2 Plant Manager CMP:lg

Enclosure:

Licensee Event Report No.89-021 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) jo Po4 8906300iOi 890623 O~i000397 +gZS PDR ADOCK S PDC

NRC Form 368 U.S. NUCLEAR REGULATORY COMMISSION (94)3)

APPROVED OMB NO. 31604104 EXPIRES: 8/31/88 LICENSEE EVENT REPORT {LER)

FACILITY NAME lll DOCKET NUMBER (2I PAGE 3)

Washington Nuclear Plant - Unit 2 0 5 0 0 03 97 s oFO 4 TITLE<neere a e y ea ure so a lons an c ua sons ue o a eac or ro ec son System (RPS) Electrical Protection Assembly (EPA) Breaker Trip - Cause Unknown EVENT DATE (5) LER NUMBER (6) REPORT DATE (7I OTHER FACILITIES INVOLVED (8)

SEOVENTIAL MONTH OAY YEAR YEAR @PI NUMBER 4) eNVMSER Mo~' OAY YEAR FACI LIT Y NAMES DOCKET NUMBER(S) 0 5 0 0 0 0 5 2 7 8 9 8 9 021 00 06 2 3 8 9 0 5 0 0 0 OPERATINO THIS REPORT IS SUBMITTED PURSUANT T 0 THE REOUIREMENTS OF 10 CFR g: /Cbecir one or more ol the lollowinp/ (11 MODE ID) 73.71(b) 20.402(b) 20.405( ~ I 50.73(e)(2) (ivl POWER 20.406( ~ Ill)Ill 50.36(cl(1) 50,73(el(2) (v) 73.71(c)

LEYEL 0 0 0 20.405( ~ II)IIII 50.38(cl(2) 60.73 (e l(2)(vS I OTHER ISpecify in Abttrect "co below enr/ In Text. NRC Form 4 20.405(e)II)(DI) 60.731 ~ l(2) III 60.73(el(2)(rill) (Al 3EEA/

20.406(el(1)(lvl 50.73(el(2)(D) 50.73( ~ l(2)(v)8) (8) 20.405(e I (1 I (rl 50.73(el(2)(ill) 60.73( ~ I (2)(x)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER AREA CODE J .D. Arbuckle, Compliance Engineer 50 937 7- 211 5 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANVFAC REPORTABLE g$ ~~~'gP CAUSE SYSTEM COMPONENT MANUFAC. EPORTASLE

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'%O TVRER TVRER NPRDS

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X J C KR GO 0 Y y&($%( N'UPPLEMENTAL REPORT EXPECTED (14) MONTH OAY YEAR EXPECTED SUBMISSION DATE I'IS)

YES III yn, complete EXPECTED SIIBAflSSIDN DA TE/ X NO AssTAAcT ILimlr to /400 /peen. /.e.. epproximerely ///teen t/npie /pete rypewr/Iten linn/ (18)

On May 27, 1989 at 1038 hours0.012 days <br />0.288 hours <br />0.00172 weeks <br />3.94959e-4 months <br /> an El ectrical Protection Assembly (EPA) br eaker (RPS-EPA-3D) 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 Engineered Safety Feature (ESF) actuations. At the time of the event the Plant was in a shutdown condition 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 HZH signal (a non-NSSSS ESF trip 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 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />.

The cause of this event is not known. Investigation at the time of the event identified a potential cause (work in progress on Differential Pressure Indicating Switch MS-DPIS-BB), but the cause can not be conclusively determined. Further corr ective action consists of installing an upgrade to the EPA breaker. equipment. This action will enhance the troubleshooting capabilities for the EPA breakers.

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.

NRC Form 368 (9 83)

NRC Form 38SA U.S. NUCLEAR REOULATORY COMMISSION (943)

LICENSEE EVE REPORT (LER) TEXT CONTINUATIO APPROVED OMS NO. 319)-O(04 EXPIRES: 9/31/(6 FACILITYNAME (1) DOCKET NUMBER (2) LER NUMBER (8) PACE (3)

SSOUSNTIAL 95%'SVISION YEAA gg NVMSSA +>NNVMSSA Washington Nuclear Plant TEXT ///more space /8 t//r/rar/ Irsa ~ - Unit HRC Form SSSA'8) (17) 2 o 5 o o o 397 89 0 1 0 0 OF 0 4 Plant Conditions a) Power Level - OX b) Plant Mode - 5 (Refueling)

Event Descri tion On May 27, 1989 at 1038 hours0.012 days <br />0.288 hours <br />0.00172 weeks <br />3.94959e-4 months <br /> an Electrical Protection Assembly (EPA) breaker (RPS-EPA-3D) tripped causing a loss of power to Reactor Protection System (RPS) Bus B. The loss of power on RPS Bus 8 caused a half-scram in RPS Division 8 and multiple Engineered Safety Feature (ESF) actuations. 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 8 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 Water Sample Valves), Group 5 [Residual Heat Removal (RHR) and Traversing In-Core Probe (TIP) Systems], Group 6 (RHR Shutdown Cooling),

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

In addition, the loss of RPS Bus 8 power causes an NSSSS Group 3 (Primary and Secondary Containment Ventilation and Purge System) and partial Group 4

[Miscellaneous Balance of Plant (6-Valves)] iso'lation. These isolations occurred because RPS Bus 8 is the power supply for Reactor Building Exhaust Plenum Radiation Monitors (Channels 8 and D). Loss of RPS Bus 8 power de-energizes these monitors, causing a NZU 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.

Plant Operators responded by investigating the situation and, finding no immediate cause, shut EPA breaker RPS-EPA-3D and returned RPS Bus 8 to the normal power supply.

Immediate Corrective Action As previously stated, Plant Operators responded by investigating the situation and, finding no immediate cause, returned RPS Bus 8 to the normal power supply. All systems were restored to pre-event lineup status within 22 minutes.

Further Evaluation and Corrective Action A. Further Evaluation

1. This event is 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)."

2. There were no structures, components or systems that were inoperable at the start of the event that contributed to the event.

NAC FOAM 388A i0 ~ 8 ~ CPOr (888 520 588r00070 (94)3)

NRC Form 366A US. NUCLEAR REOULATORY COMMISSION (94)3)

LICENSEE EVE REPORT (LER) TEXT CONTINUATIO APPROVED OMB NO. 3160-0104 EXPIRES: 6/31/BS FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (6) PACE (31 SSOUSNTIAL 'om)) IIEVISION NUMSSR if'UMSSA h'n ton Nuclear Plant - Unit 2 o3 97 89 0 2 1 0 0 0 3 OF 0 4 TEXT (Ifmore epece fe revfrvfved, ore eddvtfovrel HRC %%dvrrr 35543) (IT)

3. The cause of this event is not known. A potential cause was identified following the trip but the cause cannot be conclusively determined.

Immediately following the breaker trip, a review of all work in progress was conducted to identify any activities having the potential of causing a trip. Work on Differential Pressure Indicating Switch MS-DPIS-8B (Main Steam Line A High Flow) was in progress at the time of the trip which consisted of splice inspections and lug replacement. [MS-DPIS-8B is a flow switch in the logic circuit for relay RPS-RLY-K3B (Main Steam Isolation Valve Closed) which is utilized in the Nuclear Steam Supply Shutoff System actuation logic]. If during the work on MS-DPIS-8B a short occurred, it would be sensed on RPS Bus B and the EPA breaker might trip on undervoltage before a protective fuse on RPS-RLY-K3B opened. No method is available to confirm the probable cause; however, there have been no spurious trips during the last year of operation and none since this event.

4. Visual inspections of the breaker have also provided no indication of probable cause of the trip.

5. The EPA Breaker (RPS-EPA-3D) is manufactured by General Electric Company (Part No. TFJ226175W6).

B. Further Corrective Action

1. The EPA Breaker Assembly (RPS-EPA-3D) was replaced with a spare component.

2. An upgrade . to the EPA breaker equipment has recently been developed that provides enhanced operating characteristics, including modifications to reduce susceptability to spurious trips, a "first-in, seal-in" indicator and improved output voltage control. This upgrade will be procured and installed, subject to availability of the modified components, by the end of the next maintenance and refueling outage.

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 actuations occurred as designed.

In addition, at the 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 pr e-event lineup status within 22 minutes.

Accordingly, this event posed no threat to the health and safety of either the ~

public or Plant personnel.

Similar Events LERs87-019 and 87-025 NRC FORM SSSA ~ II,S. CPOr (966 520-569v00090 (94)3)

NRC Form 388A U.S. NUCLEAR REOULATORY COMMISSION (943)

LICENSEE EV REPORT (LER) TEXT CONTINUATIO APPROVED OMB NO. 3150&184 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) PACE (3)

LER NUMBER (8)

YEAR RZZ( SEQUSNrrAL 'P/r. ASyeloN c?r?: NUM88A '.Z~r NUMBSA Washington Nuclear Plant TEXT /I/nroro 8/roco /t rorlrr/rod, um odd/I/ono/NRC Form 388r('8/

- Unit (17) 2 o s << o 3 9 7 02 1 00 0 4>> 0 4 EI IS Information Text Reference EIIS Reference System Component EPA Breaker (RPS-EPA-3D) JC BKR Reactor Protection System (RPS) JC RPS-Bus-B JC BU Nuclear Steam Supply Shutoff System (NSSSS) BD Main Steam Line Drain Valves SN LOV Reactor Water Sample Valves AD ISV Residual Heat Removal (RHR) System BD Traversing In-Core Probe (TIP) System IG Reactor Water Cleanup (RWCU) System CE Reactor Bulding Exhaust Plenum Radiation Monitor IL MON Standby Gas Treatment (SGT) System BH Control Room Emergency Filtration System VH Reactor Building HVAC VA MS-DP I S-8B . SB PDI S RPS-RLY-K3B JC RLY NAC FOAM 388A rU.S. CPOr 1988 f)0-589r00070 (983)