ML17286B273

LER 91-022-00:on 910905,inboard RHR Sys Shutdown Cooling Supply Valve Automatically Isolated on High Suction Line Flow Signal.Caused by Instrumentation Drift.Sys Realigned & Loop B Placed Back in svc.W/911007 Ltr

ML17286B273
Person / Time
Site:	Columbia
Issue date:	10/07/1991
From:	Arbuckle J, John Baker WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-91-022, LER-91-22, NUDOCS 9110160168
Download: ML17286B273 (7)

Text

ACCELERATED D TRIBUTION DEMONS ~ATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9110160168 DOC.DATE: 91/10/07 NOTARIZED: NO DOCKET FACIL:50-397 WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION ARBUCKLE,J.D. Washington Public Power Supply System BAKER,J.W. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 91-022-00:on 910905,inboard RHR sys shutdown cooling supply valve automatically isolated on high suctio'n line flow signal. Caused by instrumentation drift.Sys realigned & D loop B placed back in svc.W/911007 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ( ENCL SIZE:

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

NOTES:

RECIPIENT COPIES RECIPIENT COPIES D ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD5 LA 1 1 PD5 PD 1 1 ENG,P.L. 1 1 D INTERNAL: ACNW ACRS 2 2 AEOD/DOA AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB 7E NRR/DLPQ/LHFB10 1 . 1 NRR/DLPQ/LPEB10 NRR/DOEA/OEAB 1 1 NRR/DREP/PRPB11 NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 RR/J3SZ/SPLBSDl 1 1 NRR/DST/SRXB 8E REG FILLE 02 1 1 RES/DSIR/EIB GN5 FILE 01 1 1 EXTERNAL: EG&G BRYCE,J.H L ST LOBBY WARD 1 1 NRC PDR NSIC MURPHYiG A

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NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE iVASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM P 1-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 33 ENCL 33

ii WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968 ~ 3000 George tVashington Way ~ Richland, i1'ashington 99352 Docket No. 50-397 October 7, 1991 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

NUCLEAR PLANT NO. 2 LICENSEE EVENT REPORT NO.91-022

Dear Sir:

Transmitted herewith is Licensee Event Report No.91-022 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, J. W. Baker WNP-2 Plant Hanager

Enclosure:

Licensee Event Report No.91-022 cc: Hr. John B. Hartin, NRC Region V Hr. C. Sorensen, NRC Resident Inspector (H/D 901A)

INPO Records Center Atlanta, GA Hs. Dottie Sherman, ANI Hr. D. L. Williams, BPA (H/D 399)

NRC Resident Inspector walk over copy 1007 gp~

91 101601+8 91 PDR ADOCK 05000397 PDR

NRC FORM 366 (84)9)

LICENSEE EVENT REPORT ILER)

U.S. NUCLEAR REGULATORY COMMISSION t APPROVED 0MB NO. 3)504)104 EXPIRES: 4/30/92 ESTIMATED BURDEN PEA RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HAS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F530). V.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (31504)104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON. DC 20503.

DOCKET NUMBER (2) PAGE 3 FACILITY NAME (I)

IhIashin ton Nuclear Plant Unit 2 o 5 o o o39 710F05 Residual Heat Remova R ys em u own oo ing so a 1on n High Suction Line Flow Due To Differential Pressure Instrumentation Drift EVENT DATE (6) LER NUMBER (8) REPORT DATE (7) OTHER FACILITIES INVOLVED (SI

,xc6 SEQUENTIAL REVS~A MONTH FACILITYNAMES DOCKS'T NUMBER(S)

MONTH DAY YEAR '.XCr/r~ NVMSER p@ DAY YEAR 0 5 0 0 0 0 905 91 2 2 00 10 7 9 1 0 5 0 0 0 THIS REPOAT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR (II /Check onr or morr of the followinPI (11)

OPERATING MODE (9) 4 20A02(b) 20.405(c) 60,73(r)(2 I(ivl 73.71(III POWER 20.406( ~ ) (1)(i) 50M (cl (I) 60,73( ~ l(2)(v) 73.7) (c)

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LICENSEE CONTACT FOR THIS LER (12I 60 73(r l(2)(rlil)(BI 50.73(r l(2) (xl TELEPHONE NUMBER NAME AREA CODE J. D. Arbuckle, Compliance Engineer 50 937 7- 145 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

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SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED SUBMISSION DATE I15)

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Abstract On September 5, 1991 at 1122 hours0.013 days <br />0.312 hours <br />0.00186 weeks <br />4.26921e-4 months <br /> while the Plant was shut down for an outage, the inboard Residual Heat Removal (RHR) System Shutdown Cooling Supply Valve (RHR-V-9) automatically isolated on a high suction line flow signal while Plant Operators were realigning RHR System shutdown cooling from Loop B to Loop A. Closure of RHR-V-9, a Nuclear Steam Supply Shutoff System (NSSSS) containment isolation valve, isolated and tripped RHR Shutdown Cooling Loops A and B which were in service at the time.

The cause of this event was instrumentation drift pertaining to Differential Pressure Indicating Switch RHR-DPIS-12B (RHR Shutdown Cooling Supply Leak Detection High Flow Indicator). As an immediate corrective action Plant Control Room Operators took appropriate steps to re-align the system and, at 1202 hours0.0139 days <br />0.334 hours <br />0.00199 weeks <br />4.57361e-4 months <br />, one loop of RHR Shutdown Cooling (Loop B) was placed back into service, well within the two hours allowed for shutdown cooling to be out of service.

NRC Form 386(64)9)

NAC FORM 366A (689)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION V.S. NUCLEAR REGULATORY COMMISSION t APPAOVED OMB NO. 31500104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST/ 50t) HAS. FORWARD COMMENTS AEGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P.530). V.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31504)104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (8) PAGE (3)

SEOUENTIAL 9+( REVLEION NUMSER NUMSER 1

Unit 0 5 0 0 0 3 9 7 9 1 022 0 0 0 2 oF 0 TEXT //f mors 4/rooo tr rortuiiod, rrro Edde)rno/HRC %%dnrr 3654's/ (Ill Further corrective actions consisted of 1) recalibrating pressure switches RHR-DPIS-12A and RHR-DPIS-12B, and 2). revising the RHR- DPIS-12B calibration procedure to require that a channel functional test be performed monthly.

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

Plant Conditions Power Level OX Plant Mode 4 (Cold Shutdown)

Event Descri tion On September 5, 1991 at 1122 hours0.013 days <br />0.312 hours <br />0.00186 weeks <br />4.26921e-4 months <br /> while the Plant was shut down for an outage, the inboard Residual Heat Removal (RHR) System Shutdown Cooling Supply Valve (RHR-V-9) automatically isolated on a high suction line flow signal during efforts associated with realigning the RHR System. Closure of RHR-V-9, a Nuclear Steam Supply Shutoff System (NSSSS) containment isolation valve, isolated and tripped RHR Shutdown Cooling Loops A and B which were in service at the time.

At the time of the event, Plant Operators were in the process of realigning RHR System shutdown cooling from Loop B to Loop A. At 1120 hours0.013 days <br />0.311 hours <br />0.00185 weeks <br />4.2616e-4 months <br />, with Loop B (RHR-P-2B) in service, Plant Operators started Loop A (RHR-P-2A) in preparation for changing the in- service RHR Shutdown Cooling System. However, within two minutes while Plant Operators were in the process of securing RHR-P-2B, RHR-V-9 automatically isolated with both loops of RHR shutdown cooling being in service. By design, the closure of RHR-V-9 also caused RHR-P-2A and RHR-P-2B to trip.

Plant design is such that RHR-V-9 will isolate on Reactor Pressure Vessel (RPV) low level (Level 3), Drywell pressure greater than 1.6B psig, or RHR Shutdown Cooling line excess flow. After verifying that there were no level or pressure transients during the event, Plant Operators concluded that the isolation of RHR-V-9 was most likely due to exceeding the high suction flow setpoint. The reason for the isolation was instrumentation drift pertaining to Differential Pressure Indicating Switch RHR-DPIS-12B (RHR Shutdown Cooling Supply Leak Detection High Flow Indicator).

NRC Form 368A (64)9)

NRC FORM 366A (689)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION t APPROVED OMB NO. 31504)104 E XPI A ES: 4/30/92 ESTIMATED BUADEN PER RESPONSE TO COMPLY WTH THIS INFOAMATION COLLECTION REOUEST: 50A) HAS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR @> SEQUENTIAL g~X IIEvrsroN NUM EA NUMBER Washington Nuclear Plant Unit 2 0 5 0 0 0 3 7 9 1 2 2 0 0 OF 0 5 TEXT Ilfrrroro spsso r'4 rlrlrrnd, oso oddldorrs/HRC Forrrr 36IS4'4/ (12)

Immediate Corrective Action Plant Control Room Operators took appropriate action to re-align the system and, at 1202 hours0.0139 days <br />0.334 hours <br />0.00199 weeks <br />4.57361e-4 months <br />, one loop of RHR Shutdown Cooling (Loop 8) was placed back into service, well within the two hours allowed for shutdown cooling to be out of service.

Further Evaluation and Corrective Action A. Further Evaluation

l. This event is reportable under 10CFR50.73(a)(2)(iv) as an event or condition that resulted in manual or automatic actuation of an Engineered Safety Feature.

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

3. Two redundant differential pressure switches (RHR-OPIS-12A and RHR-OPIS-12B), one for each trip logic, monitor the RHR shutdown cooling suction line. Circuit logic operation is such that the output trip signal of each sensor initiates a logic trip and closure of either the inboard

[RHR-V-9 (RHR-DPIS-12B)] or outboard [RHR-V-8 (RHR-DPIS-12A] isolation valve in the event of an excess flow condition in the RHR suction line.

4. The root cause of this event was instrument drift. On September 6, 1991 Plant Instrument and Control ( II(C) Technicians checked the as-found trip settings of RHR- DPIS-12A and RHR-OPIS-12B and determined that the high excess flow trip setting For RHR-OPIS-12A had drifted to 147.5 inches (H20) and the setting for RHR-DPIS-12B had drifted to 141.0 inches (H20).

The normal setpoint for these instruments is 174 inches (H20). The inaccuracy of these mechanical instruments is 1.5 percent. As a result, with both loops of RHR in service and a combined flow of approximately 15,000 gpm, the drift value was close enough to the nominal process flow variances for RHR-DPIS-12B to sense a high flow condition in the RHR suction line and, by design, RHR-V-9 automatically isolated.

A contributing factor was that the calibration procedure for RHR-DPIS-12A and RHR-DPIS-12B is only required by the Plant Technical Specifications to be performed during Operational Nodes 1, 2 and 3. Although the instrument drift (which is common to Barton mechanical instruments) has been tracked by the Plant Technical Department since 1987, the calibration surveillance NAC Form 366A (64)9)

NRC FORM 366A (64)9)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION U.S. NUCLEAR REGULATORY COMMISSION t APPROVED OMB NO. 31500104 EXPIRES; 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENTBRANCH (P.530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500(04). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER 12) LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL jYgj REVISION NUMSER NUMBER I/Iashington Nuclear Plant Unit 2 o s o o o 3 7 9 O oFO 5 TEXT ///mno EPooo /I Ioqu/Nd, uoo oddniono/HRC %%dmI 36649/ IIT) had not been performed since April, 1991 due to RHR system configuration during the extended outage (a prerequisite of the procedure is to ensure that the RHR system is not performing any core cooling function).

However, during non-outage periods, both instruments are calibrated monthly in accordance with Technical Specification surveillance requirements.

B. Further Corrective Action

l. On September 6, 1991 pressure switches RHR-OPIS-12A and RHR-OPIS-12B were recalibrated by Plant I&C Technicians.

2. Plant Procedure (PPM) 7.4.3,2.1.63, "RHR Shutdown Cooling Mode High Flow Isolation CFT/CC,R will be revised to require that a channel functional test be performed monthly during all Operational Modes.

Safet Si nificance There is no safety significance with this event. Plant Control Room Operators took appropriate and timely action to re-align the RHR System. In addition, shutdown cooling was isolated for only 40 minutes, well within the time-frame (two hours) allowed by the Technical Specifications. Furthermore, during the event period there was no increase in reactor primary cooling loop temperature, which was 126 degrees at the start of the event. Accordingly, this event posed no threat to the health and safety of either the public or Plant personnel.

Similar Events There have been several LERs pertaining to the loss of RHR shutdown cooling; however, none with a similar root cause.

NRC FomI 366A (6419)

NRS FORM 366A (649)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION U.S, NUCLEAR REGULATORY COMMISSION t APPROVED OMB NO. 3150d)04 EXPIRES: E/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150dl0E). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (21 LER NUMBER (6) PAGE (3)

YEAR rR'Ye SEQUENTIAL 3mr'EVISION NUMBER v.re NUMBER Washington Nuclear Plant Unit 2 p s p p p 3 9 7 9 1 022 0 0 0 5 oFO 5 TEXT /llmere e/Mce ls eEII//red. rrrre eddi5orre/P/RC Form 36643/ (17)

EI IS Information EIIS Reference

~sstem ~Com onent Residual Heat Removal (RHR) System 80 RHR-V-9 80 ISV Nuclear Steam Supply Shutoff System (NSSSS) BD RHR-P-2A 80 RHR-P-28 80 r

RHR-DPIS-12A BO PDIS RHR-DP IS-128 80 PDIS Reactor Pressure Vessel NH RPV NRC Forms 366A (669)