ML17349A456

LER 92-008-00:on 921005,lift of Pressurizer PORV PCV-4-455C & Isolation of MOV-4-751 Residual Heat Removal Suction Overpressure Mitigating Sys Surveillance Occurred Due to Personnel Error.Specialist counseled.W/921028 Ltr

ML17349A456
Person / Time
Site:	Turkey Point
Issue date:	11/04/1992
From:	Knorr J, Plunkett T FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-92-296, LER-92-008-01, LER-92-8-1, NUDOCS 9211030293
Download: ML17349A456 (10)

Text

<ccrc.za.~rZD D>Srmamrow DarvrOxSra~rrox svsrziw REGULATC ZNFORMATZON DZSTRZBUTZONSTEM (RZDS)

ACCESSION NBR:9211030293 DOC.DATE: 92/ll/04 NOTARIZED: NO DOCKET FACIL:50-251 Turkey Point Plant, Unit 4, Florida Power and Light C 05000251 AUTH. NAME AUTHOR AFFILIATION

'KNORRFJ.E. Florida Power & Light Co.

PLUNKETT,T.F. Florida Power & Light- Co.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 92-008-00:on 921005,lift of pressurizer PORV'CV-4-455C

& isolation of MOV-4-751 residual heat removal suction overpressure mitigating sys surveillance occurred due to personnel error. Specialist counselled.W/921028 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL 1 SIZE:

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

NOTES:NRR RAGHAVANFL 05000251 A

RECIPIENT COPIES'TTR RECIPIENT COPIES ID CODE/NAME ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 1 PD2-2 PD 1 1 AULUCK,R 1 1 INTERNAL: ACNW 2 2 AEOD/DOA- 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFB10 1 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB 1 1 NRR/DREP/PRPB11 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1 NRWR~DSTlSPLB8Dl 1 1 NRR/DST/SRXB 8E 1 1 ~EG 02 1 1 RES/DSIR/EIB 1 1 RGN2 FILE 01 1 1 EXTERNAL: EG&G BRYCEFJ..H 2 2 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHY,G.A 1 1 NSIC POOREFW 1 1 NUDOCS FULL TXT 1 1 NOTES: 1 1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK.

ROOM Pl-37 (EXT. 504-2065) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 30 ENCL 30

I~ Cl

.u. Box 029100, Miami, F!33102-9100

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ONT 28 1992 FPL L-92-296

'10 CFR 50.73 U.. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:

Re: Turkey Point Unit 4 Docket No. 50-251 Reportable Event: 92-008-00 Lift of Pressurizer PORV PCV-4-455C and Isolation of MOV-Suction During Overpressure 4-751 Residual Heat Removal Miti atin S stem Surveillance.

The attached Licensee Event Report 251-92-008-00 is being provided in accordance with 10 CFR 50.73 (a) (2) (v) .

If there are any questions please contact us.

Very truly yours, T. F. Plunkett Vice President Turkey Point Nuclear TFP/JEK/3k enclosure cc: Stewart D. Ebneter, Regional Administrator, Region II, USNRC Ross C. Butcher, Senior, Resident Inspector, USNRC, Turkey Point Plant 03Q(";6-.

92ii030293 92ii04 PDR ADQCK 0500025i i 8 PDR an FPL Group company pJ

Cl LICENSEE EVENT REPORT (LER)

FACILITY NAFZ (1) DOCKET NUMBER (2) PAGE (3)

TURKEY POINT UNIT 4 05000251 oF -3 TZTLE (4) Lift of Pressurizer PORV PCV-4-455C and Isolation of MOV-4-751 Residual Heat Removal Suction During Overpressure Mitigating System Surveillance.

EVENT DATE (5) LER NUMBER(6) RPT DATE (7) OTHER FACILITIES INV (8)

MON DAY YR YR SEQ 6 MON DAY YR FACILITY NAMES DOCKET ( (S) 10 05 92 92 008 00, 11 04. 92 ~ TURKEY POINT UNIT 3 05000250 OPERATING MODE (P) THIS REPORT IS SUBMITTED PURSUANT TO THE RE UIREMENTS OF 10 CFR 10 CFR 50.73 a 2 PONER LEVEL (10) 06 LICENSEE CONTACT FOR THIS LER (12)

James E. Knorr, Licensing Engineer TELEPKONE NUMBER 305-246-6757

'COMPLETE ONE LINE FOR EACM COMPONENT FAILURE DESCRZBED ZN TRIS REPORT (13)

CAUSE SYSTEM COMPONENT FAlCUFACTURE R NPRDSI CAUSE SYSTEM COMPONENT WNUFACTURER NPRDS2

, SUPPLEMENTAL REPORT EXPECTED (14) NO X YES 0 EXPECTED SUBMISSION MONTH DAY YEAR DATE (15)

(if Yes, co~lete EXPECTED SUBMISSION DATE)

ABSTRACT (16)

On October 5, 1992, at 1158 EDT, performance of a Technical Specification required survei:llance analog channel operational test of the Unit 4 overpressure mitigating system resulted in an opening of power operated relief valve PCV-4-455C and the closing of residual heat 'F removal (RHR) pump suction valve MOV-4-751. Unit 4 was in Mode 5 at 178 and 354 psig. The pressurizer power operated relief valve was closed immediately after the event. During the event reactor coolant pressure decreased from 354 psig to 345 psig. Prior to the event, MOV-4-751 was open allowing residual heat removal flow of 3600 gallons per minute. During the event MOV-4-751 closed and was immediately reopened. Residual heat removal pump suction flow stopped with valve closure and returned to normal after the valve was reopened. The 'A'HR pump was turned off upon loss of suction and the turned on,to .supply the flow after the valve was reopened.

'B'HR pump was The reactor coolant system temperature rose 1 'F during: the event. The cause of the event was personnel error. The controlling procedure was revised to reduce the probability of recurrence of this type of event.

t LICENSEE ET REPORT (LER) TEXT NTINURTION FACILITY NAME DOCKET NUMBER LER NUMBER PAGE NO.

TURKEY POINT UNIT 4 05000251 92-008-00 02'P 03 Z. EVENT DESCRIPTION On October 5, 1992, at 1158 EDT, performance of a Technical Specification required surveillance analog channel operational test of the Unit 4 overpressure mitigating system (EIIS-AB) resulted in an opening of power operated relief valve PCV-4-455C (EIIS-AB) (IEEE-RV) and the closing of residual heat removal pump suction valve MOV-4-751 (EIIS-BP) (IEEE-V) . Unit 4 was in Mode 5 at 178 F and 354 psig. A pre-evolution briefing concerning the controlling procedure 4-OSP-041.4, Overpressure Mitigating System Nitrogen Backup Leak and Functional Test, was conducted between Instrumentation and Control (I&C) personnel and Operations personnel. The primary loop was prepared for the test to allow for the anticipated opening of the pressurizer power operated relief valve (PORV) PCV-4-456 without depressurization of the reactor coolant system and to provide a closure signal to the residual heat removal series suction valve MOV-4-750. The primary loop preparation included the closing of the block valve for PCV-4-456 (MOV-4-535) and the opening of the power breaker for MOV-4-750 to prevent closure of the valve and isolation of residual heat removal pump suction. The analog operational test of the pressurizer power operated relief valve is accomplished by the introduction of an artificial high pressure signal to the loop being tested and verification of the appropriate operation of the loop.

After the pre-evolution briefing, the I&C specialist erroneously proceeded to apply an artificial high pressure signal to the backup loop (instead of the primary loop) of the overpressure mitigating system. The backup loop, which includes valves PCV-4-455C and. MOV 751, is a parallel loop, identical in operation and configuration to the primary loop. Since the block valve for the backup loop pressurizer PORV was not closed, the introduction of the artificial high pressure signal opened the backup loop PORV, PCV-4-455C, resulting in the depressurization of the reactor coolant system. The PORV was immediately closed. During the event reactor coolant pressure decreased from 354 psig to 345 psig. MOV-4-751, which had been open prior to the event allowing residual heat removal flow of 3600 gallons per minute, closed during the event. MOV-4-751 was immediately reopened. The residual heat removal flow was reduced to no flow but returned to normal after the valve was reopened. The pump was turned off upon loss of suction flow and the 'B'HR 'A'HR pump was turned on to supply the flow after the valve was reopened.

The reactor coolant system temperature rose 1.'F during the e'vent.

This event was reported to the NRC operations center in accordance with 10 CFR 50.72 (b) (2) (iii)'.

II. EVENT CAUSE The cause of -this event was personnel error. O-OSP-041.4 is a procedure under the control of Operations personnel. Verification of use of the correct portion of the procedure did not occur duping the surveillance. During use of the procedure for the surveillance on the primary loop, I&C personnel used the wrong section of controlling procedure 4-0SP-041.4 to perform. the introduction of the artificial high. pressure signal.

0 LICENSEE ENT REPORT (LER)., TEXT QINTINURTION FACILITY NAME 'DOCKET NUMBER LER NUMBER PAGE NO.

- TURKEY POINT UNIT 4 05000251 92-008-00 03 oF 03 Specifically,, procedure section 7.2.2.10 was performed rather than the required section 7.2.1 '0. This resulted in the artificial high pressure signal introduction to the wrong loop.

III. EVENT SAFETY ANALYSIS The overpressure mitigating system is designed to limit the pressure to which the reactor coolant system is exposed during low temperature operation. During this event the system was actuated'F inadvertently

~

resulting in a slight reduction in pressure and a 1 rise in temperature. No high system pressure actually occurred. The overpressure mitigating system and residual heat removal system responded as expected. After the event, the pressurizer power operated relief valve was closed and the residual heat removal system was returned to operation in a timely manner. Because of the short duration of the event only small changes in reactor coolant system parameters occurred. Therefore, the health and safety of plant personnel and the general public were not jeopardized by the

. actuation of the overpressure mitigating system.

IV. CORRECTIVE ACTIONS

1. The specialist was counselled on the importance of care in the execution of procedures and disciplined in accordance with plant poli'cy.

3/4-0SP-041.4 was 'revised to require a verification of the appropriate block valve closed and power removed from the appropriate residual heat removal 'suction valve. This corrective action will reduce the probability of this type of event in the future.

V. ADDITIONAL INFORMATION No actuations of this type of'he pressurizer power operated relief valve have occurred in the past.

This event was considered reportable in accordance with 10 CFR 50. 73 (a) (2) (v) .

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