ML17263A352

LER 93-004-00:on 930707,Main Feedwater Isolations Occurred Due to Secondary Side & Condensate Feedwater Pressure Transient.Returned Feedwater Regulating Valves to pre-event Controls configuration.W/930806 Ltr

ML17263A352
Person / Time
Site:	Ginna
Issue date:	08/06/1993
From:	BACKUS W H, MECREDY R C ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-004, LER-93-4, NUDOCS 9308120083
Download: ML17263A352 (22)

Text

ACCELERATED DOCUMENT DISTRIBUTION SYSTEM REGUL>RY INFORMATION DISTRIBUTE SYSTEM (RIDE)ACCESSION NBR:9308120083 DOC.DATE: 93/08/06 NOTARIZED:

NO FACIL 50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION BACKUS,W.H Rochester Gas a Electric Corp.MECREDY,R.C Rochester Gas 6 Electric Corp.RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 93-004-00:on 930707,Main feedwater isolations occurred due to secondary side 6 condensate feedwater pressure transient.

Returned feedwater regulating valves to pre-event controls configuration.W/930806 ltr.DISTRIBUTION CODE: ZE22T COPIES RECEIVED:LTR t ENCL r SIZE: LO TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).

DOCKET¹05000244 I i 1 I 1 Dl S l'.05000244 A RECIPIENT ID CODE/NAME PD1-3 LA JOHNSON,A INTERNAL: ACNW AEOD/DSP/TPAB NRR/DE/EELB NRR/DORS/OEAB NRR/DRCH/HICB NRR/DRIL/RPEB NODS SA/+PLB-BP 02 RGHl FILE 01 EXTERNAL: EGGG BRYCE,J.H NRC PDR NSIC POORE,W.COPIES LTTR ENCL 1 1 1 1 2 2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2'1 1 1 1 RECIPIENT ID CODE/NAME PD1-3 PD AEOD/DOA AEOD/ROAB/DSP NRR/DE/EMEB NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRSS/PRPB NRR/DSSA/SRXB RES/DSIR/EIB L ST LOBBY WARD NSIC MURPHYFG.A NUDOCS FULL TXT COPIES LTTR ENCL 1 1 1 1 2 2 1 1 1 1 1 1 2 2 1 1 1 1 1 1 1 1 1 1 D D$D D NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTEI CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT.504-2065)TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!D FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 30 ENCL 30

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~";-',Toe/r., state" 89 EAST AVENUE, ROCHESTER N.K 14649-0001 ROBERT C.MECREOY Vice President Clnna ttuclear Production TELEPHONE ARE/1 COOE 716 5rt6 2700 August 6, 1993 U.S.Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 93-004, Feedwater Control Perturbations, Due To A Secondary Side Transient, Causes Steam Generator High Level Feedwater Isolations R.E.Ginna Nuclear Power Plant Docket No.50-244 In accordance with 10 CFR 50.73, License Event Report System, item (a)(2)(iv), which requires a report of,"any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System (RPS)", the attached event report LER 93-004 i: s hereby submitted.

This event has in no way affected the public's health and safety.Very truly yours, xco Robert C.Mec edy U.S.Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector~r n.h 9308i20083 930806 PDR*DOCK 05000244 8 PDR

NRC FORM 366 (669)V.S.NUCLEAR REGULATORY COMMISSI LICENSEE EVENT REPORT (LER)APPROVED OMB NOA31506104 EXP IR ESI 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFOAMATION COLLECTION REQUEST: 50.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS ANO RFPORTS MANAGEMENT BRANCH (F630), V.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500'l04).

OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, OC 20503.FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET NUMBER l2)PA E 0 5 0 0 02 44 iOFO Fee water Contro ertur at ons, Due To Secon ary Trans ent, auses team Generator High Level Feedwater Isolations EVENT DATE (5)MONTH DAY YEAR YEAR LER NUMBER (61: SEQUENTIAL NUMBER REVISION NUMBER REPORT DATE LT)MONTH DAY YEAR DOCKET NUMBERIS)0 5 0 0 0 FACII.ITV NAMES 01'HER FACILITIES INVOLVED (6)0-7 79 393 0 0 4 0 0 0 806 93 0 5 0 0 0 OPERATING MODE (9)N POWER LEYEL 0 9 7 60 n(sl(2)Is>>)50.n(v)(2)

Nl 50 73(s)(2)lvsl)50.73(v)(2)(vill)(Al 50.73(v)l2l(vill)(B) 50.73(~)(2)lsl 0 THE RLQUIAEMENTS OF 10 CFR ((:/Chock ono or moro Ol trss IollovrinP/

(11)THIS REPORT IS SVBMITTFD PUASVANT T 20.402(B)20.405(~)(1)(I)20.405 (s)l1)(NI 20.406(~I ('I Hill)20405(s)(1)(lv) 20.405(s I (I)(vl 20.405(c)60M(s)(II 50.36(~)(2)60.73(sl(2((ll 50.73(s)(2)(ill 50.7 3(s)(2)(ill)73,71(B)73.71 I cl OTHER ISpscily In 4osnoct tN/oEN onr/ln Tss I, HRC Form 3664/LICENSEE CONTACT FOR THIS LEA (12I NAME Wesley H.Backus.Technical Assistant to the Operations Mana er TELEPHONE NUMBER AREA CODE 3 155 24-444 6 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)CAVSE SYS'EM COMPONENT MANUFAC.TVRER REPORTABLE TO NPADS P jQg CAUSE S STEM Xj'>>AA'RC~C.3>>">>T>>~'" IKU'::hl.'.@<COMPONENT MANUFAC.TVRER EPORTABLE TO NPROS?>>o.>>...%R.:.R6)SA'NO 4 PcA.N:: SUPPLEMENTAL REPORT EXPECTED llc)YES Ill y<<, compiots EXPECTED Sl/SSI/SSION DATE/NO AssTRAcT I(.imit to/400 spscsr, l.s., spproslmsrsly lilusn tinpis.rpocs typssvrinon

/inos/(16)EXPECTFD SUBMISSION DATE (15)MONTH DAY YEAR On July 7, 1993-at approximately 0915 EDST, with the reactor at approximately 974 full power, main feedwater isolations occurred on the"BFI Steam Generator (S/G).These feedwater isolations were caused by overfeeding the"B" S/G, following a secondary side condensate and feedwater pressure transient.

Immediate operator action was to manually control the Feedwater Regulating Valves (FRVs)to restore the S/G water levels and stabilize the plant.The immediate cause of the event was due to a secondary side condensate and feedwater pressure transient.

The underlying cause of the event was determined to be not isolating the 5A heater high level dump valve prior to trouble-shooting.(This event is NUREG-1022 (X)Cause Code).Corrective actions taken or planned are discussed in Section V of the text.NRC Form 366 (669)

NRC FORM 388A (SJ)9)US.NUCLEAR REGULATORY COMMISSION LICENSEE ENT REPORT (LER)TEXT CONTINUATION APPROVED OMB NO.3(504)04 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST: 508)HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F430), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31500)08), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.FACILITY NAME (ll DOCKET NUMBER (2)YEAR LER NUMBER (8)SEOUENTIAL o@IIEVISION NUM E II PAGE (3)R.E.Glnna Nuclear Power Plant TEXT/I/more e/reoe/I rer/uuN/, u>>//I/rme/HRC Frurn 35883/()7)o s o o o 24 49 3 004 0 0 0 2 OF 0 9 LANT COND T ONS The plant was at approximately 974 steady state reactor power with the following pertinent activities in progress: o The"B" All Volatile Treatment (AVT)mixed bed demineralizer was being placed in service per operating procedure T-6.9A (Condensate Polishing Mixed Bed DI Unit Start-up).

Also per T-6.9A, the standby"C" condensate pump start/stop switch was placed in pull stop, and the low pressure heaters condensate bypass valve switch was changed from the auto to closed position.The above switch manipulations were performed to prevent any inadvertent operations during placing the AVT mixed bed demineralizer in service..o The Instrument and Control (I&C)Department was troubleshooting 5A high pressure heater level control problems.DESCRIPTION OP EVENT A.DATES AND APPROXIMATE TIMES OP MAJOR OCCURRENCES:

0 July 7, 1993, 0915 EDST: Event date and approxi-mate time.0 July 7, 1993, 0915 EDST: Discovery date and approximate time.0 July 7, 1993, 0925 EDST: "A" and"B" Steam Generator (S/G)levels restored to pre-event normal operating band.NRC Form 388A (889)

NRC FORM 388A (SJ)9)US.NUCLEAR REGULATORY COMMISSION LICENSEE ENT REPORT (LER)TEXT CONTINUATION APPROVED OMB NO.31600108 EXPIRES: E/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST: 60.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (PW30), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20666, AND TO 1'HE PAPERWORK REDUCTION PROJECT (3160410E).

OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20603.FACILITY NAME ('I)DOCKET NUMBER 12)YEAR LER Nl/MBER (8).'os: sEovENTIAL NVMSER REVISION NVM ER PAGE (3)R.E.Glnna Nuclear Power Plant TEXT///maro g>>co JI nq>>)ed, 1>>o~H/IC FomI 38SAB/(ll) o s o o o 2 4 4 9 3 004 0 0 0 3 QF0 9 B.EVENT On July 7, 1993, at approximately 0800 EDST, an Instrument and Control (I&C)Technician began trouble-shooting 5A high pressure heater level control problems, in accordance with Maintenance Work Order (MWO)9301086.The I&C Technician suspected there was an air leak in the air supply tubing or in the valve positioner for one of the 5A heater level control valves.With concurrence from a Control Room operator, the technician disconnected and plugged the tubing to the normal level control valve, and observed that the high level dump valve (which fails open on loss of air)briefly opened for a few.seconds (until the tubing was plugged)and then closed, Seeing no improvement in air pressure, the tubing was then unplugged and restored to normal.Shortly before 0915 EDST, the technician disconnected and plugged the tubing to the high level dump valve.With air pressure to the valve plugged, the dump valve went full open and remained fully open.Approximately ten (10)seconds later, the technician noted that the Heater Drain Tank (HDT)discharge valve began closing, and expeditious'ly unplugged the tubing and restored normal air supply to the dump valve.Upon restoration of air pressure, the dump valve closed.On July 7, 1993 at approximately 0915 EDST, with the reactor at approximately 974 full power, a secondary side condensate and feedwater system decreasing pressure transient occurred.At this time the configuration of the S/G feedwater regulating valves were as follows: o The"A" S/G Main Feedwater Regulating Valve (FRV)was in manual mode (approximately 454 open)to reduce feedwater system flow oscillations that were occurring.

NR C F oIRI 388A (689)

NRC FORM 366A (SSS)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE ENT REPORT ILER)TEXT CONTINUATION APPROVED OMS NO.31600104 EXPIRES: 4/30/02 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST: 50A)HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P430), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20655, AND TO 1HE PAPERWORK REDUCTION PROJECT (3150d)04), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.FACILITY NAME (<<DOCKET NUMBER (2)YEAR LER NUMBER (6)'SOUSNTIAL NUMSSII 4 IISVIS ION NUMSSII PAGE (3)R.E.Gonna Nuclear Power Plant TEXT///mau Spuce/4 Isqu/>>I/, u>>a/dc0htuh4/I/RC FumI 36643/((Tl 0 5 0 0 0 2 4 4'004 0 0 0 4oF 0 9 o The<<A<<S/G Bypass FRV was in the auto mode controlling the<<A<<S/G level (approximately 804 open).o The<<B<<S/G Main FRV was in the auto mode controlling the<<B<<S/G level (approximately 464 open).o The<<B<<S/G Bypass FRV was in the auto mode (full open).The secondary side condensate and feedwater system decreasing pressure transient was initiated by a significant momentary decrease in HDT pump flow due to the closing of the HDT discharge valve.(HDT pump flow decreased approximately 754 from normal).HDT pump flow normally supplies approximately one third of the suction flow to the S/G main feedwater pumps.The other two thirds of the suction flow to the S/G main feedwater pumps is supplied by the condensate pumps through the low pressure feedwater heaters.The decreased flow coupled with the inability of the standby condensate pump to start and/or the low pressure heaters condensate bypass valve to open, (this line-up explained in Pre-Event Plant Conditions,Section I), decreased pressure throughout the con-.densate and feedwater system.The<<B<<S/G Main FRV (in auto)followed the loss of main feedwater pressure by opening more (approximately 574 open)to maintain main feedwater flow to the<<B<<S/G.The<<A<<S/G Main FRV (in manual)remained at approximately 454 open.Approximately 40-50 seconds after initiation of the transient, HDT pump flow was restored to normal, restoring condensate and feedwater pressure to normal, and causing<<B<<S/G feedwater flow to exceed the calibration range of the flow transmitters

(<<B".S/G FRV had opened to 574).Because the"B<<S/G feedwater flow values exceeded the calibration range of the flow transmitters, the"B<<FRV automatically switched to manual mode.The preceding events occurred in a short period of time.During this short period of NRC FoIm 366A (640) 0 NRC FORM 366A (6BB)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE ENT REPORT (LER)TEXT CONTINUATION APPROVEO DMS NO.3)504(OE EXPIRES: E/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 508)HRS.FORWARD COMMENTS REGARDING BURDEN FSTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (PJ)30), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31504)04).

OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.FACILITY NAME (I)DOCKET NUMBER 12)YEAR LER NUMBER (6)I'oR'j SEQUENTIAL S>NUMBER REVISION NUM ER PAGE (3)R.E.Gonna Nuclear Power Plant TEXT/II mort Epoco/E IEOVPEI/ooo ea//I/ono/HRC FomI 366AB/(17)o s o o o 24 49 3 004 0 0 0 50F 09 time the"B" S/G was being supplied with more feedwater than required, and at approximately 0917 EDST (approxi-mately two (2)minutes into the transient) main feedwater isolation on high level (i.e.>/=674 narrow range level)occurred five (5)times to the"BL(S/G over a period of fourteen (14)seconds.The Control Room operators took immediate manual actions to restore S/G levels and at approximately 0925 EDST the"A" and"B" S/G levels were restored to their normal operating band and the plant stabilized.

Subsequently, the Control Room operators notified higher supervision and the Nuclear Regulatory Com-mission per 10 CFR 50.72, non-emergency 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification.

C.INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT The 5A high pressure heater high level dump valve to the HDT inoperability, due to troubleshooting efforts, in accordance with MWO 9301086, contributed to the event.Inoperability of the"C" Standby condensate pump and condensate bypass valve, due to procedure T-6.9A'equirements, contributed to the difficulty in responding to the event.D.CÃHiER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED: None.NRC FoIIn 366A (649)

NAG FOAM SKSA (5$9)U$.NUCLEAR REGULATORY COMMISSION LICENSEE NT REPORT (LER)TEXT CONTINUATION APPROVED OMB NO.3)500(04 EXPIRES: r/30/92 TIMATED BUADEN PEA RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REGUEST: 50A)HRS.FORWARD COMMENTS REGARDING BUAOEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P$30), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON.

OC 20555.AND TO'IHE PAPERWORK REDUCTION PROJECT (31500(04), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503 FACILITY NAME (II DOCKET NUMBER (2)YEAR LER NUMBER FS)SKQVKNTIAI NVMSKR RKVIKION NUMSKR PAGE (3)R.E.Gonna Nuclear Power Plant TEXT///mao r/vcr/r rrqokaf, uw edcVdoIN/HRC Fomr 35SA'r/(Ill o s o o o 2 449 3 004 0 0 0 6 OF 0 9 E.METHOD OP DISCOVERY The event was immediately apparent due to alarms and indications in the Control Room and indications at the 5A high pressure heater high level dump valve to the Heater Drain Tank.OPERATOR ACTION: The Control Room operators took immediate manual actions to control S/G levels and stabilize the plant.Subsequently, the Control Room operators notified higher supervision and the Nuclear Regulatory Commission per 10 CFR 50.72, non-emergency 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification.

G SAFETY SYSTEM RESPONSES:

The>>B>>S/G Main and Bypass FRVs began to close automatically as a result of the feedwater isolation signals.Due to the short duration that the signals were present, the FRVs never fully closed.III.CAUSE OP EVERT A.IMMEDIATE CAUSE: The feedwater isolation signal to the>>B>>S/G Main and Bypass FRVs was due to the>>B>>S/G narrow range level being>/=67%.B.INTERMEDIATE CAUSE: The>>B>>S/G narrow range level was>/=674 due to increased flow to the>>B>>S/G caused by the perturba-tions in main feedwater header pressure and automatic FRV operation.

This situation resulted in overfeeding the>>B>>S/G.NRC FomI 358A (5$9)

NRC FORM 366A (609)US.NUCLEAR REGULATORY COMMISSION LICENSEE ENT REPORT (LER)TEXT CONTINUATION APPROVED OMB NO.31500104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPI.Y WTH THIS INFORMATION COLLECTION REQUEST: 60gl HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (PJ)30), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON.

DC 20503.FACILITY NAME (II DOCKET NUMBER (2)YEAR LER NUMBER (6)(I63: SEQUENTIAL NUM664+p REVISION NUM 64 PAGE (3)R.E.Gonna Nuclear Power Plant TEXT///RNVP 4/>>Ce/4 JNIUPN/II>>PIS/OR4/HRC FORR 36542/((7)0 s 0 o 0 2 4 4~3 004-0 0 7oF 09 The perturbations in main feedwater header pressure were caused by a momentary large reduction in HDT pump flow followed by a rapid return to normal flow.The"B" S/G Main FRV (in auto)followed the loss of main feedwater header pressure by opening more to maintain feedwater flow and when HDT pump flow was rapidly restored, main feedwater header pressure returned to normal.When main feedwater header pressure returned to normal, the"B" S/G FRV was open approximately 114 more than normal and the"B" S/G feedwater flows exceeded the calibration range of the flow transmitters and by design the"B" S/G FRV switched to manual mode.In the manual mode'he"B" S/G Main FRV could not control the"BLI S/G level without operator intervention.

It should be noted here that if the low pressure heaters condensate bypass valve had been capable of opening, the above transient could have been less severe and the"BFI S/G would most likely have not sustained a high level feedwater isolation.

The momentary large reduction in HDT pump flow, followed by a rapid return to normal flow, was caused by the opening and subsequent closing of the 5A high pressure heater high level dump valve to the HDT, due to the actions of the I&C Technician's troubleshooting activities.

It is believed that the opening of this dump valve, coupled with the 5A high pressure heater low level condition, decreased the HDT level substan-tially.This level decrease was sensed by the HDT level control system and it began to close the HDT pump discharge control valve to reduce flow and return the HDT level to its operating band.The closing of the 5A high pressure heater high level dump valve, due to the actions of the ISC Technician, reversed the above conditions and the HDT pump flow returned to normal.NRC FoIRI 366A (64)9)

J 0 NRC FOAM 368A (649)FACILITY NAME (1)U.S NUCLEAR REGULATORY COMMISSION LICENSEE ENT REPORT ILER)TEXT CONTINUATION DOCKET NUMBER (2)APPROVED 0MB NO.31504(OS EXPIRES;S/30/92 ESTIMATED BUADEN PER AESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50J)HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F430).U.S.NUCLEAR REGULATORY COMMISSION.

WASHINGTON, DC 205S5, AND TO THE PAPERWORK REDUCTION PROJECT (3(504104).

OFFICE OF MANAGEMENT AND 8UDG ET, WASHINGTON.

DC 20503.PAGE (3)LER NUMBER IS)R.E.Glnna'NUclear Power Plant TEXT///mare Saece/e re/a/rerL Iree~HRC Farm 355AB/(12)YEARo s o o o 24 493 SEQUENTIAL.oA NUMSSR CN 004 REVISION NUM SR 0 0 0 8 oF 09 ROOT CAUSE: The underlying cause of the event was determined to be not isolating the 5A high pressure heater high level dump valve to the HDT prior to commencement of troubleshooting the air supply concerns for the'dump valve.ANALYSTS OP&TENT This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(iv), which requires reporting of,"any event or conditi'on that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF)including the Reactor Protection System (RPS)".The feedwater isolation of the"B" S/G was an automatic actuation of an ESF system.An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:

There were no operational or safety consequences or implications attributed to the feedwater isolations because: o The feedwater isolations occurred at the required S/G level.0 0 The plant was quickly stabilized and manual control of FRVs was accomplished to mitigate any consequences of the event.I, As the feedwater isolations occurred as designed, the assumptions of the FSAR for steam line break were met.Based on the above, it can be concluded that the public's health and safety was assured at all times.NRC FomI 365A (589)

NRC FORM 366A (BJIS)UA.NUCLEAR REGULATORY COMMISSION LICENSEE ENT REPORT ILER)TEXT CONTINUATION APPROVED OMB NO.3(500106 E XP I R ES: 6/30/62 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST: 60.0 HRS.FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F430), U.S.NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1ME PAPERWORK REDUCTION PROJECT (3(500106).

OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.FACILITY NAME (1)DOCKET NUMBER (2)LER NUMBER LSI YEAR r~r'SEOUENTIAL P~S REVISION NUMBER"~NUM ER PAGE (3)R.E.Gonna Nuclear Power Plant TEXT///mom Jpon/1 mqII/mI/II'////JJN/I HRC%%dmI 36M B/l(7)o s o o o 2 449 3 004 00 0 90F V.CORRECTIVE ACTION A ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: After the S/G levels were stabilized, the FRVs were returned to their pre-event control configuration.

B.ACTION TAKEN OR PLANNED TO PREVENT RECERMFNCE:

To ensure a more rapid mitigation of this type of event and possibly eliminate the feedwater isolation, operating procedure T-6.9A will be changed to delete the steps that place the low pressure heaters condensate bypass valve switch to the closed position.Operations, Planning, and I&C personnel have been notified that isolation of a heater high level dump valve, prior to troubleshooting, should be considered a normal practice.ADDITIONAL INFORMATION A.FAILED COMPONENTS:

None.B PREVIOUS LERs ON SIMILAR EVENTS A similar LER event historical search was conducted with the following results: LER 91-009 and LER-92-006 (Revision 1)were similar events with different root causes.C SPECIAL COMMENTS: LER 92-006 (Revision 1)indicates the problems that have been experienced with the Advanced Digital Feedwater Control System (ADFCS), the causes and corrective actions taken or planned.NRC Form 366A (686) h T