ML17261A772

LER 87-007-00:on 871218,during Review of Westinghouse Ltr Re Operating Plant Feedback,Potential for Loss of Cooling During High Head Recirculation Phase Discovered.Caused by Design Flaw.Affected Valve repositioned.W/880118 Ltr

ML17261A772
Person / Time
Site:	Ginna
Issue date:	01/17/1988
From:	Snow B, Widay J ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
LER-87-007, LER-87-7, NUDOCS 8801250358
Download: ML17261A772 (8)

Text

AC CELERATZD DISTRIBUTION DEMONSTRATION SYSTEM.

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:8801250358 DOC.DATE: 88/01/17 . NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet, Ginna Nuclear Plant, Unit 1; Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION WIDAY,J.A. Rochester Gas & Electric- Corp.

SNOW,B.A. Rochester Gas & Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 87-007-00 on 871218,discovery of apparent design

.-inadequacy caused potential loss of core cooling. R W/8 DISTRIBUTION CODE: IE22D COPIES RECEIVED:LTR 3 ENCL TITLE: 50.73 Licensee Event Report (LER), Incident Rpt, etc.

I SIZE:

NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72) ~

05000244'ECXPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 LA 1 1 PD1-3 PD 1 1 A STAHLE,C 1 1 INTERNAL ACRS MICHELSON 1 1 ACRS MOELLER 2 2 AEOD/DOA 1 1 AEOD/DSP/NAS 1 1 AEOD/DS P/ROAB 2 2 AEOD/DSP/TPAB 1 1 ARM/DCTS/DAB 1 1 DEDRO 1 1 NRR/DEST/ADS 1. 0 NRR/DEST/CEB 1 1 NRR/DEST/ELB 1 1 NRR/DEST/ICSB 1 1 NRR/DEST/MEB 1 1 NRR/DEST/MTB 1 1 NRR/DEST/PS B 1 1 NRR/DEST/RSB 1 1 NRR/DEST/SGB 1 1 NRR/DLPQ/HFB 1 1 NRR/DLPQ/QAB 1 1 NRR/DOEA/EAB 1 1 NRR/DREP/RAB 1 1 NRR/DREP/RPB 2 2 NRR IB 1 1 NRR/PMAS/I LRB 1 1 EG ILE 02 1 1 RES TELFORD i J 1 1 EIB 1 1 RES/DRPS DIR 1 1 RGN1 FILE 01 1 1 EXTERNAL: EG&G GROH,M 5 5 FORD BLDG HOY,A 1 1 H ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC HARRIS,J 1 1 NSIC MAYS,G 1 1 8

A TOTAL NUMBER OF COPIES REQUIRED: LTTR 46 ENCL 45

UL. NUCLEAh AEOULATONY IXHOA~

ARAOVEO NN INL Et~&

EXIWNL EO I M UCENSEE EVENT REPORT tLEA)

FACILITY ~ Ill R.E. Ginna Nuclear Power Hant OOCXET 0 5

~h 0 0 ITI 0 2 Q 1 0 0 T'TLE~ Discovery Of Apparent FAUNAE Design Inadequacy Causes Potential For Loss of Core Cooling Durin The Hi LEh ~hd Recirculation UENYIAL IO QONTII Phase 1&OAT CATE Ill OAY OTlIEh lACILITIEEINVOLVEO IEI ABACI UT Y NA~ OOCXET NUNEEXIEI YEAh YEAh N Ih 0 6 0 0 0 0 0 7 0001 17 0 I 0 0 0 1 2 1 wx% ol 8 7 7 8 8 TINE ASfOIlf IE EUEANTTEO PUNEUANT TO TIIE hEIXIIXENENTE tV IO Cfh f. (ONat eae a ~ et VN hWOeaayt It1l

~

N N.TININHlel KNEh ~IHE ~ CENeHII  % TelsHNNI LEVEL IIEI

's<'>vrr'I

'v.

A.;a+m4 S~HIHII AAHUHIIINI

~HI Sll&WIllNI tlirt

~IXI M.TINIIEHE NPINHEHEI NPSslNI INI

~teEt MPSslIEINNIIAI WPQQNHtOHEI NPSlelIEHII OTNEh IEAwNY N

~ NNI ~he T~.NACA~

LICQQEE CONTACT tCM TH% LEh IIXI AIIEA COOl Joseph A. Widay, Technical Manager MANUIAC TUhEh tfOIITAELl "

TO NNIOE CCN&tETE ONE LINE tOO EACH COAOONENT EAILUhE OCNAIEEO IN TNN hEEONT 11$

CAUEE EYETEN aaeONENT IIANUEAC TUN Eh LPtLNAENTALhEIONT EÃftCTEO It& OAY YEAh EXtOCCf0 SJNJ IEQON IIII

~TIIACTICAL N ION NNN, l.a. ~~

YEE (II ye@. cesyJset CÃJVCTEO EUENIENON OA Tfl tlfOee ~ gWVsoce yeaetrae Z(~ IIOI CATE On December 18, 1987 during the review of a Westinghouse Corpora-tion letter entitled "Operating Plant Feedback - Non-vital Power Supply Used in Valve Interlock Logic," it was discovered that the potential existed for a loss of core cooling during the High Head Recirculation Phase. Even though the review determined no suscept-ability to the condition as described in the referenced letter, further evaluation revealed the below described deficiency.

1 The apparent root cause of the event was identified as a design flaw, in that a common power supply was utilized to power a motor operated valv'e on each train of the High Head Recirculation System. A postulated failure of the electrical power supply prior to opening of the subject valves would result in both

.flow. paths leading to the safety injection and containment spray pumps being blocked, creating potential loss of core cooling.

~

(See Ginna USFAR for configuration description.)

Corrective action to prevent recurrence was to position the affected valve in the one-of-two series arrangement to the open position, thereby eliminating the potential for common mode failure.

88Pi25P>58 88Pii7 p5PPP2

@DR ADOCK 9C9 8

U.S, NUCLKAR RKCULATORY COMMISSION NRC Fe//A 094A

<949 I LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPROVEO OM9 NO. 3150&105 K IIPIR K 9'/5 /891 OOCRKT NUMKKR IXI L'KR NUMKKR I ~ I ~ AOK ISI FACILITY NAMK 11I SSOUS//TIAL A II/l5 IQ N

// UM 5 I A

//UMSS A R.E; Gonna Nuclear Power Plant 0 5 0 0 0 007 00 02 OF 0 6 TKXT Ill/I/II/54M't /FSUIFSL I/SP tdMISS/M//F/IC FO//II JRLK't/>>TI I ~ PRE-EVENT PLANT CONDITIONS The unit was at 100% Reactor Power at the time of the discovery.

II'ESCRIPTION OF EVENT A. EVENT:

On December 18, 1987 during the review of a Westing-house Corporation letter entitled "Operating Plant Feedback - Non-vital Power Supply Used in Valve Interlock Logic," it was discovered that the potential existed for a loss of core cooling during the High Head Recirculation Phase. Even though the review determined no susceptability to the condition as described in the referenced letter, further evaluation revealed the below described deficiency.

The apparent root cause of the event was identified as a design flaw, in that a common power supply was utilized to power a motor operated valve on each train of, the High Head Recirculation System. A postulated failure of the electrical power supply prior to opening of the subject valves would result in both flow paths leading to the safety injection and containment spray pumps being blocked, creating potential loss of core cooling. (See Ginna USFAR for configuration description.)

The Residual Heat Removal (RHR) discharge valves MOV 857A, B, C leading to the safety injection and containment spray pump suctions, are normally closed.

MOV .857A is in the 6>> line from RHR pump>>B>>. MOV 857B and C are series valves in the 6>> line from RHR pump>>A>>. MOV 857A is powered from Bus 14 through MCC 1C position 7M. MOV 857B is powered from Bus 16 through MCC 1D position 7M and MOV 857C from Bus 14 through MCC 1C position 152. RHR pumps >>A>> and

>>B>> are powered by Bus 14 and 16, respectively.

N /I C P 0 IIM SSS A I945 I

P e U.S. NUCLEAR RECULASORY COMMISSION NRC Pmm 944A APPROVEO OM9 NO, 9150WI04 I945 I LICENSEE EVENT REPORT ILER) TEXT CONTINUATION EXPIRESEISI/95 PACE ISI PACILITY NAME I'l OOCKEY NUMSER ISI LER NUMEER 14I SSOVSNSIAL ASVISION NSAA NVMSSA NVMSIA 0 OF 0

R.E. Ginna Nuclear Power Plant 0 5 0 TEXT IIP meS <<Mce <<mSNS<<E v44 <<ANbnel HIIC %%dnn ~El 1171 There is an interlock for MOV 857A, B, C which prevents their opening if MOV 850A or MOV 850B are open and AOV 897 and AOV 898 are both open and MOV 896A and MOV 896B are both open. This prevents discharging water from sump B into the Refueling Water Storage Tank (RWST) during sump recirculation.

MOV 857A also has a pressure interlock which prevents its opening when system pressure exceeds 250 psi.

This prevents overpressurizing the safety in)ection suction piping.

During the sump recirculation phase following a LOCA, if high head recirculation is necessary, the 857A, B, C valves must be opened. Present procedure ES-1.3, "Transfer To Cold Leg Recirculation" calls for alignment to the sump and discharging to the reactor through MOVs 852A and 852B. Zf high head recirculation is required due to reactor coolant system pressure being above the discharge capabilities of the RHR pumps, the 857 valves must be opened. A failure of Diesel Generator (D/G) 1A prior to opening the 857 valves would result in both trains leading to the safety in) ection/containment spray pumps being blocked and potential loss of core cooling.

Procedure changes have been submitted to require positioning MOV 857C in.the open position during normal safeguards system alignment. This action provides assurance of at least one flow path (assuming single failure criteria) during the high-head recircu-lation phase.

B. ZNOPERABLE STRUCTURES P COMPONENTS P OR SYSTEMS THAT CONTRZBUTED TO THE EVENT:

None C. DATES AND APPROXIMATE TZMES FOR MAJOR OCCURRENCES:

o December 18, 1987, 1000 EST: discovery date and time NRC PORM SSSA 1945)

NRC fetM SSSA V.S. NUCLEAR 8EOULATOR Y COMMISSION (ILES I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROV EO OMS NO. SISS&105 EXPIRES Sfll/88 fACILITYNAME III OOCKET NUMSER (tl LER NUMSER Itl PACE (SI YEAR NUM tP 'UM

','~: StautNTIAL 8:: IICVl5ION PA R.E. Ginna Nuclear Power Plant o s o o o 244 87 0 07 0 0 0 4oF 0 6 TEXT IP ~ PPNN N MMES. ~ MSSSNNPNRC rO aaCSI OTI D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

As noted, only the potential loss of core cooling due to the failure of the 857 valves was identified. All other valves within these systems were provided with proper train separation.

E. METHOD OF DISCOVERY:

Self-identified through an engineering assessment of a Westinghouse letter regarding the non-vital power supply used in valve interlock logic. Even though there was no susceptability to the condition as described in the letter, further evaluation revealed the above described design inadequacy.

F. OPERATOR ACTION:

None.

G. SAFETY SYSTEM RESPONSES:

None.

III. APPARENT CAUSE OF EVENT A. ROOT CAUSE:

The root cause of the event was determined to be an apparent design inadequacy in that a common mode failure renders both trains of core cooling for the condition of high head recirculation inoperable.

NRC PORM SSSA ISSSI

U.S. NUCLEAR REGULATORY COMMISSION NRC farm SSSA 15451 LICENSEE EVENT REPORT {LER) TEXT CONTINUATION APPROVEO OMS NO, 5150MIOA EXPIRES SISI IS5 PACILITY NAME 111 POCKET NUMSER ISI LER NUMSER ISI ~ AOE Ill v, 55QvorrTrAL rrlvr5rorr NUMOEA NUMO 5 A 0 0 OF R.E. Ginna N ar Power Plant 0 5 0 TEXT TIP mora Rxko il rooMPPIE Mop NARlonol NPC %%drm ~'Pl (171 IV. ANALYSIS OF EVENT This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(v)(D) "any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.>>

An assessment was performed considering both the safety consequence and implications of this event with the following results and conclusions:

0 The RHR pump and 857 valves for each train are not powered by the same source. Pump >>A>> (Bus 14) is in the line with MOVs 857B (MCC 1D) and MOV 857C (MCC 1C). Pump >>B>> (Bus 16) is in the line with MOV 857A (MCC 1C). With this configuration a failure of D/G 1A just prior to opening of 857 valves will cause RHR pump >>A" to stop while 857A (Train >>B>>) and 857C (Train >>A>>) could not be opened from the control room. Although pump >>B>>" would remain operable in this case, Train >>B>> would be unavailable since MOV 857A could not be opened. Manual operator action would be required to mitigate the consequences due to the above described failure.

Failure of MOVs 857A and .857C to stroke to the open position due to the common power supply failure could be alleviated through operator action by manually opening the affected valves. This action, though possible, is not in the interest of ALARA and was not provided in existing procedures.

0 An alternate line has been provided so that high head recirculation can continue in the event that the recirculation flowpath downstream of the 857 valves is unavailable. This design can also be utilized to provide a flow 'path to the 1C safety injection pump piping in the event of failure to open the 857 valves. Two normally closed manual valves in series, Nrlc porIM ssoA IS 45 I

Q,S, NUCLEAR REGULATORY COMMISSION NRC Form 3SSA 104SI LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO. QI 5OMIQE T

EXPIRES I/)I I55 FACILITY NAME III OOCRET NUMSER ITI LER NVMSER I ~ I YEAR SEOVEN'IIAL REVISION NVMOER NUMSSR R.E. Ginna Nu ar Power Plant 0 5 0 0 0 007 0 0 06o" 06 TECT TIP IIRRp <<Mop <<rooMPPIE Mop or<<RRSW HIIC %%drm ~'sl IITI 1816A and B, in a separate 4" line upstream of MOV 857B and C can provide core cooling from the "A" train of the RHR system to the 1C safety injection pump suction. This action is recognized to mitigate valve opening failure only, and not entire train failure.

V. CORRECTIVE ACTION A. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

o MOV 857C has been placed in the open position during normal safeguards alignment to eliminate the potential for common mode failures.

o The Emergency Operating Procedures have been changed, validated, and verified to reflect the new normal safeguards system alignment.

o An Engineering Work Request has been generated to evaluate the current configuration design.

VI. ADDITIONAL INFORMATION A. FAILED COMPONENTS:

None.

B. PREVIOUS LERs ON SIMILAR EVENTS:

A similar LER event historical search was conducted with the following results: No documentation of similar LER events could be identified.

Co SPECIAL COMMENTS:

o The corrective action planned will be tracked by Corrective Action Report (CAR) 5 1831.

o Westinghouse Electric Corporation has been notified of the above apparent design deficiency.

NRC PORM SOSA 1545 I

ROCHESTER GAS AND ELECTRIC CORPORATION e 89 EAST AVENUE, ROCHESTER, N.Y. 14649.0001

, I., le sera co."t ""i 5-'C.27QO January 18, 1988 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 87-007 Discovery Of Apparent Design Inadequacy Due To Susceptibility To A Common Mode Power Supply Failure Causes Potential For Loss of Core Cooling During The High Head Recirculation Phase R.E. Ginna Nuclear Power Plant Docket No. 50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (v) (D) which requires a report of, "any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident", the attached Licensee Event Report LER 87-007 is hereby submitted.

This event has in no way affected the public's health and safety.

~a~

Very truly Bruce A. Snow yours, XC: U. S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406 Ginna USNRC Resident Inspector