ML17262A292

LER 90-017-00:on 901212,reactor Trip Relay de-energized & Reactor Tripped When Dc Switches in Distribution Panel Opened.Caused by Procedural Inadequacy.Procedure Change Process Being evaluated.W/910111 Ltr

ML17262A292
Person / Time
Site:	Ginna
Issue date:	01/11/1991
From:	BACKUS W H, MECREDY R C ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-017, LER-90-17, NUDOCS 9101160181
Download: ML17262A292 (20)

Text

ACCELERAI I=D D!S~RIBU t ION DEMONSTP 4.TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9101160181 DOC.DATE: 91/01/11 NOTARIZED:

NO FACIL:50-244 Robert Emmet'Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION

'BACKUSjW.H.

Rochester Gas&Electric Corp.MECREDY,R.C.

Rochester Gas&Electric Corp.RECIP.NAME RECIPIENT AFFILIATION

.DOCKET 05000244 R

SUBJECT:

LER'90-017-00:on 901212,reactor trip relay de-energized

&reactor tripped when dc switches in distribution panel opened.Caused by procedural inadequacy.

Procedure change process being evaluated,W/910111 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR I ENCL SIZE: 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).

D 05000244 A RECIPIENT ID CODE/NAME PD1-3 LA.JOHNSON,A INTERNAL: ACN W AEOD/DSP/TPAB NRR/DET/ECMB 9H NRR/DLPQ/LHFBll NRR/DOEA/OEAB NRR/DST/SELB SD NRR/DST/SPLBSDl REG F RGN1 01 EXTERNAL EG&G BRYCE I J~H NRC PDR NSIC MURPHY,G.A 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 3 3 1 1 1 1 RECIPIENT ID CODE/NAME PD1-3 PD AEOD/DOA AEOD/ROAB/DS P NRR/DET/EMEB.7E NRR/DLPQ/LPEB10 NRR/DREP/PRPB11 NRR/DST/SICB 7E NRR/DST/SRXB SE RES/DSIR/EIB L ST LOBBY WARD NSIC MAYS,,G NUDOCS FULL TXT COPIES LTTR ENCL 1 1 1'2 2 1~1 1..-1 2 2 1 1 1 1 1'1 1.1 1 1'R jV t/!IYg~jl'i 9 Xr~NOTE TO AI.I"RIBS" RECIPIENTS:

D D PLEASE HELP US TO REDUCE lVASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOXI P!-37 (EXT.20079!TO ELlb IINATE YOUR NAil!E FROM DISTRIBUTION LISTS I OR DOCUb'IENTS YOU DON'T NEED!FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQU1RED: LTTR 31 ENCL 31 ROCHESTER 8AS AN'i Ef.L'RI CORPORATION 89 EAST AVENUE, ROCHESTER N.Y.14649.0001 yola$1k1C ROBf Rl f.s" c V<<e f're<dt" Cit, Tf.EPjtQN.AREA cGiJf'7't6 546 270 January 11, 1991 U.S.Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 90-017, Opening of DC Switches (Procedural Inadequacy)

Disables Manual and Auto Actuation of Safeguards Sequence Initiation Causing a Condition Outside the Design Basis of the Plant R.E.Ginna Nuclear Power Plant Docket No.50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(ii)(B), which requires a report of,"any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded, or resulted in the nuclear power plant being in a condition that was outside the design basis of the plant", the attached Licensee Event Report LER 90-017 is hereby submitted.

I'his event has in no way affected the public's health and safety.Very ix ugly youk.8, xc: Robert C.Me redy/U.S.Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector g Pu p(g.Z(pP 9101160181 5101ii PPFy A"tAI Y n~rirt,r"~.i&DR 0

HAC Arw 500 ($441 LI C EN S EE EVENT REPORT tLER)VL NUCLCAA AIOULATOAQ COAWI~AAYAOVIO OU!HO.SIN 0<OV CIS<ACS I/SQI 5 5ACILITY HAUC 111 OOCAIT NUMSCA Ol 0 6 0 o 0)24410Fl R'nna Nuc1ear Power Plant 1 IT LC 14I Qponmq, of DC Switches Dz.sables Manual and Auto Actuation o Sa egu s Se ence I'nitiation Causin a Condition Outside.the Desi'asis of the Plant UOHTH OAY YCAA YIAA IVCHT OATS IN LIA NVMICA IN~I OUI AYIAL>>UUYIA MY'>>>>UUSI 1 AtMAT O*TI ITI UOHTH OAY Y CAN~ACILIYY>>AMIA OOCKCT HUll~CA(ll 0 6,0 0 0 OTHCA IACILITICI INVOLVCO OI 1 2 1 2 9090 0,1 7 0 0 01119 1 0 6 0 0 0 OAC AATINO UOOC I~I~OYI C1 LCVIL p p 10.405 III 50.500 4(I(I II 50 A00 4 I I I I I I I TOPIC 4(l I I (W I 50AOS 4 I (11(HI 50.50S 41(1 llrl TOAOII~I NM(al Ill~OMIAIOI N.T 54(O I II N.T54((5(ISI

~I I C4(QI IWI N,T 54)Q I I HI N.T 54(OI I VI~O.T54(QI(vt(l0254IOIIYWIIAI N.1 5 4 l(5 llvW I I~I IO.TSNIQI(A(0 THI AtOVI1CMtNTI 0>ISCSA$!ICMYA rv r rrr Al IM YA>>rr>>51 111 THIS ACKIAT Il SUIMIIICO TUASVAHT'I TSJIWI 5 1514(OTHI1 (5Ar>>r 4 AAvvrr~>>or AAI M TAAL>>AC lrr JICAI HAUC LIC INC C I CONTACT SOA THIS LCA (ill Wesley H.Backus Technical Assistant to the Operations Mana er TCLCTHOHC HUMSIA AAIA COOI 31 524-446 COM5LCTC OHC LIHC SOA CACH COMM>>tNT SAILUAC OIICAIIIO IN THI~ACMAT (ill CAU5C SYSTSU COMMHCHT UANUYAC TVACA C~"ILC"'">'.>~~:.-,1 C Ull lrSICM~'jPi44~+'v+(+Ir rj(Q CAJ,.I A,IV 5lO.UVIOL 1+cvw COMMHCHT MANVSAC TVACA CMATAIL TO HtAOS g g>~$~(5(ÃrP A~SUYYLCMCHTAL ACMAT I trtCTCO II4l CX~CCICO SV I M I 55 IO H OATC II II MONTH CAY'YCAA Ill III rr.Y>>vU>>H CIIPTCTEO 5UCU(55(O>>

OA TCI HO AJCTAACT IL>>A>>N I 500 Mecr, I I, AArv srvswr rtrvA~YAAUY npvrrw>>A AANI (IN)On December 12, 1990, at 2310 EST, with the reactor at approximately 3%full power, the Control Room Foreman opened two'='tc)'e , cd by a Mainte>>.rce procedure, cav~inL)disabling of manual (pushbutton) and automatic actuation of the safeguards sequence initiation.

The two DC switches were closed, as directed by the Maintenance procedure, approximately twenty (20)minutes later, restoring manual (pushbutton) and automatic actuation initiation.'he underlying cause of the event was procedure inadequacy due to insufficient attention to detail.A Extensive corrective actions are being taken to prevent recurrence, including communication of management expectations, HPES evalua-tions, identifying procedural'nadequacies, and a comprehensive upgrade of the procedure change process.>>AC/rs 50C.la

~~~~~SRRMRRHSKHH8i%

I I I~~~I~~~N~~I~0~~~~~~~~~~~~0 0 0 0 f~e~~~-~~~~~~~~-~~~~~~~~~~~~~~

~~REEHQH8895%%9PiB I~l~r y~$g g$g~'~~~o~-~0~~~~~~~~~~~~~J~~~~~~~~~~~~~~~~e~~~~~~~~

)'~A~-~~~~~~~-~-~'~~~~--~-I~I~I--~-~~~~~~~~I o~~s~~~~~~~~~~~~

IIAC SIUttt SSSA 19451 LICENSEE EVENT REPORT ILER)TEXT CONTINUATION V.S.IIVCLKAA ASOVLATOAT COUUISSIOII ASSAOV50 OU9 AO 515OMIOS 5IISIA55'9/SI I95 SACILITY IIASIC ill OOCKST IIVUSSA ITI LTA AVSSICA ICI SSOVSIITIAL U 1 IIS U IS IO It lI U~AOS ISI R.E.Ginna Nuclear Power Plant TTXT III'IIOIS AUSS tI ISSIUSS.USs NASOSIUS ArAC IItttlt WS'll I ITI I Ol 24 49IQQ1 7 00 Qi5oF l i5 The Control Room operators immediately.performed the applicable actions of E-0 (Reactor Trip or Safety Injection) and ES-0.1 (Reactor Trip Response)and stabilized the plant in hot shutdown.I After completing the applicable steps of E-0 and ES-0.1, the Control Room operators completed their part of M-48.14, by closing the two DC switches that had been opened in step 5.5.1 of M-48.14.This was accomplished at approximately 2330 EST, December 12, 1990.The oncoming SS, who had been in the Control Room during this event, resumed the evaluation of the consequences of alarm L-31 after'plant conditions had stabilized.(The cause of the alarm had already been determined.

)He performed another review of M-48.14 and called other knowledgeable members of the plant staff at their homes (at approximately 0100 EST, December 13, 1990)to discuss his concerns about the effect of opening these two DC switches.After receiving confirmation that'his concerns were legiti-mate, he made the proper notifications to higher supervision and the Nuclear Regulatory Commission (NRC).3.NOPERABLL'TRUCTURAL&

s COMPONEN'1'6 s OK SYSTEI'sh THA'J.CONTRIBUTED TO THE EVENT: D.None.OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED-None..E.METHOD OF DISCOVERY:

The event was made apparent during the oncoming SS review of the consequences of Control Board Alarm L-31 (Safeguards DC Failure)and subsequent discussions with knowledgeable plant staff.~IAC SOASS SOSA l9451 MAC lotm 9$$A I943I LICENSEE EVENT REPORT ILERI TEXT CONTINUATION

.II.9.HIICLTAII 1$4ULATOIIY CO>>AII9$IOH A9PIIOVlO OM9 HO$I$0WIOa XXPIA$$9/$I 4$SACILITY NA>>l III OOCrlT eu>>9$A LTI vtAA L$1 HII>>9$II ICI 9 I QVl NTI AL~tVu ATVISIOH V TA AAOl ITI R.E.Ginna Nuclear Polar Plant TTXT I~~>>eccl~.

v>>e<<rWW+ACSn

~'IIIITI oIoI24 490-017-00 0'6oF1 5 F.OPERATOR ACTXON: Factors that influenced operator actions, during the event were as follows: v The Control Room operators questioned step 5.5.1 in procedure M-48.14, but information in M-48.14, the DC switch labels, and Alarm Response procedure AR-L-31 did not provide sufficient operational information to determine the consequences of opening these two switches.o The Control Room operators had confidence in a Plant Operating Review Committee (PORC)approved procedure that had also been reviewed by the Electrical Planner.As the event was over prior to discovery, no operator actions other than normal were performed.

G.SAFETY SYSTEM RESPONSES:

None.XXI.CAUSE OF EVENT A.IMHEDlATE CAUSE: A condition outside the design basis of the plant was caused by the disabling of manual (pushbutton) and automatic actuation of the safeguards sequence initiation (i.e.auto and manual SI).B.INTERMEDXATE CAUSE: The disabling of manual (pushbutton) and automatic-actuation of the safeguards sequence initiation was-caused by switch gl2 in the 1A DC Distribution Panel and switch g9 in the 1B DC Distribution Panel being open at the same time.Both of these panels are on the back of the Main Control Board.MAC>414 999A (9WI

~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~0~~~~~~\~~~~~~~~~~~~~~~~~~~~~~~

a~t~~~r r~~~~~~~~~~~~~~0~~~~~~~~~~~~~~~~~o~~~~~~~~

, 0 HRC>wiA SAEA IE4S I AACILITY IIAME (II LICENSEE EVENT REPORT ILER)TEXT CONTINUATION COCKET iIUMEER IEI V.t.IIVCLEAR REOULATORY COMMISSION i'ttROYEO OME HO SISO&I04 EIItIRES'EPICS LER HVMEER Iti SEQVERTiAL M EA AtVitiOH HQIJ EA tAOE ISI R.E.Ginna Nuclear Power Plant TEXT IA'i>>it A>>i>>e~.v>>AiAAOCrW NAC Aiiiii~'el I ITI o 5Io(oIo gI4 4 9 0-OI1i7-0 0 90F15 The effect of the potential delay in actuating safeguards equipment upon those events analyzed in the UFSAR was evaluated.

The accidents effected by this action are those accidents which result in depressurization of the primary system causing SI.These are primarily the following:

0 0 0 FeedLine Break (FLB)Steam Generator Tube Rupture (SGTR)Small Break Loss of Coolant Accident (SBLOCA)o Large Break Loss of Coolant Accident (LOCA)o Small Steam Line Break (Small SLB)o Large Steam Line Break (SLB)An analysis of these accidents was performed to determine the effect of the disabling of manual (pushbutton) and automatic actuation of the safeguards sequence initiation with the following results: Feed Line Break t This accident was analyzed by the Ginna Updated Final Safety Analysis Report (UFSAR)as a heat up event with auxiliary feedwater available in ten (10)minutes.As a heatup event, RCS pressure never decreased below the SI setpoint, but rapidly increased above the SI pump shutoff head.Therefore, SI was not necessary and auxiliary feedwater, when available within ten (10)minutes, is sufficient to mitigate the event.Operator actions to start auxiliary feedwater within ten (10)minutes is consistent with the Ginna licensing basis.If the FLB was re-evaluated as a cooldown event from 34 power the results would be bounded by a SLB.RRC AORM SEEA it AS I IIAC laew 494A I941I LICENSEE EVENT REPORT ILER)TEXT CONTINUATION V.4.IIUCLSAA ASCUL*TOAY COMMI44IOII A99AOYSO OM4 IIO SI SO&104/4)e+IIIK$4ISI'4$9 ACILITY IIAM4 (Il OOCIIST IIVM44A (1I LSA MVM44II I~I S~QUSHTIAL 4UM A ASVIQl08~i Q 9A~AQ4 ISI R.E.Ginna Nuclear Power Plant TEXT lll~CWCe M~,~AAAIMAM WIC AtW AM'll I Ill 0 5 0 0IO 2 4 4 90-017-00,10 oFl 5 Steam Generator Tube Ru ture SGTR is bounded by SBLOCA from the RCS depressurization standpoint.

The leak rate from a SGTR is small compared to break flow for a SBLOCA.There is no significant effect due to lack of manual (pushbutton) or automatic SI since the main steps in the procedure deal with isolation of the ruptured SG, depressurization of the RCS, and termination of SI.Small Break Loss of Coolant Accident When manual (pushbutton) and automatic SI was de-activated, the reactor was operating at 34 power.The reactor had been at 3%power for approximately ten (10)hours.Prior to that, the'eactor had been subcritical for twenty-two (22)hours following a trip.Westinghouse Owner's Group letter WOG 90-113, dated July 2, 1990,"Shutdown LOCA Program-Draf<Report", evaluated a mode 4 LOCA using a generic two (2)loop plant with a six (6)inch break assumed to occur two and a half (2.5)hours after shutdown.Acceptable results were obtained provided SI was started ten (10)minutes after the break.Assumptions of the mode 4 LOCA analysis are compared with the Ginna Event conditions below: WOG MODE 4 GINNA EVENT Decay Heat 1.34 No accumulators available RCS pressure.1000 psig RCS temperature 425 F Decay Heat 0.864 Accumulators available RCS pressure 2235 psig RCS temperature 547 F The availability of accumulators and the lower decay heat offset the higher RCS temperature and pressure.Sufficient time~s available to manually start the.SI and RHR pumps and open appropriate valves from the Control Room, and to recover from the SBLOCA.In any case, SBLOCA is bounded by LOCA because less time is available for operator action during a Large Break LOCA.4AC 90AM 994A<9A01 MAC eOrm SSSA 114SI LICENSEE EVENT REPORT ILER)TEXT CONTINUATION V.S.HVCLSAA ASOuLATOAv COMMiSSIOle AeeAovso OMs Ho sl so&Ice r See>ASS SISI4S I'ACILITY eIAMS III OOCIIST HUMOSII (11 LSA MuMSSII ISI SSCMSHTIAL M 1 AS Q 4 10 4 U~AOS ISI R.E.'Ginna Nuclear Power'Plant TSxr nr eeee Meee e~.we eeeMeew'AC

~Xa4 Tv I ITI 0 5 I0 0 io 2 4 4 90-017 0 IO 1)1 os'I 5 Lar e Break Loss of Coolant Accident An assessment of disabling manual (pushbutton) and automatic SI at 3%power was performed by Westinghouse with respect to the LOCA analysis.The assessment assumed the RCS was at 547 F, 2235 psig.The fuel rods were assumed to be at 600 F which would be the approximate pellet and clad temperature at the end-of-blowdown phase.The vessel lower plenum and the lower portion of the core would be covered with accumulator water.Further, it was assumed that SI must be initiated when the fuel rods are at 1800 F to turn around the cladding temperature before it reaches 2200 F.Decay heat is based on an approximation of power history prior to the event, using the 1971 ANS Model.An adiabatic heatup calculation was performed using properties for a 14 x 14 array Optimum Fuel Assembly (OFA).The calculation indicated SX was necessary in 5.5 to 6 minutes.Simulations on the Ginna specific simulator indicate a 5 to 6 minute operator response duringa LOCA is achievable.

Small Steam Line Break This accident is bounded by the Large SLB because longer times are available for operator response.Lar e Steam Li:ne Break Westinghouse assessed the effect of no manual (pushbutton) or automatic SX on the Steam Line Break analysis.Based on their experiences with Steam Line Break analysis as well as a review of the available margin to the acceptance criteria, it was judged that if the accident were re-analyzed at 34 power with no manual (pushbutton) or automatic SI, acceptable results would be obtained.~AC eo14 sssA IS 4S I

0 MAC Sarw$SEA 104$1 LICENSEE EVENT REPORT ILERI TEXT CONTINUATION U,L MIJCLEAII AEOULATOAV COMMITEIOII I APPROVED OME IIO$1$0&IOJ E JTPi A E$8 JT I 4$SACILITY IIAME III DOCKET IIU4HEA ITI LEII IILNNEII IEI SEOVS JJTJAL IvIJU 1 1 PEV>SIC U UVU SA PACE 1$I R.E.Ginna Nuclear Power Plant TEXT JJJ eOrP JPPPP A newer.UPS PJ>>1>>MJ JTAC M Ja4'Il lltl 0 5 0 0 0 2 4 4 90-'017-00 12 OF Rochester Gas and Electric Corporation (RG&E)performed a computer analysis of the SLB using the Westinghouse LOFTRAN Code.A base case was compared to a case where SI was delayed for ten (10)minutes.The comparison indicated negligible change in, minimum DNBR.There was an insigni-ficant change in mass released to containment because mass release is dominated by initial steam generator level and auxiliary feedwater flow, neither of which are affected by delayed SI.Comparing energy out the break for both cases, showed negligible differences.

Therefore, delaying SI has negligible effect on minimum DNBR and mass/energy out the break.In conclusion, delay of manual (pushbutton) and automatic SI with the reactor at 34 power would not cause Non-LOCA events to'xceed the acceptance criteria.A delay of 5.5 to 6 minutes in the LOCA can be tolerated without unacceptable results.Based on operator training, this is sufficient time for operator response.Based on the above, it can be concluded that the public's health and safety was assured at all times.V.CORRECTIVE ACTION A.ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: The affected system was restored to normal when the two (2)DC switches were closed twenty (20)minutes after they were opened.%AC SCAM SSAA I$431

~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

l~~r~~I t~~~~0~0~~l 0~~~~~~~~~~~~~~-~~~~~~~~~~~0~~~0~~~~~~0~~-~~~~~~~~~~~~~~~~~~~~~~

>>~~s~i i i~I I I I~~~~~~~~~~~~~~~I I~~~~~~I>>~Ol~~~~QI'~~~~~~~~~~~~~~I IB I