ML20024H228

LER 91-003-00:on 910421,actuation of EFW Sys During Plant Heatup Occurred Due to Low Once Through Steam Generator Level.Caused by Leaking Feedwater Recirculation Valve.Plant Startup Procedure OP 1102.02 Will Be revised.W/910521 Ltr

ML20024H228
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/21/1991
From:	James Fisicaro, Scheide R ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1CAN059107, 1CAN59107, LER-91-003-02, LER-91-3-2, NUDOCS 9105300251
Download: ML20024H228 (6)

Text

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Operations Hay 21, 1991 1CAN059107 11 S. Nuclear Regulatory Commission Document Control Desk Hall Station PI-137 Washington, D. C, 20555

SUBJECT:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 I,1 cense No. DPR-51

' Licensee Event Report 50-313/91-003-00 Gentlemen:

The enclosed 1,1censee Event Report is bcIng submitted in accordance with 10CFR50.73(a)(2)(iv),

Very truly yours,

, .,,.4 [.hM James , Fisicato Director, I.icensing JJF/RilS/mmg Enclosure cc: Regional Administrator Region IV U. S. Nuclear Regulatery CommJssion 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 INPO Records Center Suite 1500 1100 Circle, 75 Parkway Atlanta, GA 30339-3064 4

9105300251 910521 POR ADOCK 05000313 ((2 L 5 PDH

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NRC Fot!n 3/6 U.S. Nucimr Regulatoiy Ccmissim (6-fD) , Ar5mval (lib No. 3150-0104 Txpires: 4/30/92 L1CENSEE EVENT REPORT (1. E R)

FACIL11Y NAME (1) Arkansas Nuclear Om, Unit One IHMir Nlti!!!R (2) PKIE (3)

Autmatie _

Ojsjo[0j0l_3{_1l_31](Fjoj5 TITIE (4) Actuatim of tim Frergmcy ren! water Systm Dtriig Plant limtup due to lm Qice throgh Stem Gmerator I E01 Wil idLResultai Frun a Icaking Fmtwater Reciaulatim Valvo.

31hT IMIE (5) UREDER (tl M;l%T IMIE f 7) UMR_13C&lI1DSJh%1Lil;l)JO Sapimtial Revisim tkoth Ihy_ Ymr Ymr Nimaler Ntsnier tkuth Diy_ Ymr Facility)mos _ Ihlet_Numlpr(sL QLQ LO 0j4_ _2]l 9 1 _9{l-- 0j.0[3 --

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ti1E (9) N (Check me or more_ of tic _fgllwine) (11) 1%1R _ 20.402(b) _ 20.405(c) X._ 50.73(a)(2)(iv) _ 73.71(b)

IEVEI, _ 20.405(a)(1)(i) _ 50.36(c)(1) _ 50.73(a)(2)(v) _ 73.71(c)

.(10)_Oj0}0 ___ 20.405(a)(1)(ii) _ 50.36(c)(2) _ 50.73(a)(2)(v11) _ Other (Specify in

?0.40Va)(1)(ilf) _ 50.73(s)(2Xi) _ 50J3(aM2XvHiM A) Nrtr ct !n!x n:rl

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_ 201 405_(aKl)[v)__ _ 50.73(aX 2X 111] 50.p[aX2).(x)__ _ 366A)

IJCIETE GhTACT RE 11115 IIR (12)

Name Telminn Nuder Ar m Richant 11. Scheide, Nucimr Safety at:1 Licmsing S i ncialist Caio 5J91191614Hs191ql9 Ogil]ITEJhE_QNE FIR FACll (UiIUJNr FA1!L'RE IFSfKIBElliN_IllIS 104T _(13)

Retortable Regortable Cause Svnis CcummLJ130ufactitrer to !qME Gssn Systm _ficuoteuL thrmisturer to NIME X _B! A _1lSlV{ V! Ol &! 5 Y l 1 l I l l l L_ J_lJ_ _LLl 1 J_LL J_LL_

_1mr StjlHIET REITT EXIEllD (14) EXITIID _tkulh. Ihv SLWi!SSICN

_LLYes Gf yes, cmmt.tranctmi Sur issim nittlxl No st_a sL _.l_ _L __L ABS 1RACT (Limit to 1400 spaces, i.e. , apprmimately fiftem single-since ty;writtm lines) (16) j On April 21, 1991, with the reactor subcritical, the Emergen<.y Feedwater System i (EFW) was automatically actuated due to a low level in the "A" Once Through Steam Generator (OTSG). At the tin,e of the event, OTSG levels were being lowered to 30 inches in accordance k hh the "Plent Startup" preceduro. However, when the "A" OTSG reached its programmed level, the auxiliary feodwater pump (P-75) was unable to maintain the required level. OTSG level continued to decrease until the EFW system automatically actuated (13 inches). The EFW system operated as designed and quickly returned the OTSGs to their programmed level. The cause of this event was leakage through a closed valve (FW-8A) in the "A" OTSG feedwater recirculation 1ine which resulted in a partial void in the feedwater line to the "A" OTSG. This void caused P-75 to reach a " runout" condition which rendered it incapable of feeding the OTSGs.

FW-8A will be repaired prior to heatup from the next refueling outage. In addition, the " Plant Startup" procedure will be revised by July 31, 1991 to include steps requiring backup isolat. ion for FW-8A and B and verifying that P-75 is capable of feeding the OTSGs prior to reaching 30 inches during hentup.

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NHC Fohn 366A' U. S. Nuc1 car Regulatory Omnissian 16-89) Appmm101B No. 3150-0104 Expires: 4/30/92 LIONSEE ININf RElWf (!JR) DAT (mTINUATim F/CILFlY NVE (1) (IEf7 MHNR (2) 11R RMER (6.) PATf. (3)

Sequential Revisim Artansas Nuclear One, Unit One Year Nab 1____l@lrL Oj5j0}0j0lJlJj3.3]1-- 0j0}3 --

0_j_0_ Oj2l@}0h UAT (If mre space is rnluirni, use ailithnal NRC Fonn 366A s) (17)

A. Plant Status ,

. t the time of this event , Arkansas Nuclear One, Unit One ( ANO-1) was preparing for startup following a maintenance outage. The reactor was subcritical with the Group 1 control rods withdrawn to their upper limits to establish immediate negative reactivity addition capability. Reactor Coolant System (RCS) [AB]

pressure was approximately 2250 psig for an elevated system pressure walkdown and-temperature was 538 degrees. Once Through Steam Generator (OTSG) pressure was approximately 900 psig.

B. Event Description On April 21, 1991, at approximately 2150, the Emergency Feedwater System (EFW)

[BA] was automatically actuated by the Emergency Feedwater Initiation and Control system (EFIC) in response to a low level in the 'A' OTSG.

The EFIC system monitors OTSG 1evels and pressures,fiain Feedwater pump status, reactor coolant pump (RCP) status and Engineered Safeguards Actuation System

[JE] channels 3 and 4 in order to initiate EFW or OTSG isolation should an actuation setpoint be reached. The EFW system, which includes a motor driven as well as a steam driven feedwater pump, is actuated to protect the reactor core from an overheating condition upon loss of main foodwater or RCp circulation.

OTSG isolation is actuated to protect the core from an overcooling condition if a main steam line rupture should occur.

At the time of the event, OTSG Icvels were being lowered from 185 inches to 30 inches in accordance with the " Plant Startup" procedure (OP1102,02). The auxiliary feedwater pump (P-75) was in nervice, being supplied by one condensate pump, and the startup valves (CV-2623 and CV-2673) were closed and in automatic control to allow them to open and control OTSG levels when the low level limits (30 inches) were reached. The 'B' OTSG reached its low level limit first and CV-2673 began controlling level at approximately 29 inches. At 2144, when the

'A' OTSG reached its low level limit, the licensed control board operator l observed that the level continued to decrease below 30 inches. It was also observed that the 'B' OTSG level had begun to decrease. Operators then determined that P-75 discharge pressure was less than OTSG pressures. At this time, additional-operations personnel were dispatched to search for possible l

1eaks in the condensate / feedwater system. OTSG levels continued to decrease

' until, at 2150, the EFW system was automatically actuated providing feedwater to I

the OTSGs. The introduction of cold feedwater to the OTSGs caused a decrease in l pressure to approximately 850 psig, Upon observing the decreasing OTSG pressures, and in-conjunction with the other abnormal secondary system Indications, the control board operator manually tripped the Group 1 control rods in consideration of the possibility of a main steam line break. The-EFW system quickly returned the OTSGs to their programmed levels. The steam driven EFW

NFC Form 366A' U. S. Nuclear Regulatory Cmmissian (6-89) Anuwn! 018 No. 3150-0104 Expires: 4/30/92 1.lGNSIT. INihT RElWF (IIR) 'IDT U Nf1MJATI(N i FACILITY NME (1) 11nTT MMRER (2) IIR MMNR (6) PNE (3)

Sa p utial Revislan Artansas Nacimr One, Unit Die _lmt _

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'11XT (If more spice is requiral, use akiltimal NIC Fom 366A(s) (17)3}_1{3 pump (P-7A) was secured at 2200, while the motor driven pump remained in service (P-7B) supplying the OTSGs. The auxiliary feedwater pump was secured and inspected _to determine if it had incurred any damage. The inspection revealed no obvious damage to the pump. It was restarted at 2205. At 2210, after verifying that the auxiliary feedwater pump was operating normally, P-7B was secured and the EFIC system was reset. At 2237, the reactor trip was reset and plant startup was continued.

C. Root Cause The cause of this event, was determined to be leakage past the seat of FW-8A (feedwater recirculation line) which was closed at the time of this event, and -

through FW-9A, which was open, to the condensnr (see attachment). This leakage, over a period of time when OTSG feedwater was not required, resulted in the feedwater line between CV-2623 and FW-7A becoming voided. When the 'A' OTSG reached its lower limit, CV-2623 negan to open as designed, llowever, since flow was entering a partially voided line, the OTSG 1evel continued to drop and CV-2623 continued to ranp open in response to the decreasing level. As CV-2623 continued to open, P-75 discharge pressure dropped below OTSG pressure, and the

'B' OTSG level began to decrease, resulting in the 'B' startup valve (CV-2673) beginning to open. farther. It is believed that P-75 pumping into thn voided feedwater line resulted in the pump reaching a ' runout' condition .'hich rendered it incapable of supplying feedwater to t.be OTSGs.

-D. Corrective Actions FW-8A will be repaired prior to plant heatup from refueling outage IR10, which is scheduled to begin in April, 1992. FW-9A is presently closed, offectively i- isolating-the feedwater recirculation line. This line is not expected to be L used prior to restart from 1R10.

Additional corrective actions which will be taken to aid in preventing the occurrence of similar events during future plant startups include:

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• The ' Plant Startup' procedure (OP 1102.02) will be revised by July 31, 1991 i to require that FW-9A and a be closed whenever FW-8A and B are required to be closed.

+- The ' Plant Startup' procedure will also be revised to include a requirement to verify the ability to feed the OTSGs with the auxiliary feedwater pump prior to reaching the low level limits when lowering OTSC levels during startup. This revision will also be completed by July 31, 1991.

• A repetit2ve maintenance t ask has been developed for FW-8A, FW-8B, FW-9A and t

FW-9B.

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NRC Fopn 366N U. S. Nuclear Regulatory Gunissim A;ptuval GiB No. 3150-0104 (6-89)

Expires: 4/30/92 1,10NSEE IMNr idHET (IG) 11XF UNilNUATI(N F/CII.lTV NM (.1) IIGET NLMER (2) IllMM2L(6) IW;E (3)

Snjuential Revisim Arkansas Nuclear nie, Unit One int ltint Ols1010l91 al 113 MLi -nL913 . Nunder__r0l0el41aFigts au" (if nore since is rnluirni, use riklit icml NRC Fonn 366A's) (17)

E. Safety Significance The EFW system was actuated and operated as designed during this event. In addition, the reactor was subcritical at the time and no significant- RCS perturbations resulted from the event. Therefore, there was no safety significance associated with the event.

F. Basis For Reportability The automatic actuation of thn EFW system as well as the manual tripping of the Group I control rods is reportable pursuant to 10CFR 50. 7 3( a )( 2 )( iv ) .

This event was also reported in accordance with 10CFR50.72 at 2345 on April 21, 1991.

G. Additional Information There have been no previous similar events reported by ANO.

Energy Indust ry Information System (EIIS) codes are identified in the text as

[XX].

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NRC Fem 3/4 U. S. NttlNtr Rgilatory Omissiai (6-99} Al ptwai GiB No. 3150-0104

• . Expires: 4/30/92 )

L10NSEE E\1NT REIET (UR) 11XT (INTLNUATICN F/CILlw NAME (1) DIHT NllWR f 2) UI NUMBER (6) PNE (3)

Sorpential Revisim Attansas Ntelnar Om, Unit Om .

Year Nurfer ld.rg._

01.5[0{gjQj 3} _1}_3 .,9l 1 r

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ATTACitMENT CV-26SO FW-7A N 1

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.3, STEAM FE-2627 GENERATOR F\W6A CV-2623 FW-6A

, s E-1 A HEATER NFW -9A TO THE CONDENSER FW-19 FROM AUX. FWP

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