05000423/LER-1993-004

LER 93-004-01:on 930331,RT Occurred Due to electro-hydraulic Control Power Supply Failure.Replaced Faulty Power Supply. W/930831 Ltr

ML20056H501
Person / Time
Site:	Millstone
Issue date:	08/31/1993
From:	Scace S, Young J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-PT21-93 LER-93-004-01, LER-93-4-1, MP-93-687, NUDOCS 9309090435
Download: ML20056H501 (5)

LER-2093-004,

Event date:
Report date:
4232093004R00 - NRC Website

text

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i EORTHEAST UTILITIES a = S = ces s a en street Beran connecocut 1 I*cif NaY,$ dtN EIC ah

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' rwineast Noea Erew comran, HARTFORD. CONNECTICUT 06141-0270 (203)665-5000 August 31, 1993 MP-93-687 Re: 10CFR50.73('a)(2)(iv) 1 l U.S. Nuclear Reculatorv Commission Document Control Desk l Washington, D.C. 20555 1

Reference:

Facility Operating License No. NPF-49 Docket No. 50-423 Licensee Event Report 93-004-01 Gentlemen:

1 This letter forwards Licensee Event Report 93-004-01 which is being submitted to report on root cause and revise the analysis of event. Licensee Event Report 93-004-00 was l

submitted pursuant to 10CFR50.73(a)(2)(iv). any event or condition that resulted in l

manual or automatic actuation of any Engineered Safety Feature (ESF), including the l

Reactor Protection System (RPS).

I Very truly yours, NORTHEAST NUCLEAR ENERGY COh1PAhT ephen .

Vice President - hiillstone Station SES/JSY:ljs

Attachment:

LER 93-004-01 cc: T. T. h1artin, Region 1 Administrator P. D. Swetland, Senior Resident inspector, hiillstone Unit Nos.1, 2 and 3 )

V. L Rooney, NRC Project hianager, hiillstone Unit No. 3 l

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L.CENSEE CONTACT FOR THIS LER (12) )

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1 Jettrey S. Young. Engineer. 6442 (203) 447-1791 COMPLETE ONE LINE FOR E ACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

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SUPPLEMENTAL REPORT EXPECTED (14) E xPE CTED YES SUBM:ESION No NE W f4 ses com.e'e E ADE CTED SJBM SSON L ATE. X ABSTRACT um.t tc 60; antes . e. , an ycea:e'.15 s mo e-u.a:*c tgw mea res)06)

At 0103 on March 31. 1943. wnh the plant in Mode i at 1009 pouer, a turbine vahe closure resulted in a )'

reactor inn fol! owed by a turbme inp.

Turbme vahe closure was the result of a faut > power supply m the Electo-Hydraunc Control (EHC) system.

With the excepuon of a (team generator code safety sahe not completely reseatmc the plant responded normall> to the tr2nuent Extenme uou& shooung determmed that a power supply in the EHC system was fauhy arid caused the turbine sahes to close. The faulty power supph was replaced. At action to present recurrence, the power supplies m the EHC system mll be replaced or refurbnhed on a 10 year penod.

Subsequent mvesugation determmed that the steam generator safety vahe which did not completely re> eat had i an mcorrect lower adjustrnent rmg setung. Addiuonal mtpect on revealed that other safety sahes also had i mcorrect setungs Three of these vahes mdnated that they hited and reseated durmg the transient. The other four vahes d:d not open.

The root cause of the improper setungs was madequate work control b3 the sendor Crosby Yahe and Gage Compans, who performed rnamtenance on the saieues. conducted an mvesupuon and determmed that their personnel madvertent!) used the wrong procedure to set the lower adrustment rmgs.

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On March 31,1993 at 0103 wnh the plant m Mode 1 at 1009 pouer (M6 degrees Fahrenhen and 2250 psta) a turbme load teJection alarm was recened as the turbine control vahes began to close. A normal response to turbme valve closure occurred. This conusted of all steam dumps opemng, atmosphenc durnps on 3 steam generators opening, and 1 Pressunzer Power Operated Rehef Yahe (PORY) cythng ,

open for 6 seconds. In add: tion. mdicauon was recen ed that several code safet) vahes on 2 steam generators hhed. 19 seconds aber the turbme vahes began to close, steam generator levels shrank to the low low setpomt which resulted m a reactor tnp followed by a turbme tnp. After the reactor and turbme inp, the atmosphenc dumps on the fourth steam generator opened and mdicahon was recened that code safeues on a 'htrd steam generator hfted. At steam pressure decreased. all dumps and safenes rescated with the excepuon of I safety on the "D" steam generator.

At the ume of the inp. operaton venhed that the Reactor Tnp and Bypass Breakers were open, that ali i

control roos uere fvlh inserted. and that neutron flux was decreasmc. An automatic start of the Auuhan f eeduater System occuned due to the low low water lesel'm one steam generator and the  ;

turbme drnen Auxihart Feedwater Pump staned on low low water level in two steam generators. A Fteawater holanon occurred due to the lov average Reactor Coolant System temperature (564 degrees Fatuenheiy comtidem with the reactor tnp Subsequentiv, the plant stabihzed at the no load temperatore of 557 ceprees Fahrenheit. Thece were expected sptem responses. No addinonal Engmeered Safet3 Feature (ESF) actuanons were required or irnuated.

Aher the plant had stabihred, the salen was gagged but still did not close completeh' A normal plant cooldown was performed and the safety reseated at approximately 1000 psig steam generator pressure, Subsequent irnesuganon resealed that the louer adjustment rmg of the malfunctiomng safety vahe was i mc ot r ect. When set properly, the louer nng insures a clean forceful popping action when the safety lifts ,

and a cushioned rescating when the safety closes. If this nng is set 100 h:ph, steam cannot escape as i rapidh> u hen the vahe bepms to close. As a result the blowdown is mereased. The settings of the lower adlustment rmps on 11 other safeties which had been worked on by Crosby Vahe were checked. All 7 of the vahes which were mstalled were found to be set mtproperly. Of these, 3 mdicated that the3 had hfted and restated dunng the transient and 4 did not open because their setpomts were not reached.

11. Caute of Esent I l

The root cau*e of the turhme vahe closure was equipment failure. A capacnor failure m one of the EHC power rupphes caused noise to EHC solenoids resulting m a slow closure of all turbme vahes with servos.

The root cause of the safety vahe fathng to completeh reseat was madequate work control by the l t endor. When the contractor performed mamtenance on 4 safeties in 19S7 and 19%, NNECO Quahty I Control perforrned an mtpecuon of the uork on 2 that required ring adjustment. No discrepancies were noted and the nng setungs of these safenes were correct. Work on another group of b safenes was performed between 19E9 and 1991. Dunng this penod, the ring settings were documented and a certihcanon statement was issued. Investiganon by the vendor after the event revealed that a procedure +

for Crosby salen salves was inachenently u'ed m setting the lower adjustment rmg instead of the procedure specihed b) Nonheast Etihues. Northeast Cuhties performed a surveillance of Crotby and found no d:screpancies. An mdustry aud t team who had previously scheduled an audn of Crosb3's l Append:x B program was informed of this event and the results of the surveillrnce. This auda team had i no Upmhcam Endmgs and reported only a mmor order entry discrepanc3 Ill. Analvm ollvent Nac wmanA wo

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This esent is bemp reponed m accordance with 10CFR50.73faH2Hn9 as any esent or condinon that resulted m automauc actuauon of an ESF meludme the Reactor Protecuon System. An immediate nonficauon was made m accordance with 10CFR50.72(bH2Hii).  ;

All safet3 systems funcuoned as designed as a result of the reactor inp. The Auxthary Feedy ater System started automancally due to the low low steam generator water level. A Feedwater Isolauen occurred due to the low Reactor Coolant System aserage temperature comcident urh the reactor inp. No other 6 E5F signals were initiated and the esent posed no signif cant hazard to the health and safety of the public. Sc ondary plant equipment was returned to normal operation. and the unn was returned to pow er  ;

The improper setung of the louer adiustment rings on the safety vahes did not place the plant out side of the design basis for a stuck open safen vahe. The plant is analyzed for a flow of 97'.200 lbm'hr for this accidem Each safety vahe is destgned to pass f r~r (970.000 lbm'hr) of total steam flow. How ever, parameter mdicauens shoued that the affected safet3 vahe had not completely reseated but was not  !

stuck fully open The partially open safety did not result m an uncontrolled cooldown and the plant  ;

operators were able to mamtam normal shutdown plant temperature and pressure. Indicated proper operanon of 3 other safeues which had improper nng adjustment durmg the transient showed that ,

~

improper setimg of the lower adiustment nng does not necessarily mean that a safety will not reseat.

4 While actual operanon and subsequent testmg of the malfunctionmp safety (see below) showed that n re'eated at lower pressures. these pressures are normall3 reached several hours aher p! ant shutdown. l Theref ore, arn. other vahe which did not fully reseat would eventualh. close by uself.

l W yle Valve Reteated Reseat Pressure {

Test No. Durmg Test (PSIG) 1 No < 1055 a 2 Yes 1068 i I

3 Yes 1062 4 No < 1065 I Therefore. the 4 vahes that did not open may hase rescated properly if they had opened. In addition. l I

they would not have stuck fully open if they had failed to reseat and would hase closed completely by

t hemseh e3 after a controlled plant cooldown to approximatelv 530 decrees Fahre1 hen i

Because there is no measurable pnmarv to secondary leakace m the "D" steam generator, there was no .,

release of radioactne matenal as a result of the partially open safety vahe. l l

A review of the impact of an extended blowdown of a main steam safety vahe on other postulated accidents has been completed. This review concluded that the consequences o.f a main steam safety failmg to reseat properly are bounded by the assumptions and results of the accident analysis for ifnit 3 This completes review of this erem as a 10CFR21 concern.

IV. Correctn e Action  ;

As immediate correcure action for the tr p, the fauhy power supply was replaced. As action to prevent  !

recurrence. the power supply will be replaced or refurcished on a 10 year period. Due to replacement j unavailabihty. the other power supplies m the EHC system will be replaced durmg the upcoming refuelmg j outa ge , i Nec s o-m awA 15-923 4

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At ammed: ate correctne action for the 'mproper ring setungs, the 11 other safet3 vahes which had been r refurbnhed. repaired or set by Crosby \ ahe were examined for proper rmg setungs. The 7 (excludmg  !

the ma!!uncuonmp onel which were improperly set were adjusted as necessary to their proper nng -

setungs All spare safeues currentl3 out for mamtenance at any contractor will be serihed before return to the ute.

i V. Minonn! inf r rrv ati;/n No other Licensee Event Reports (LER31 hase been submitted for reactor inps resulung from a f atied  ;

power supph in use EHC synem. [

Tlus LER also satishes NNECO's evaluation. nou6 cation and reporung obhrauon to report defects under 10CFR21. In addition, a copy 0.1 this LER was sent to the contractor that performed mamienance on l ,

the mam steam safeues. I i No sirmlar pouer supply malfuncuons were found m a renew of NPRDS. t Ells codn hm Egg,nnyn, Mam Turbine Control power Supply, Electric - JX I

Flmd 53 stem - TG hn' Reheat Steam Rehe! Vahe - RV l Sutem - SB  !

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