ML20045F688

LER 93-004-02:on 930222,reactor Trips on SG Low Water Level Occurred.Caused by Automatic Trip of MFW Pump B on Low Suction Pressure Due to High FW Flow Rates.Classroom Training on Main & Afwc Received by Licensed Operators

ML20045F688
Person / Time
Site:	Millstone
Issue date:	07/02/1993
From:	Binkowski M NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20045F687	List: ML20045F686 ML20045F688
References
LER-93-004, LER-93-4, NUDOCS 9307080246
Download: ML20045F688 (5)

Text

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NHC Fom 366 US NJCLE AR RE GULATORY OOMM;SSOfs A M oVE D OME NO 3 bO-01Ni (6-891 EXPIRE 5. 4.30/92 E stmatec nuoen .w. rewcme 1c compiy w'th 1+s mto mation c@ection reovest 50 0 ms F ovato Co9mDnts reQarom@ Di. Foe *. est Aate lo lhe ROCo'o5 LICENSEE EVENT REPORT (LER) a,c Renons Management B.anen to-53en c s osciear Repd. story Comm$stos Wasnmgtom DC 2255 anc to

!' the Faperwo-w Recat~ tion P o ett (3150-0104; C4ce c4 Va*egement enc Buoget A W .nctor.. DC 20503 F ACl#v NAME n) DOCPU NJMEER (L "'-

J Millstone Nuclear Power Station Umt 2 oj si ol ol oja [3 is 1lod0l5 i T J LE 143 i

, Reactor Tnps on Steam Generator Low Water Level l l

l EVENT O ATE (5) LEm NJVBEA iEi REPOATDATF m OTWE A F ACh mE S PNO;VED .8 t

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1 20 405.a' fit M SG r3 a);2)hW 50 73iai f2!m f UCENSEE CONTACT FOR TH:S LER f T2s l NAVE TE EPHONE NJMBER f a ma A cOpi Michael Bir$owski. Mechanical Engmeer, Ext. 6845

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i On February 22,1993, at 0144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br />, with the plant in Mode 1 at 1009 power, the " A" steam generator ,

atmosphenc steam dump vahe ( ADV) failed open. Dunng the subsequent transient plant conditions, the "E" '

mam feeduater pump automatically tnpped on low suction pressure, and an automatic reactor trip on low steam j generator water level occurred at 0151 hours0.00175 days <br />0.0419 hours <br />2.496693e-4 weeks <br />5.74555e-5 months <br />. Operators then performed Emergency Operatmg Procedure i j EOP 2525. " Standard Post Trip Actions " All safety related eqwpment responded as expected and the unit was placed m a stable condiuon. The cause of the reactor tnp was the mabihty to recoser the water Inel m the

" A" steam generator followmg the inp of the "B" main feedwater pump.

J i

Dunng the subsequent plant startup on Februar) 23,1993, at 2037 hours0.0236 days <br />0.566 hours <br />0.00337 weeks <br />7.750785e-4 months <br />, with the plant m Mode 1 at 159 power, high vibranon of teveral mam turbme bearmgs required shutdown of the mam turbme. Reactor power j

, was quickl> reduced to apprownately 149 power and the mam turbme was tripped at 2040 hours0.0236 days <br />0.567 hours <br />0.00337 weeks <br />7.7622e-4 months <br />. During the j i subsequent transient plant conditions, an automanc reactor inp on low steam generator water lesel occurred at l l 2043 hours0.0236 days <br />0.568 hours <br />0.00338 weeks <br />7.773615e-4 months <br />. Operators then performed Emergency Operating Procedure EOP 2525. " Standard Post Trip '

j Actions? Ali safety related equ pment responded as expected and the urut was placed in a stF condition.

The cause of the reactor tnp was insuf ficient feedwater flow to the steam generators for the exe ,

reactor power lesel.

]

These events are being reported pursuant to the requirements of Paragraph 50.73(ay2)fiv), reportmg any event or condition that resuhed m manual or automatic actutuon of any Engineered Safet) Feature System.

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OF 0l 5l 0l 0l 0l3 l3 l6 9l3 0 l 0l 4 0l2 0l 2 0[5 f TEC # mee s; ace is veauced use aca tiota NRO Form 3t6 A s f 07;

1. Decenpnon of Event j On February 22,1993, at 0144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br />, with the plant in Mode 1 at 1009 power, the " A" steam generator atmospheric steam dump valve (ADV) failed open. The ADV failure caused the indicated steam flow rate to decrease and the feedwater lesel control system to automatically decrease feedwater flow to the " A" steam generator. thus decreasmg the steam generator water level. The operatms then took manual control of the " A" steam generator feedwater control valve m an effort to restore steam generator water lesel. Dunng the subsequent transient, the " A" steam generator water level increased to the high level setpoint, causing the feedwater control valve to automancall> close and the water level to decrease. After the " A" steam generator feedwater control valve was reopened, excessive f eedwater flow rates caused the "B" mam feedwater pump to automaucally tnp on low suction pressure, resuhmg m an automatic reactor inp on low steam generator water level at 0151 hours0.00175 days <br />0.0419 hours <br />2.496693e-4 weeks <br />5.74555e-5 months <br />. Operators then performed Emergency Operatmg Procedure EOP 2525. " Standard Post Trip Actions? Auxihary feedwater automaucallj m:aaled with no comphcations. All safety related equ pment responded as expected and the umt was placed in a stable condition. ,

Dunng the subsequent plat startup, on February 23,1993, at 2037 hours0.0236 days <br />0.566 hours <br />0.00337 weeks <br />7.750785e-4 months <br />, with the plant in Mode 1 at 159 power. h)gh vibration of several mam turbme bearings required shutdown of the main turbme within l 15 mmutes. Reactor power was quickly redned to approximately 149 power and the mam turbme was l inpped at 2040 hours0.0236 days <br />0.567 hours <br />0.00337 weeks <br />7.7622e-4 months <br />. Following the mam turbme trip, the steam generator pressures increased. causme the differenual pressure between the main feedwater pump and the steam generators to decrease, and consequently decreaung the feedwater flow rate to the steam generators. Addiuonalh, the operators i were minally concerned about oserfeedmg the steam generators, and stopped the automatic actions of the ,

feeow. iter control system and consequently only partially opened the feedwater bypass control vahes. I When ream generator levels were observed to be decreasmg. the feedwater flow rates were then j mereasec. Due to differences m the feedwater bypass valve positions between the two steam generators. -

more feedwater flow was directed to the B" steam generator and the " A" steam generator water level dd not recover prior to reaching the low lesel inp setpomt. An automatic reactor inp on low steam generator water level occuned at 2043 hours0.0236 days <br />0.568 hours <br />0.00338 weeks <br />7.773615e-4 months <br />. Operators then performed Emergency Operating l Procedure EOP 2525, " Standard Post Tnp Acuons " All safety related equipment responded as

! expected and the umt was placed m a stable condition.

Causm of Evem

, ae root cause of the automatic reactor inp on low steam generator water lesel on February 22. 1993, I was the automatic trip of the "B" mam feed rater pump on low sucuon pressure due to the high I feedwater flow rates beme demanded to recover steam generator water levels.

l The cause of the imuatmg event on February 22. 1993, was the " A" steam generator ADV f ulmg open.

The ADV f ailed open when the sprmg attachment fitting broke at the point where the spring attaches to

! the mner diaphragm asse nbly. The break at the attachment point resulted from a combination of l rmsalignment of the feedback spnng and bending of the range spnng attachmert (possibly durmg installation). and positioner vibrauon dunng normal plant operanon. See Figure 1 for a diagram of the AE.V valve poutioner.

The root cause of the automatic reactor inp on low steam generator water lesel on February 23. 1993, was msufficient feedwater flow to the steam generators for the existmg reactor power level. Ses eral i contributing factors were mvolved in this root cause:

i Following the main turbme trip, steam generator pressures increased, causing the differential pressure between the main feedwater pump and the steam generators to decrease, and consequently decreasing the feedwater flow rate to the steam generators.

I NAC F o-m 366

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i NPc emm mA u.s Nue2AR RE3xArony ceuwssoN APPRCVED OMS NC 31sD-CiD4 t f-FE E XPiREs : 4'33 92 E stmatec b.rcen per response to compiy we this m'"'a t

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LICENSEE EVENT REPORT (LER) comments' eparomo"'taaen"estmate to the Recor:'s TEXT CONTINUATION y,c necom vanage,ent e anen m_sx; u s Nsciea, Repsiators com%ston Wasnmpton De POEM anc to the Pape wow AecacSon E~o:ert i3'6D-e D4) o".ce o' Vanapeme'd 6'C Budget V6 EssT om DC EsL3 E AClUTY NAP /E (t DOCKET NJMBER (2) Lemturmcm g; F AGE 01 YEAR SN N mem em

\hlist,one Nuclear Power Station y '- , - -

0l 0j 4 0l2 0l 3 OF ol 5l 0l cl 0l3 l3 l6 9l3 0l5 TEXT of more space :s reosaea ase ac:mona' NRc Fo m 366A si 07)

• Subsequent to a turbme tnp signal, the main feedwater regulating vahes automatically close, and the feedwater regulatmg bypass valves automaucally open to approximately -'5"c. The operators were minally concerned about overfeedmg the steam generators, and stopped the automauc actions of the feedwater control system and consequently only parually opened the feedwater regulaung bmass valves. When steam generator levels uere observed to be decreasmg. the feedwater flow rates were then mcreased. Due to differences in the feedwater bypass vahe posiuons between the two steam generators, more feudwater flow was directed to the "B" steam cenerator and the " A" steam generator water lesel did not recoser prior to reaching the low level inp setpoint.

! A contribunng cause to both events was an unfamilianty by the operators to feedwater transients involving the new steam generators. Dunng licensed operator training on the new steam generators, the operators ,

l were informed that the transient water level response of the new steam generators uould be more stable

! when compared to the origmal steam generators. This statement is correct for water lesels greater than 50cc , but is mcorrect for water levels less than 50cc , where the level changes are much more rapid and more indicauve of the response obsened on the ongmal steam generators.

l The cause of the high main turbme bearmg vibration ispears to be a turbine " rub." The turbine was placed on the turning gear and was restarted on February 24, 1993 with no problems.

111. Anahmic of Es ent These events are bem; reported pursuant to the requirements of Paragraph 50.'3(aH2)(iv). reporting any esent or condiuon that resuhed in manual or automatic actueuon of any Engmeered Safety Feature Sy stem.

There were no safety consequences from these reactor tnp events. All safety related equipment responded as expected and plant operators executed apphcable Emergency Operaung Procedures accordmgly.

IV. Corrective Action The vahe positioner for the "A" steam generator ADV was replaced and the feedback spring was reabgned. The vahe posmoner for the ~B" steam generator ADV was replaced. and the as found i feedback spnny ahpnment was sausfactory.

Operator shih bnefmps vere conducted to provide mformanon on the obsened steam generator lesel response dunng the transient conditions. The water level response for the new steam generators, which were installed durmg the previous refuehng outage. is as follows:

For indicated steam generator water levels greater than 509 (moisture separator region), the indicated lesel changes are slower when compared to the original steam generators.

For indicated steam generator water levels less than 50"c (downcomer region), the indicmed level changes are the same as the original ste: generators.

For mdicated steam generator water levelt in the downcomer region (less than 50Fc ), the mdicated lesel changes are sigmficantly faster as compared to changes in the mosture separator region (greater than 50cc ).

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NRO Fvm 360A U.S NUCLE AR REGULATORY COMM 5SiON APPROVED OME NO. 3150-C104 (6-89) EXPAEs 4 0012 Est. mated buroen per response to como'y w'th tNs

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LICENSEE EVENT REPORT (LER) Commer'ts regarding"DurDen estmate to the ReOoros TEXT CONTINUATION ano neoorts unnagement sranen (p-530). U. S Nuc taar Aegulato y Comemssion. Wasmngtoa DC 20555. ana to

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Ma - gement

".aperwork anoAeduction Buopet. Pro,eet (3150-01N Wasmngton DC 20603 L Off ce of F AOILITY NAME 11 DCOKET NUMBER (2) t FA NUMP:A@ PA GE (34 YEAR "#" N

- umm hitllstone Nuclear Power Station Unit 2 - ~

OF 0l 5l 0l 0l 0l3 l3 l6 9l3 0l 0l 4 0l2 0l 4 0l5 l TEXT Uf more scate is reamre: use aodrttonat NRc Forre 366A s) (17) 1 Additional actions implemented to prevent recurrence were as follows:

a) The ADV Positioners have been added onto the Production Alaintenance hianagement System ,

(PA1515) and will be replaced esery outage. I b) All licensed operators have recened specific classroom traimng on the construction of the new steam generators with emphasis placed on how the differences between the old and new steam generators will affect lesel response.

i in addition, specibe classroom traming on Alain and Auxihary Feedwater Controls emphasizmg steam generator response has been received by all licensed operators and associated simulator trammg has been received by all except one licensed operator. The simulator model has been upgraded to represent the changes in level control which resulted from the replacement of the steam generators. Normal / routine simulator training will further enhance the operators' response to level l transients. l The final operator is scheduled to receive the specific simulator traming during the month of July. l The combination of the classroom and simulator training will provide the operators with the I ,

necessary knowledge to accurately control level in the new steam generators.

V. Additional Information Similar LERS: f*7-12, 87-11, 57-09, 87-02 I i

Ells Codes for referenced components: '

Atmospheric Steam Dump Valve: SB-PCV-C635 Feedwater Control Valve: SJ-FCV-C635

• Alain Feedwater Pump: SJ-P-1075  !

Alain Turbme: TA-TG-GOS4 The following component failed during this event:

Atmosrheric Steam Dumo Valve Pocitioner Manufacturer: Moore Industries Model: 12372-74GS10GC Ells Code: SB-00S4-M4:2 1

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NRC Form 366A U. S. NUCLEAR REGULATORY COMMISSION ADPROVED OMB NO. 3150-0104

. (6-89) EXPIRE S: 4 '30'G2 Estimated burden per response to comply witn this

' = '*n o"*cton vemest ; 50.0 hrs. Forward .

LICENSEE EVENT REPORT (LER) c'"omments re"garomy butoen estimate to the Records

. TEXT CONTINUATION ane neoorts Management sranen <p-sact u.s Nuc+ ear-Regatatory Commission. Washington, DC 20555. and to t e Paperwo a Reducton P-o.!ect (315D-01C4 i Office of Management and Badget. Wasningtort DC 20503 FACILITY NAME (11 DOCKET NUMBER (2) LEA NUMBpo e PAGE131 YEAR N N w empi

( l Millstone Nuclear Power Station Unit

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DETECTOR N  !

NOZZLE \ l h t (T T)  :

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t VALVE 1 l r' . I I PLUNGER 1 PRESSURE fr a t 1 ,9-INPUT BEAM ""

PILOT VALVE  :

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ggppLy l PLUNGER 2 l RANGE SPRING - ll 6i (F;i li, S SPAN "U i ADJUSTMENT VALVE 2 l

PRESSURE ZERO OUTPUT PRESSURE ADJUSTMENT LEVEL ADJUSTMENT Figure 1 Atmospheric Steam Dump Valve Positioner i

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