ML18036A727

LER 92-004-00:on 920427,reactor Scrammed on Low Reactor Water Level.Caused by Failure of Feedwater Control Sys Resulting in Feedwater Pump run-back & Low Flow to Reactor Vessel.Maint Planning Guide revised.W/920526 Ltr

ML18036A727
Person / Time
Site:	Browns Ferry
Issue date:	05/26/1992
From:	Hsieh C, Zeringue O TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-004-02, LER-92-4-2, NUDOCS 9206010233
Download: ML18036A727 (20)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9206010233 DOC.DATE: 92/05/26 NOTARIZED: NO DOCKET FACIL:50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 AUTH'AME AUTHOR AFFILIATION HSIEH,C.S. Tennessee Valley Authority ZERINGUE,O.J.'ennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 92-004-00:on 920427,reactor scrammed on low reactor water level. Caused by failure of feedwater control sys resulting in feedwater pump run-back & low flow to reactor vessel.Maint planning guide revised.W/920526 ltr.

DISTRIBUTION CODE IE22T COPIES RECEIVED LTR ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event Report.(LER), Incident Rpt, etc.

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL SANDERS,M. 1 1 HEBDON,F 1 1 ROSS,T. 1 1 INTERNAL: ACNW 2 2 ACRS 2' AEOD/DOA 1 1 AEOD/DS P/TPAB 1 1 AEOD/ROAB/DS P 2 2 NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFB10 1 1 NRR/DLPQ/LPEB10 '1 1 NRR/DOEA/OEAB 1 1 NRR/DREP/PRPB11 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1 NRR/DS~QLB8 Dl G~~

RGN2 FILE 01 02 1

1 1

1 1

1 NRR/DST/SRXB 8E RES/DSIR/EIB 1

1 1

1 EXTERNAL: EG&G BRYCE,J.H 3 3 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHY,G.A 1 1 NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAIWIE FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 32 ENCL 32

, I Tennessee Valley Authority, Post Office Box 2000. Decatur, Alabama 35609 O. J. 'Ike'eringue Vice President, Browns Ferry Operations NAY P6 ]ggg U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

Dear Sir:

TVA BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 2 DOCKET NO. 50-260 FACILITY OPERATING LICENSE DPR-52 LICENSEE EVENT REPORT LER-50-260/92004 The enclosed report provides details concerning the Unit 2 reactor scram on low reactor water level. The low water level was caused by the failure of the feedwater level control system. The failure of the level control system resulted in feedwater pumps run back to a low flow condition.

The report is submitted in accordance with 10 CFR 50.73(a)(2)(iv).

Sincerely, 0 J. Zeringue Enclosure cc: see page 2 fthm'y~

9206010233 920526 PDR ADOCK 05000260

(/j/ I 8 PDR

I 1

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U.S. Nuclear Regulatory Commission Yi4Y 26 1992 cc (Enclosure):

INPO Records Center Suite 1500 1100 Circle,75 Parkway Atlanta, Georgia 30339 Paul Krippner American Nuclear Insurers The Exchange, Suite 245 270 Farmington Avenue Farmington, Connecticut 06032 NRC Resident Inspector Browns Ferry 'Nuclear Plant Route 12, P.O. Box 637 Athens, Alabama 35609-2000 Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 2900 Atlanta, Georgia 30323 Thierry M. Ross U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852

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NRC Form 366 Ll. LEAR REGULATORY COMMISSION Approved OHB No. 3150-0104 (6-89) Expires 4/30/92 LICENSEE EVENT REPORT (LER)

FACILITY NAHE (1) IDOCKET NUHBER (2)

(4) Automatic Reactor Scram on 1'ITLE Low Reactor Water Level Due to Failure of the Feedwater Level Control I . I I I I SEQUENTIAL I I REVIS ION I I I I FACILITY NAHES IDOCKET NUHBER(S)

I I I I I I I I 05 I 26 I 92 I OPERATING . I ITHIS REPORT IS SUBHITTED PURSUANT TO THE REQUIREHENTS OF 10 CFR g:

NODE F20.402(b) (20.405(c) ]50.73(a)(2)(iv) I73 71(b)

POWER J f )20.405(a)(1)(i) )50.36(c)( 1) 150.73(a)(2)(v) l73 71(c)

LEVEL / f /20.405(a)( l)(ii) )50.36(c)(2) )50.73(a)(2)(vii) )OTHER (Specify in

[20.405(a)(l)(iii) J

)50.73(a)(2)(i) ]50.73(a)(2)(viii)(A) Abstract below and in (20.405(a)(l)(iv) f (50.73(a)(2)(ii) (50.73(a)(2)(viii)(B) Text, NRC Form 366A)

.4 1 v ~ 111 NAHE I AREA CODE I AH HP T N T R T I I IREPORTABLEI I I I IREPORTABLEI H HP N T I I I I I I I I P T 14 I I SUBHISSION I I I N 0 T ABSTRACT (Limit*to 1400 spaces, i.e., approximately fifteen single-space typewritten lines) (16)

On April 27, 1992, at approximately 1432 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.44876e-4 months <br />, Unit 2 reactor scrammed on low reactor water level. Engineered safety feature actuations included primary containment isolation system Groups 2, 3, 6,"and 8, and actuation of the control room emergency ventilation and standby gas treatment (all trains) as expected.

The low reactor water level was due to the failure of the feedwater level control system. This resulted in the feedwater pumps run back and low flow to the reactor vessel. The operator was unable to regain control of the feed pumps before the reactor scrammed on low water level.

The cause of this event was the ~aster feedwater level controller output failed downscale. The downscale failure was due to an unexpected and random failure of an electrolytic capacitor in the controller.

The following corrective actions have been or will be taken to address this event:

1) The maintenance planning guide has been revised to include testing wet type electrolytic capacitors, 2) TVA will evaluate the power stores procedure on monitoring shelf life of high-risk components in storage, and 3) TVA will evaluate implementation of the Scram Frequency Reduction Cormnittee recormnendation to design and install a fault tolerant digital feedwater control system.

NRC Form 366(6-89)

Ol NRC Form 366A U.S. CLEAR REGULATORY COHHISSION Approved OHB No. 3150-0104 (6-89) Expires 4/30/92 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAHE (1) IOOCKET NUHBER (2)

I J SEqUENTIAL / / REVISION J Browns Ferry Unit 2 I I I I I TEXT (If more space is required, use additional NRC Form 366A's) ( 17)

I. PLANT CONDITIONS Unit 2 was operating in the run mode at approximately 100 percent power (3288 MW Thermal). The Feedwater Level Control (FWLC) system was in automatic three-element control and was controlling reactor vessel water level at approximately +33 inches.

Units 1 and 3 were shutdown and defueled.

II. DESCRIPTION OF EVENT A. gvggt:

On April 27, 1992, at approximately 1432 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.44876e-4 months <br /> 21 seconds, the Unit 2 reactor scrammed on low reactor water level (+ll inches), resulting in the actuations of the Engineered Safety feature (ESF) [JE] systems. The ESF actuations included Primary Containment Isolation System (PCIS) [JM]

Groups 2, 3, 6, and 8 on low water level, "and actuation of the Control Room Emergency Ventilation [VI] and Standby Gas Treatment [BH] systems (all trains) as expected.

The low reactor water level was due to the failure of the FWLC system.

(The level controller output signal dropped from 100 percent to 20 percent.) This failure in turn caused all three feedwater pumps to run back to a low flow condition which resulted in a reduced makeup feedwater flow to the reactor vessel. The reactor water 1'evel dropped rapidly as flow was reduced, resulting in a recirculation runback initiated by water level below 27 inches and a feedwater low flow condition. The lead Unit Operator (UO) (utility, licensed), noting the Reactor Feed Pump Turbine (RFPT) A, B, and C abnormal alarms and the decreasing water level, placed the master feedwater level controller into the manual mode to increase the flow demand signal. However, the operator was unable to regain control of the feed pumps before the reactor tripped on low reactor water level. A manual scram signal was immediately inserted into the Reactor Protection System (RPS) [JC].

NRC Form 366(6-89)

0 II NRC Form 366A U.S. UCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104 (6-89) Expires 4/30/92 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) IOOCKET NUMBER (2)

I I I I SEQUENTIAL I [REVISIONI I I I I Browns Ferry Uni t 2 I I I I I 4 F TEXT (If more space is required, use additional NRC Form 366A's) (17)

At approximately 1432 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.44876e-4 months <br /> 35 seconds, the reactor water level stabilized at -8.5 inches and started to increase. Eighteen seconds later the level had increased above +11 inches. During this time, both condensate booster pump 2A and RFP 2A tripped on low net positive suction head. The remaining feedwater pumps and the main turbine tripped on high reactor water level (+54 inches), and a manual turbine trip was initiated almost simultaneously with the high reactor level trip signal. All turbines tripped as expected', and all heater strings (high pressure and low pressure) isolated during the, water level transient. (The A string heaters were later returned to servi.ce.) After all feedwater pumps had tripped, their individual controllers were moved to the manual position.

At approximately 1435 hours0.0166 days <br />0.399 hours <br />0.00237 weeks <br />5.460175e-4 months <br />, the reactor scram was reset and the control rods were verified to be fully inserted. At 1436 hours0.0166 days <br />0.399 hours <br />0.00237 weeks <br />5.46398e-4 months <br />, the PCIS was reset and RFP 2C was returned to service to control reactor water level.

The reactor was brought to a shutdown condition in accordance with TVA's emergency operating instruction and maintained in a hot condition per normal operating procedures.

As a result of the ESF actuations, including the automatic actuation of the RPS, TVA reports this event in accordance with 10 CFR 50.73(a)(2)(iv) as an event or condition that resulted in manual or automatic actuation of any ESF.

B. t t th

@ggt:

The master controller output in the FWLC system failed downscale.

Troubleshooting the controller circuitry, TVA found a failed electrolytic capacitor (General Electric; C-29, 250 microfarad, Part No. 2098K42-002).

NRC Form 366(6-89)

!t NRC Form 366A U.S. NUCLEAR REGULATORY COHHISSION Approved OHB No. 3150-0104 (6-89) Expires 4/30/92 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAHE (1) IDOCKET NUHBER (

I I I I SEQUENTIAL I I REV IS ION I I I I I Browns 'Ferry Uni t 2 I I I I I TEXT (If more space is required, use additional NRC Form 366A's) (17)

C April 27, 1992 at 1432 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.44876e-4 months <br /> Reactor scram on low water level 21 seconds (+11 inches) 1432 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.44876e-4 months <br /> Reactor water level stabilized 35 seconds (-8.5 inches) 1432 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.44876e-4 months <br /> Reactor water level increased 53 seconds above low level trip (+11 inches)

, 1433 hours0.0166 days <br />0.398 hours <br />0.00237 weeks <br />5.452565e-4 months <br /> Main turbine trip on high reactor water level (+54 inches) 1435 hours0.0166 days <br />0.399 hours <br />0.00237 weeks <br />5.460175e-4 months <br /> Reset reactor scram

'1436 hours0.0166 days <br />0.399 hours <br />0.00237 weeks <br />5.46398e-4 months <br /> RFP 2C returned to control reactor water level and reset PCIS 1745 hours0.0202 days <br />0.485 hours <br />0.00289 weeks <br />6.639725e-4 months <br /> Four-hour non-emergency report to NRC as required by 10 CFR 50.73(a)(2)(iv)

D th None E. th fD This event was immediately known to the UO upon receiving indications of the reactor scram and the ESF actuations.

F.

Operations personnel promptly responded to the scram and immediately proceeded to stabilize reactor water level transient.

NRC Form 366(6-89)

S NRC Form 366A U.S. CLEAR REGULATORY COHHISSION Approved OHB No. 3150-0104 (6-89) Expires 4/30/92 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAHE (1) lOOCKET NUHBER '(2) N I I I SEQUENTIAL I IREVISIONI I I I I Browns Ferry Uni t 2 I ~

I I I I 4

TEXT (If more space is required, use additional NRC Form 366A's) (17)

G t t ESF actuations occu'rred as designed on low reactor water level scram.

These actuations included PCIS isolation in Group 2 (residual heat, removal), Group 3 (reactor water- cleanup) Group 6 (reactor building ventilation and primary containment purge and venting), and Group 8 (reactor low level). Control room emergency ventilation and standby gas treatment Trains A, B, and C started as expected.

III. CMSE OF THE EVENT The immediate cause of the scram was reactor water level decreasing below the low level setpoint (+11 inches).

B. QmLttgggg:

The cause of this event was the master feedwater level controller output failed downscale (i.e., controller can only provide 20 percent output signal irrespective of the input signal demand). The downscale failure

'as due to an unexpected and random f'ailure of an electrolytic capacitor in the controller.

C None IV. SAFETY'ALYSIS Loss of feedwater flow due to feedwater control system failures (or feedwater pump trips) is evaluated in the final safety analysis report as an abnormal operational transient.

The ESF actuations and safety systems functioned as design during the scram.

Plant safety was not adversely affected and the safety of plant personnel and the public was not compromised.

NRC Form 366(6-89)

~I NRC Form 366A (6-89)

U.S. 'pproved CLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

OMB No. 3150-0104 Expires 4/30/92 TEXT CONTINUATION FACILITY NAME (1) IOOCKET NUMBER (2)

I I I I SEQUENTIAL I IREVISIONI I I I Browns Ferry Unit 2 I I I I I TEXT (If more space is required, use additional NRC Form 366A's) (17)

V CORRECTIVE ACTIONS A.

1. The reactor was placed in a stable condition with the reactor pressure being maintained at 920 psig by the electro-hydraulic control system on turbine bypass valves. Excessive steam loads were transferred to the auxiliary boilers.

2. RFP 2C was returned to service to control reactor water level and maintain the water level in the vessel in the normal range.

B. V

1. TVA has revised the maintenance planning guide to include testing the wet type electrolytic capacitor.

2. TVA will evaluate the power stores procedure on monitoring shelf life of high-risk components.

3. TVA will evaluate implementation of the scram frequency reduction committee recommendation to design and install a fault tolerant digital feedwater control system.

VI. ADDITIONAL INFORMATION A. t The failed electrolytic capacitor in the master controller was found leaking.

B.

An automatic reactor scram due to a problem with the master level controller occurred on Unit 1 in 1985 (LER 259/85016). Although one electrolytic capacitor,was later identified to be out of tolerance and leaking, it was not believed to be the problem that caused the reactor trip. The scram was found to be caused by a cold solder joint in the controller.

NRC Form 366(6-89)

0 NRC Form 366A U.S. UCLEAR REGULATORY'OHHISSION Approved OHB No. 3150-0104 (6-89) Expires 4/30/92 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAHE (1)

Browns Ferry Unit 2 iOOCKET NUHBER I

I (2)

I, I E N B R I SEQUENTIAL I I REVI SION I I I

I I

I' I I TEXT (If more space is required, use additional NRC Form 366A's) (17)

VII.

1. TVA will evaluate the power stores procedure on .monitoring shelf life on high-risk components by July 21, 1992.

2. TVA will evaluate implementation of the scram frequency reduction committee recommendation to design and install a fault tolerant digital feedwater control system by the beginning, of Unit 2, Cycle 7 outage.

Energy Industry Identification System (EIIS) codes are identified in the text as [XX].

NRC Form 366(6-89)

0 0