ML18036A743

LER 92-004-01:on 920427,reactor Scrammed on Low Reacter Water Level Resulting in Actuation of Control Room Emergency Ventilation & Standby Gas Treatment Sys.Caused by Controller Output Failure.Maint Planning Guide revised.W/920608 Ltr

ML18036A743
Person / Time
Site:	Browns Ferry
Issue date:	06/08/1992
From:	Hsieh C, Zeringue O TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-004, LER-92-4, NUDOCS 9206150002
Download: ML18036A743 (20)

Text

ACCELERATED D+1UBUTION DEMONS~TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9206150002 DOC.DATE: 92/06/08 NOTARIZED: NO DOCKET FACIAL:50-260 Browns,Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 AUTH. NAME AUTHOR AFFILIATION HSIEH,C.S. Tennessee Valley Authority ZERINGUE,O.J. Tennessee Valley Authority RECZP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 92-004-01:on 920427,reactor scrammed on low reacter water level resulting in actuation of control room emergency ventilation a standby gas treatment sys.Caused by controller output failure.Maint planning guide revised.W/920608 ltr.

DISTRIBUTION CODE: IE22T TITLE: 50.73/50.9 Licensee COPIES RECEIVED:LTR Event Report (LER),

t ENCL IncidentU SIZE:

Rpt, etc.

NOTES:

REC1PIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL SANDERSiM. 1 1 HEBDONiF 1 1 ROSS,T. 1 1 1

INTERNAL: ACNW 2 2 ACRS 2 2 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFB10 1 1 NRR/DZPQ/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/DST/SPLB891 1 1 NRR/DST/SRXB 8E 1 1 REG EZ-L"~ 0'2 1 1 RES/DSIR/EIB 1 1 RGN2~- -'" -=01 1 1 EXTERNAL'G&G BRYCEi J ~ H 3 3 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHYiG.A 1 1 NSIC POOREiW. 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 NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

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

I~ 0 J

Tennessee Valley Authority. Post Office Box 2000, Decatur.'Alabama 35609 O. J. 'Ike'.Zering Ue Vice President, Browns Ferry Operations JUN O8 1992 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'0-260

'FACILITY OPERATING LICENSE DPR-52 LICENSEE EVENT REPORT LER-50-260/92004$ REVISION 1 Enclosed is Revision, 1 to LER 260/92004 submitted on May 26, 1992, concerning the Unit 2 reactor scram on low reactor water level that occurred on April 27,,

1992. The Unit '2 reactor scram caused actuation of the control room emergency ventilation and standby gas treatment on all trains. This revision provides additions of Unit 1 and Unit 3 as the other facilities involved in Item 8 of the LER Form 366. Additionally, the requirement for the four-hour, nonemergency report to NRC was changed to 10 CFR 50.72(b)(2)(ii).

Sincerely,,

0 J. Zeringue Enclosure cc: see page 2

/r I 920b150002 050002b0 'P20b08'DR ADOCK PDR

,2 U.S. Nuclear Regulatory Commission JUN 0 8 1992 cc (Enclosure):

INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30339 Paul Krippner American Nuclear Insurers Town Center, Suite 300S 29 South 'Main Street West Hartford, Connecticut 06107-2445 NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, P.O. Box 637 Athens, Alabama 35611 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

NRC form 366 U.. LEAR REQJULTORY 'COIIGSSION Approved OHB No. 3150-0104 (6-89) Expires 4/30/92 LICENSEE EVENT REPORT (LER)

FACILITY NAHE (1) (DOCKET NUHBER (2) w rr N TITLE (4) Automatic Reactor Scram on Low Reactor Water Level Due to Failure of'he Feedwater Level Control V

,I I I I I SEQUENTIAL (REVISION(' I I FACILITY NAHES (DOCKET NUHBER(S)

T r

,I I I I I I I I I I r w r i OPERATING I (THIS REPORT IS SUBHITTED PURSUANT TO THE REPUIREHENTS OF. 10 CFR 5:

HODE I I w (20.402(b) (20.405(c) (Z (50.73(a)(2)(iv) (73.71(b)

POWER I (20.405(a)( 1)(i ) (50.36(c)( 1) (50.73(a)(2)(v) (73 71(c)

LEVEL I (20.405(a)( 1)(ii) (50.36(c)(2) (50.73(a)(2)(vii) (OTHER (Specify in-(20.405(a)(l)(iii) I 150.73(a)(2)(i) (50.73(a)(2)(viii)(A) Abstract below and in (20.405(a)( l)(iv) (50.73(a)(2)(ii) (50.73(a)(2)(viii)(B) ( Text, NRC Form 366A) 1 N RTH 1 I.AREA CODE I mli n I I IREPORTABLEI I I IREPORTABLEI NT HA A T N P NT 0 I I I, 'I I SUBHISSION I I (

ABSTRACT (Limit to 1400 spaces, i.e., approximately fifteen single-space typewritten lines) (16)'n 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 wa'ter 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 master 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'valuate implementatj.on of the Scram Frequency Reduction Committee recommendation to design and install a fault tolerant digital feedwater control system.

NRC Form 366(6-09)

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 (2)

I /SEQUENTIAL t ./REVISION)

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

I. PLANT CONDITIONS

'Unit 2 was operati'ng i'n the run mode at approximately 100 percent power (3288 MM 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 l. and 3 were shutdown and defueled.

II. DESCRIPTION OF E(TENT A. J&~t:-

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]'ystems. 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'reatment [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 level 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'mmediately inserted into the Reactor Protection System (RPS) [JC].

-NRC Form 366(6-89)

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

TEXT 'ONTINUATION FACILITY NAHE (1) IOOCKET NUHBER (2)

I /SEQUENTIAL / /REVISION/

Browns Ferry Unit 2 I ~

I I' I TEXT (If more space's 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'll 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 service.) 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'eports 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. n t t t t t th Ivgnt:

The master controller output in the FMLC 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)

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

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

I I (SEQUENTIAL f [REVISION) )

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

G ESF actuations occurred 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's expected.

III. CAUSE OF THE EVENT A.

The immediate cause of the scram was reactor water level decreasing below the low level setpoint (+ll inche's).

B- RIII~m-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 was due to an unexpected and random failure of'n electrolytic capacitor in the controller.

C.

None IV. SAFETY ANALYSIS 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)

0 0 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 (2)

I ( SEQUENTIAL / )REVISION/

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

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 t t v t

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.

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)

I 0

NRC Form 366A U.S. NUCLEAR REGULATORY 'COHHISSION Approved OHB No. 3150-0104 (6-89) Expires 4/30/92 LICENSEE'VENT REPORT (LER) i TEXT CONTINUATION FACILITY NAHE (1) IDOCKET NUHBER (2)

I I I ISEQUENTIAL I IREVISIONI I I I I Browns 'Ferry Unit 2 I I I I I 4

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

VXX

1. TVA will evaluate, the power stores procedure on monitoring shelf life on

. high-risk components by July 21,, 1992.

2. TVA will evalua'te 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