ML17223A386

LER 89-007-00:on 890923,unit Experienced Dropped Control Element Assembly in Regulating Group 1.Most Probable Cause of Event Was Blown Fuse.Root Cause for Breaker Trip Not Conclusively Identified.Blown Fuse replaced.W/891023 Ltr

ML17223A386
Person / Time
Site:	Saint Lucie
Issue date:	10/23/1989
From:	Andrea Johnson, Sager D FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-89-381, LER-89-007-02, LER-89-7-2, NUDOCS 8910310177
Download: ML17223A386 (7)

Text

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ACCESSION NBR:8910310177 DOC.DATE: 89/10/23 NOTARIZED: NO -DOCKET FACIL:50-389 St. Lucie Plant, Unit 2, Florida Power & Light Co. 05000389 AUTH. NAME AUTHOR AFFILIATION JOHNSON,A.B.

Florida Power & Light Co.

SAGER,D.A. Florida Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 89-007-00:on 890923,manual reactor trip resulting from multiple dropped rods due to unrelated equipment failures.

W/8 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR I ENCL I 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 "NCL h.

PD2-2 LA 1 1 PD2-2 PD 1 1 NORRIS,J 1 1 INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 ACRS WYLIE 1 1 * 'EOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 DEDRO 1 1 NRR/DEST/ESB 8D 1 1 NRR/DEST/ICSB 7 1 1 NRR/DEST/MEB 9H 1 1 NRR/DEST/MTB 9H 1 1, NRR/DEST/PSB 8D 1 1 NRR/DEST/RSB 8E 1 1 NRR/DEST/SGB 8D 1 1 NRR/DLPQ/HFB 10 NRR/DLPQ/PEB 10 1 1 NRR/DOEA/EAB NUDOCS-ABSTRACT ll 1 1

1 1

1 1 REG FIL B 10 02 2

1 2

1 RES/DsiR/EIB 1 1 GN2 FILE 01 1 1 EXTERNAL 'G&G WILLIAMSg S 4 4 L ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC MAYSgG -1 1 NSIC MURPHYiG.A 1 1 NUDOCS FULL TXT 1 1 h

p,o-"I FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 38 ENCL 38

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P.O. Box14000, Juno Beach, FL 33408 0420 OCTOBER 2 5 3989 L-89-381 10 CFR 50.73 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:

Re: St. Lucie Unit 2 Docket No. 50-389 Reportable Event: 89-07 Date of Event: September 23, 1989 Manual Reactor Trip Resulting From Multiple Dro ed Rods Due To Unrelated E i ment Failures The attached Licensee Event Report is being submitted pursuant to the requirements of 10 CFR 50.73 to provide notification of the subject event.

Very truly yours, D. A. er Vice P ident St. Lucie Plant DAS/JRH/gmp Attachment cc: Stewart D. Ebneter, Regional Administrator, Region Senior Resident Inspector, USNRC, St. Lucie Plant II, USNRC 8S'10310177 891023 P DR *DOCK 0 5000389 8 PNU l ~ i t ~

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NRC Form 3SSA U.S. NUCLEAR REOULATORY COMMISSION (943)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS ND. 3(SOWIOE EXPIRES: 8/31/88 FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (8) PACE (3)

YEAR SEOUENTIAL re%'EvoloN Q$ ./F NUMSER ST. LUCIE, UNIT 2 0 5 0 0 0 3 8 9 8 9 007 0 0 02 "0 5 TEXT ///more e/reoe /F>>r/rr/rat, o>> edd/done/ HRC Fomr 3SSA8/ ll/I DESCRIPTION OF EVENT At 0558 on September 23, 1989, with St. Lucie Unit 2 in Mode 1 at 100% power, the unit experienced a dropped Control Element Assembly (CEA)(EIIS:AA) in Regulating Group 1. CEAItt: 70 had dropped fully into the core. The Reactor Control Operator (RCO) manually reduced the main turbine load to match reactor power. The unit entered the Technical Specification ACTION statement 3.1.3.1.d for Movable Control Assemblies and had 63 minutes to realign GEE 70 within 15 inches of other CEAs in its group or be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The control room crew began to investigate the GEE 70 drop and implemented the "CEA Off>>Normal Operation And Realignment" procedure. The Assistant Nuclear Plant Supervisor (ANPS) went to the local Control Element Drive Mechanism Control System (CEDMCS)(EIIS:AA) panel and observed that the disconnect breaker for CEASE 70 had opened. The disconnect breaker for GEE '70 was reclosed and the attempt to withdraw CEAIP 70 was unsuccessful.

The Nuclear Plant Supervisor (NPS) ordered a controlled power reduction due to an inoperable CEA. The RCO selected Manual Group 5 on CEDMCS and began inserting rods. At 0613, after inserting the control rods 2 to 3 inches, four CEAs in Regulating Group 5 dropped in the core. The ANPS ordered the RCO to manually trip the reactor and carry out the Standard Post Trip Actions. The reactor was immediately tripped from 96% power. The Steam Bypass Control System (SBCS)(EIIS:SB) operated to reduce primary average temperature (T-avg.) to the zero percent power setpoint of 532 degrees F.

Auxiliary Feedwater (AFW)(EIISIBA) actuated on low steam generator levels and functioned as designed with feedwater being supplied to both steam generators for Reactor Coolant System (RCS)(EIISIAB) heat removal. The 2A AFW and 2B AFW pumps (electric motor driven) were supplying feedwater to the "A" and "B" steam generators, respectively. The 2C AFW pump (steam driven) also supplied feedwater to both the "A" and 'B" steam generators. The Standard Post Trip Actions were completed and the unit was quickly stabilized in Hot Standby, Mode 3.

The trip was an uncomplicated reactor trip with all safety functions verified as being met; however, some equipment had inadequate performance following the reactor trip. The RCO had to take manual control of PCV-8802, 10% steam bypass valve, because the valve did not automatically open. Subsequent to operator action, the SBCS functioned properly to maintain RCS heat removal. Approximately 45 minutes into the event, it was identified that the "A" AFW pump discharge valve, MV-09-9, did not respond to demand change from the control room nor could the valve be operated manually. MV-09-9 was mechanically bound such that AFW flow was approximately 200 gpm. The "B" AFW pump discharge valve, MV"09-10, limit switch position in the control room did not agree with feed flow response; i.e., MV-09-10 indicated full open but providing only 50% of maximum feed flow rate. The control room elected to locally operate MV-09-10.

The 2C AFW pump, steam driven, was available to supply feedwater NRC FORM 388A (983)

r /3 NRC Form 366A U.S. NUCLEAR REOULATORY COMMISSION (943)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OMS NO. 3160-0104 EXPIRES: 8/31/88 FACILITY NAME (11 DOCKET NUMBER (2) LER NUMBER (6) PACE (3)

@8 SEOVENTIAL REV ISIQ N YEAR NUMBER 'd(F: NVMEER ST. LUCIE, UNIT 2 0 5 .0 0 0 3 8 9 0 0 7 0 003 OF 0 5 ~

TEXT ///moro EOoco /F ror)rr/rod, oFF odd/dorM/HRC Form 3684'F/ (17) to both steam generators via its own'edicated discharge valves, MV-09-11 6 MV-09-12. The 2C AFW pump is designed with pump capacity equivalent to two electric motor driven AFW pumps. The SBCS and AFW problems did not impact the control room crew's ability to maintain the plant stable in Hot Standby, Mode 3. The cause of inadequate equipment performance will be discussed in the Analysis Of Event section.

CAUSE OF EVENT The cause for the reactor trip was a manual reactor trip by the RCO following four dropped CEAs. The four CEAs that dropped were GEE 69, CEAIP 72, CEAIP 75, and GEE 78, which are powered from CEA Subgroup 818. The immediate cause of the four CEAs dropping was the tripping of CEA Subgroup $ 18 breaker. The root cause for the subgroup breaker trip has not been conclusively identified; however, testing of the subgroup breaker revealed that the breaker tripped at a current of approximately 30 amps, which is less than designed.

The subgroup breaker is designed for a 40 amps continuous load.

The CEA subgroup breaker was replaced.

The root cause for CEAIP 70 dropping has not been conclusively identified. CEAfk 70 is powered from CEA Subgroup (P19. The most probable cause was a blown fuse in CEA Subgroup 819. This caused the CEA to lose one phase of the three .phase power supply. The fuse was replaced for CEA Subgroup 419.

ANALYSIS OF EVENT This event is reportable under 10 CFR 50.73 (a)(2)(iv), "any event or condition that resulted in a manual or automatic actuation of any Engineered Safety Feature, including the 'eactor Protection System."

Having one full length CEA drop followed by a full length CEA Subgroup drop is not an analyzed event in the St. Lucie Unit (P2 Final Updated Safety Analysis Report. The Fuel Resources Group analyzed this event and determined th'at no Departure from Nucleate Boiling Ratio (DNBR) or Local Power Density (LPD) limits were violated at any time during this event. In addition, no Incore Neutron Detector alarms were received prior to the reactor trip.

The reactor trip was observed to be a routine manual reactor trip.

The resulting transient was well enveloped by the St. Lucie Unit 8'2 Final Updated Safety Analysis Report.

PCV-8802 failed to automatically open on demand. PCV-8802, receives input from a T-avg. program to maintain RCS average temperature at 532 degrees F. The RCO transferred the " individual controller of PCV"8802 to manual control for RCS heat removal. The failure NRC FORM 366A (903)

NRC Form 366A US. NUCLEAR REOULATORY COMMISSION (943)

'LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMB NO. 3)60M(06 EXPIR ES: 8/31/68 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PACE (3)

YEAR SEQUENTIAL REVISION ST. LUCIE, UNIT 2 NUMBER iver NUMSER TEXT fifnroro sposo/1 6//rr)od, Irso oddnrisno/'NRC Form BB)r('s/ () 7) 0500038989 007 0 0 4 OF 05 of the valve to auto open did not significantly affect the performance of the Steam Bypass Control System. In addition', the Atmospheric Dump Valves were available to provide an alternate means to accomplish the RCS heat removal safety function.

MV-09-9 lost the ability to be throttled due to mechanical binding in the valve's actuator after the RCO had throttled the discharge valve to a feed flow rate of 200 gpm. The cause for mechanical binding in the actuator was due to excessive wear of the upper stem threads leading up to galling and eventual binding between the upper stem and stem nut. MV-09-10 had erroneous limit switch position indication in the control room. The cause for the erroneous indication was the lack of engagement between the limit switch drive pinion and the drive sleeve bevel gear in the valve's actuator.

The 2C AFW pump, steam driven, was available to supply feedwater to both steam generators via its own dedicated discharge valves, MV-09-11 6 MV-09"12. The 2C AFW pump is designed with pump capacity equivalent to two electric motor driven AFW pumps.

From the analysis of the event, all Safety Functions were met and maintained; therefore, the health and safety of the public were not affected by this event.

CORRECTIVE ACTIONS C

1. The fuse was replaced for CEA Subgroup 819.

2. CEA Subgroup 818 breaker was replaced; Future plans to test CEA Subgroup breakers during upcoming refueling outages.

3. Instrument & Control De'partment investigated the inadequate performance, of the Steam Bypass Control System (SBCS) and did not find any problems with PCV-8802. The SBCS was tested and the system functioned properly. No cause could be found why PCV-8802 did not open automatically.

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4. Mechanical Maintenance replaced 'the Limitorque operator upper stem and stem nut for MV-09-9. ,The actuator tested satisfactorily and was returned to service.

5. Electrical Maintenance replaced the worn limit switch drive pinion gear for MV"09-10. The limit switch pinion gear and the drive sleeve bevel gear was verified for proper gear engagement. The actuator tested satisfactorily and was returned to service.

NRC CORM 3664 (94)3)

NRC Form 3SSA U.S. NUCLEAR REOULATORY COMMISSION (983 l LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED OM8 NO. 3(9)W(04 EX PIRE 8( 8/31/88 FACILITYNAME (1) OOCKET NUMEER (2) LER NUMeER (e) PACE (3)

YEAR SSOVENTIAL RSV IS IO N NVMSSR NVM SR ST. LUCIE, UNIT 2 3 8 9 8 9 007 0 0 5 QF0 TEXT ///moro FPooo /F /or)o/rdd, VFO Odd/dorN/H/IC /orrrr 38843/ Ill)

ADDITIONAL INFORMATION FAILED COMPONENT:

Component: CEA Subgroup 8'18 Breaker, 250 VAC 50/60 Hz Manufacturer: Heinemann Electric Company Model (I)s AM3-AZA2A3-A Component: Semiconductor Fuse Manufacturer: International Rectifier Model I/)2 SF25X60 Component: MV-09-9 & MV-09-10 Actuators Manufacturer: Limitorque Model (/'( SMB-000 PREVIOUS SIMILAR EVENTS.

See LER 8335-80-050 for previous manual reactor trip due to multiple dropped rods.

NRC FORM SddA (9.83)