ML17228B079

LER 95-002-00:on 950221,Unit 2 Automatically Tripped.Caused by Coalescing of Microscopic Conductive Particulates in Full Fluid,Acting as Short Circuit Between Ctr Diaphragm of Transmitter.New Level Transmitter placed.W/950321 Ltr

ML17228B079
Person / Time
Site:	Saint Lucie
Issue date:	03/21/1995
From:	Breen J, Sager D FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-95-093, L-95-93, LER-95-002-01, LER-95-2-1, NUDOCS 9503280194
Download: ML17228B079 (7)

Text

R.IC) H.lY PCCELERATED RIDS PROCESSING)

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESS3'C'4 NBR-9503280194 DOC.DATE: 95/03/21 NOTARIZED- NO DOCKET FACIL:~0-389 St. Lucie Plant, Unit 2, Florida Power & Light Co. 05000389 AUTH. NAME AUTHOR AFFILIATION BREEN,J. Florida Power & Light Co.

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

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 95-002-00:on 950221,Unit 2 automatically tripped. Caused by coalescing of microscopic conductive particulates in full fluid, acting as short circuit between ctr diaphragm of transmitter. New level transmitter placed.W/950321 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED LTR ENCL SIZE TITLE: 50.73/50.9 Licensee Event Report (LER), Inciden Rpt, etc.

NOTES RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 PD 1 1 NORRIS,J 1 1 INTERNAL: ACRS 1 1 AEOD/SPD/RAB 2 2 AEOD/SPD/RRAB NRR/DE/ECGB 1

1 1

1

~LE CENTERS NRR/DE/EELB 02 1 1

1 1

NRR/DE/EMEB 1 1 NRR/DISP/PI PB 1 1 NRR/DOPS/OECB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRSS/PRPB 2 2 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1' RES/DSIR/EIB 1 1 RGN2 FILE 01 1 EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H 2 2 NOAC MURPHY,G.A 1 1 NOAC POORE,W. 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM P!-37 (EXT. 504-2083 ) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 28 ENCL 28

Florida Power 8c Light Company, P.O. Box 128, Fort Pierce, FL 34854-0128 March 21, 1995 L-95-093 10 CFR 50.73 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Re: St. Lucie Unit 2 Docket No. 50-389 Reportable Event: 95-002 Date of Event: February 21, 1995 Automatic Reactor Tri on Low Steam Generator Water Level Due to a Failed Level Transmitter 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. ger Vice esident St. ie Plant DAS/EJB Attachment cc: Stewart D. Ebneter, Regional Administrator, USNRC Region Senior Resident Inspector, USNRC, St. Lucie Plant II Psttg gh 9503280194

~l P PDR ADOCK 950321 05000389 PDR an FPL Group company

NARC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEH PER RESPONSE TO COMPLY WITH LICENSEE EVERT REPORT (LER) THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEHENT BRANCH (HNBB 7714), U.ST NUCLEAR REGULATORY COMMISSION, (See reverse for required nunber of digits/characters for each block) 'WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3150-0'104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) PAGE (3)

St. Lucie Unit 2 05000389 1OF4 TITLE (4) Automatic Reactor Trip on Low Steam Generator Water Level due to a Failed Level TI~sinitter EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

SEQUENTIAL REVISION FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUHBER 02 21 95 95 --002 00 3 21 FACILITY NAHE DOCKET NUMBER OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10'CFR 5: (Check one or mor e) (11)

MODE (9) 20 '02(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b)

POWER 20.405(a)(1)(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c) 100 LEVEL (10) 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) OTHER 20.405(a)(1)(fii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) (Speci fy in 20.405(a)(1)(iv) 50.73(a)(2)(ii) 50.73(a)(2)(viii)(B) Abstract below and in Text, 20.405(a)(1)(v) 50.73(a)(2)(iii) 50.73(a)(2)(x) NARC Form 366A)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUHBER (Include Area Code)

Jack Breen, Shift Technical Advisor 407-465-3550 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

I CAUSE REPORTABLE REPORTABLE SYSTEH COMPONENT MANUFACTURER CAUSE SYSTEH COHPONENT MANUFACTURER TO NPRDS TO NPRDS X SJ LT X999 SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR YES SUBHISSION (If yes, complete EXPECTED SUBMISSION DATE).

X NO DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On February 21, 1995, at 1317 hours0.0152 days <br />0.366 hours <br />0.00218 weeks <br />5.011185e-4 months <br />, St. Lucie Unit 2 autaTatically tripped frcm 100%

power, due to low water level in the 2A Steam Generator. In accordance with Emergency Operating Procedure (EOP) -1, Standard Post Trip Actions were performed. When all safety function status checks were satisfactorily completed, the operating staff exited EOP-1 and entered EOP-2, "Reactor Trip Recovery". Normal Steam Generator water levels were regained and the Unit was stabilized in Mode 3.

The event was initiated when level transmitter LT-9011 failed high. The most likely root cause of the level translDitter failure was determined, by the vendor, to be due to coalescing of microscopic conductive particulates in the circuit between the center diaphragm of the transmitter and one of the sensor cell fill fluid which acts as a short capacitor plates.

Corrective Actions for this event include: 1) Verification that no other ccmponents failed in the Feedwater Control System. 2) A bench test on LT-9011 to confirm the level transmitter had failed. 3) A new level transmitter for LT-9011 was placed in service. 4)

The failed transmitter was sent to the manufacturer for a detailed failure analysis.

5) Engineering is evaluating the failure analysis, to determine the appropriate actions to prevent reoccurrence of this event and its impact on other plant transmitters.

NARC FORH 366 (5-92)

NARC FORH 366A U ~ S ~ NUCLEAR REGULATORY COHHISSION APPROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS.

F FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EV1WZ REPORT (LER) THE INFORMATION AND RECORDS MANAGEHEHT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMHISSIOH ~

TEXT CONTINUATION WASHINGTON, DC 20555-0001 AHD TO THE PAPERWORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SEQUENTIAL REVISION YEAR NUHBER NUHBER 2 OF 4 St. Lucia Unit 2 05000389 95 002 00 TEXT (It more space is required, use additional copies of NARC Form 366A) ('17)

On February 21, 1995, at 1317 hours0.0152 days <br />0.366 hours <br />0.00218 weeks <br />5.011185e-4 months <br /> with St. Lucie Unit 2 in Mode 1 at 100% power, the reactor autcnatically tripped due to valid low water level in the 2A Steam Generator(SG) .

Just prior to the reactor trip, a Reactor Operator (RCO) witnessed the alarming of annunciator G1, "2A S/G LEVEL HI/LOa (EIIS:IB) Upon acknowledgement of the annunciator,

~

the RCO chserved SG level indications. He looked at both wide range and narrow range SG level indications. The wide range level indication indicated that level in the 2A SG was rapidly decreasing, in contrast the narrow range SG level indication indicated that the 2A SG level was at 3.00~. The RCO then observed the four nannw range safety channels. The narrow range safety channels indicated that level in the 2A SG was rapidly decreasing.

The RCO prcmptly switched the feedwater indicator controller frcm aut(mat).c to manual, in an attempt to regain SG water level. At that time, the Assistant Nuclear Plant Supervisor directed the RCG to trip the reactor. However, before that action could be completed the reactor tripped autanatzcally due to valid low water level in the 2A SG.

All plant safety functions were met. The Auxiliary Feedwater Actuation System (AFAS) (EIIS:BA) functioned as required during this event. Additional observations, were the lifting of the suction relief valve of tge 2A Main Feedwater Pump (MFP) (EIIS:SJ),.

followed by the tripping of the 2A MFP due to a low flow condition. This was a result of the closure of the MFRV. The lifting of the SB High Pressure Feedwater Heater (HPFH) (EIIS:SN) safety relief valve and several Mam Steam Safety Valves (MSSV) (EIIS:SB) were also observed, due to the increase in pressure of the secondary system following the MFRV closure. The SB HPFH safety valve and the MSSV's reseated and the plant was stabilized in Mode 3, Hot Standby.

The event was initiated when level transmitter LT-9011 failed high. This transmitter sends an input signal to the Feedwater Control System, which then sends an output signal to FCV-9011, the MFRV, level recorder 9011 and FIC-9011. When the transmitter failed high, indicating 100% level in the 2A SG, MFRV went closed, all it sent a signal to close the MFRV. When the feedwater to the 2A SG was stopped. The level in the 2A SG started decreasing rapidly. When the level in the 2A SG reached 56%, annunciator G1, "2A S/G LEVEL HI/LO", alarmed. Annunciator G1 receives its signal frcm level transmitter 9005 (LT-9005) . Similar to LT-9011, LT-9005 is also a narrow range level tranmlitter. By the time annunciator G1 had alarmed, water level in the 2A SG was already decreasing rapidly.

The reactor tripped autanatically, as required, at 20.5% SG level.

The most likely root cause of the level trEInsmitter failure was determined, by the vendor, to be due to coalescing of microscopic conductive particulates in the acts as a short circuit between the center diaphracpn of the transItu.tter and one of the fill fluid which sensor cell capacitor plates. A review of St. Lucre maintenance history did not indicate any previous similar failures of this type of level transnlitter.

I I

I i

NARC FORM 366A S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY MITH THIS INFORMATION COLLECTIOH REQUEST: 50.0 HRS.

FORHARD COMMENTS REGARDIHG BURDEN ESTIMATE TO ZZCmSEE EVE REPORT (LER) THE INFORMATION AND RECORDS MANAGEHENT BRANCH (MHBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, TEXT CONTINUATION NASHIHGTOH, DC 20555-0001 AHD TO THE PAPERNORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILiTY NAME ('1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SEQUENTIAL REVISION YEAR St. Lucie Unit 2 05000389 NUMBER NUMBER 3OF4 002 00 TEXT (If more space is required, use additionaI copies of HARC Form 366A) (17)

This event is reportable under 10 CFR 50.73 a.2.iv. as, "any event or condition that the resulted in manual or autcmatic actuation of any engineered safety feature, including Reactor Protection System.

The plant response to this event was bounded by the accident analysis of the St. Lucie Unit 2 Final Safety Analysis Report (FSAR), section 15.2, "Decreased Heat Removal by the Secondary System". The actual plant respcB)se was IIK)re conservative because of the following:

1) Only one Feedwater Regulating Valve closed in this event, instead of a total loss of normal feedwater.

2) The reactor autcmatically tripped due to low 2A SG water level. In the accident analysis, the reactor is assumed to trip on high pressurizer pressure.

During the transient primary pressure never exceeded 2300 psia, therefore the high pressurizer pressure trip setpoint of 2370 psia was never challenged.

All plant safety functions were met. The Auxiliary Feedwater Actuation System functioned as required during this event. Therefore, the health and safety of the public were not affected by this event.

1) System verification showed no other components failed in the Feedwater Control System.

2) LT-9011 was bench tested and confirmed to have failed high.

3) A new level transmitter was placed in service for LT-9011.

4) The corresponding level transmitter for the 2B SG (LT-9021) was replaced.

5) The failed level transmitter was sent to the manufacturer for a detailed failure analysis.

6) Engineering is evaluating the failure analysis, to determine the appropriate actions to prevent reoccurrence of this event and its impact on other plant transmitters.

7) Engineering will evaluate the feasibility of design modifications, to minimize or eliminate plant trip single point vulnerabilities in the Feedwater Control System.

8) Training will evaluate this event for use in Licensed Operator Requalification training .

C FORM 366A (5-92)

HARC FORM 366A U~S ~ NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5-92) EXP I RES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 NRS.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO LICENSEE EVENT REPORT (LER) THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, TEXT CONTINUATION ~

WASHIHGTOH, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3180-0104)) OFFICE OF MANAGEMENT AND BUDGET WASHINGTON DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

SEQUENTIAL REVISION YEAR NUMBER NUMBER St. Lucie Unit 2 05000389 4 OF 4 95 002 00 TEXT (If more space is required, use additional copies of NARC Form 366A) (17)

Rosermmt Differential Pressure Transmitter, Model g1152DP5E92PB LER 335-91-05, "Reactor Trip from 100% Power on Low Steam Generator Water Level caused by a De-energized Feedwater Regulating Valve due to a Deficient Procedure."

LER 335-88-08, "Reactor Trip on Low Steam Generator Level Due to Inadvertent Closure of a Main Feedwater Recrating Valve."

LER 335-88-03, "Reactor Trip on Low Steam Generator Level Due to Main Feed Regulating Valve Ecpipment Failure. "

HARC FORM 366A (5-92)