ML17228B230

LER 95-003-00:on 950708,automatic Reactor Trip Occurred During Turbine Overspeed Surveillance Testing Due to Personnel Error.Personnel Involved in Event Counseled & Procedure Changes Being made.W/950803 Ltr

ML17228B230
Person / Time
Site:	Saint Lucie
Issue date:	08/03/1995
From:	Benken E, Sager D FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-95-218, LER-95-003-02, LER-95-3-2, NUDOCS 9508080069
Download: ML17228B230 (7)

Text

jpRIORITY. 1e, (ACCELERATED RIDS PROCESSING)

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9508080069 DOC.DATE: 95/08/03 NOTARIZED: NO DOCKET g FACIL:50-335 St. Lucie Plant, Unit 1, Florida Power & Light Co. 05000335 AUTH. NAME AUTHOR AFFILIATION BENKEN,E.J. Florida Power & Light Co.

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

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 95-003-00:on 950708,automatic reactor trip occurred during turbine overspeed surveillance testing due to personnel error. Personnel involved in event counseled &

procedure changes being made.W/950803 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 PD2-1 PD 1 1 NORRIS,J 1 1 INTERNAL: ACRS 1 1 2 2 AEOD/SPD/RRAB 1 1 EHZE~R 1 1 NRR/DE/ECGB 1 1 NRR/DE/EELB 1 1 NRR/DE/EMEB 1 1 NRR/DISP/PIPB 1 1 NRR/DOPS/OECB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SPSB/B 1 1 NRR/DSSA/SRXB 1 1 RES/DSIR/EIB 1 1 RGN2 FILE 01 1 1 EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H 2 2 NOAC MURPHY,G.A 1 1 NOAC POOREiW 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 NOTE TO ALL "RZDS" RECZPZENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D8 (415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 27 ENCL 27

Florida Power & Light Company, P.O. Box 128, Fort Pierce, FL 34954.0128 August 3, 1995 L-95-218 10 CFR 50.73 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Re: St. Lucie Unit 1 Docket No. 50-335 Reportable Event: 95-003 Date of Event: July 8, 1995 Automatic Reactor T i Duri Turb' 0 e s e d Sur eilla ce Tes i due t Pe so el Erro 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 r sident St. Lu ie Plant DAS/EJB Attachment cc: Stewart D. Ebneter, Regional Administrator, USNRC Region Senior Resident Inspector, USNRC, St. Lucie Plant II 9508080069 950803 PDR ADOCK 05000335 8 PDR

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~ NRC FORH 366 U.S. IN)CLEAR REGUULTORY C(NITSS ION APPROVED BY QS NO. 3150-0104 (5-92) EXPIRES 5/31/95 EST IHATED BURDEN PER RESPONSE TO C(WPLY lllTH THIS INFORHAT10N COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) FORMARD COHHENTS REGARDING BURDEN EST I HATE TO THE INFORHAT ION AND RECORDS HANAGEHENT BRANCH (HNBB 7714), U.S. NUCLEAR REGULATORY COHHISSIONg (See reverse for required number of digits/characters for each block) MASHINGTON, OC 20555-0001 AND TO THE PAPERNORK REDUCTION PROJECT (3140-0104), OFFICE OF HANAGEKENT AND BUDGET MASHINGTON OC 20503.

FACILITY IWK (1) DOCKET IRWBER (2) PAGE (3)

St. Lucie Unit 1 05000335 1OF5 TITLE (4) Automatic Reactor Trip During Turbine Overspeed Surveillance Testing due to Personnel Error.

EVENT DATE 5 LER NNBER 6 REPORT DATE 7 OTHER FACILITIES INVOLVED 8 SEQUENT IAL REVISION FACILITY NAHE DOCKET NUHBER DAY YEAR HONTH DAY YEAR HONTH YEAR NUHBER NOSER N/A FACILITY NAHE DOCKET NUHBER 07 08 95 95 003 0 08 03 N/A OPERAT INGa THIS REPORT IS SUSHI TTED PURSUANT To THE REQUIREHENTS OF 10 CFR: Check one or more 11 IRmE (9) 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b) 20.405(a)(1)(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)

LEVEL (10) 100 20.405(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vii) OTHER 20.405(a)(1)(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A) (Specify 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) NRC Form 366A LICENSEE CONTACT FOR THIS LER 12 NAHE TELEPHONE NUHBER (Include Area Code)

Edwin J. Benken, Licensing Engineer (407) 468-4248 COMPLETE ONE LINE FOR EACH C(NPONENT FAILURE DESCRIBED IN THIS REPORI'3 REPORTABLE REPORTABLE CAUSE SYSTEH COHPONENT HANUFACTURER CAUSE SYS'IEH COHPONENT HANUFACTURER TO NPRDS TO NPRDS SUPPLEHENTAL REPORT EXPECTED 14 EXPECTED HONTH DAY YEAR YES SUBHI SSI ON (If yes, EXPECTED SUBHISSION DATE).

X No DATE (15) coagulate ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On July 8, 1995, Unit 1 was operating at 100 percent reactor power. Operations personnel were conducting a scheduled Turbine overspeed trip surveillance per an approved plant procedure. During the portion of the surveillance that tests a solenoid valve for Overspeed Protection Control (20-1 OPC) a utility non-licensed Operator failed to close an isolation valve as directed by the procedure. Failure to close this valve allowed electro-hydraulic (EH) fluid from the Governor valves (GV) and Intercept valves (IV) to drain when the solenoid valve was opened in a subsequent step. Draining of the EH fluid caused closure of the Main Turbine Governor and Intercept valves which resulted in an automatic reactor trip.

The root cause of this event was cognitive personnel error on the part of a utility non-licensed operator who failed to properly implement a procedural step during performance of a surveillance.

Corrective actions for this event: 1) Operations personnel involved with the event were counselled. 2)

Procedure changes are being made to incorporate human factors improvements and additional step verifications. 3) Other load threatening surveillances are being reviewed to determine if generic changes are warranted. 4) A technical subcommittee is evaluating this event for additional corrective actions to prevent reoccurrence. 5) Site management held a trip review meeting open to all disciplines for lessons learned from this event.

NRC FORH ( )

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(, KRC FORH 366A U.S NUCLEAR REGULATORY CQKI SSI(NI APPROVED BY CHB NO 3150-0104 (5-92) EXPIRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO COMPLY NITH THIS INFORHAT ION COLLECTION REQUEST: 50.0 HRS.

FORNARD COHHENTS REGARDING BURDEN ESTIHATE TO THE LICENSEE EVENT REPORT (LER) INFORMATION AND RECORDS HANAGEHENT BRANCH (HNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COHHISSION, MASHINGTOM, DC 20555-0001( AND TO THE PAPERNNK REDUCT IOM PROJECT (31/0-0104), OFF ICE OF HAMAGEHEMT AND BUDGET MASHIMGTON DC 20503.

FACI LITY MAHE 1 LER MINBER 6 PAGE 3 SEQUEHT IAL REVISION St. Lucie Unit 1 05000335 95 003 2OF5 TEXT If mo e s ce is r ired use additional co ies of M C Form 3 (17)

DESCRIPTION OP THE EVENT On July 8, 1995, St. Lucie Unit 1 was operating at 100 percent Reactor power. A utility non-licensed Operator was performing the monthly turbine overspeed trip te "t in accordance with an approved plant procedure. The non-licensed operator was performing the steps of the procedure while a utility licensed Operator maintained radio communication with the control room.

During the portion of the test which checks the operability of an Overspeed Protection Control (OPC) solenoid valve, SE22138 (EIIS:TG), the procedure directed the operator to unlock and close V22482 (EIIS:TG), "EH Test Header to 20-1/OPC Isolation." This is the electro-hydraulic (EH) fluid inlet isolation to the OPC solenoid valve. This step ensures that the OPC solenoid valve is isolated from the actual EH fluid system (EIIS:TG) supplying the turbine Governor (GV) and Intercept valves (IV) (EIIS:SB) prior to testing the solenoid. The NPO removed the locking device from isolation valve V22482, but was momentarily distracted by placing the locking device in a secure position, and failed to close the valve as directed by the procedure. When the next step of the procedure was executed (the actual stroke testing of solenoid valve SE22138) EH fluid was drained from the GVs and IVs causing the GVs and IVs to rapidly close. Closure of the turbine valves quickly reduced steam flow through the turbine which resulted in a reactor trip from high pressurizer pressure at 1122 hours0.013 days <br />0.312 hours <br />0.00186 weeks <br />4.26921e-4 months <br />.

Emergency Operating Procedure (EOP)-1, "Standard Post Trip Actions" was immediately implemented.

The Reactor Coolant System (RCS) Power Operated Relief Valves (PORV) (EIIS:AB) actuated as designed during the time the high Pressurizer pressure signal was present (less than 4 seconds), and then reclosed. The maximum RCS pressure reached during this event was 2430 psia. The maximum secondary pressure reached was 1023 psia.

Operators observed increasing level in the 1A SG after the trip and closed the 15 percent feedwater bypass valve. Level continued to increase and the Control Room Operators closed the isolation valve for the 1A Feedwater Regulating Valve (EIIS: JB). The 1B Main Feedwater Pump (MFW)(EIIS:SJ) subsequently tripped from a low flow condition, and the 1A MFW Pump tripped due to high level in the 1A SG. The 1B MFW Pump was restarted and SG levels were then controlled within the normal band.

A relief valve in the Letdown Level Control System (EIIS:CB) opened during the event due to the system transient, and subsequently closed when Control Room operators reduced the letdown pressure controller (EIIS:CB) setpoint. The Steam Generator Safety Valves (EIIS:SB) functioned as designed to limit SG pressure during the initial transient. The Steam Bypass Control System (SBCS)

(EIIS:Jl) functioned properly to control RCS temperature during this event.

NRC- FORH 366A (5-92)

NRC FORN 366A U.S. NUCLEAR REGUIATOIY COBIISSIOI APPROVED BY OS NO. 3150-0104 (5 92) EXP IRES 5/31/95 ESTIHATED BURDEN PER RESPONSE TO CONPLY MITH THIS INFORHATION COLLECTION REQUEST: 50.0 HRS.

FORMARD CONHENTS REGARDING BURDEN EST IHATE TO THE LICENSEE EVENT REPORT (LER) INFORHATION AND RECORDS HANAGEHENT BRANCH (HHBB TEXT CONTINUATION 7714), U.S ~ NUCLEAR REGULATORY COHIISSIOI,

'MASHINGTOH, DC 20555.0001( AND TO THE PAPERMDRK REDUCT IOI PROJECT (3140.0104), OFFICE OF IIANAGEHENT AND BUDGET MASH INGTON DC 20503.

FACILITY NANE 1 DOXET NNBER 2 LER NWBER 6 PAGE 3 YEAR SEQUENTIAL REVI SIOI St. Lucie Unit 1 05000335 95 003 0 30F5 TEXT lf aero s ce is r ired use edditiooeI co ies of NRC Fo 366A (17)

DESCRXPTXON OF THE EVENT conti ued The Control Room crew completed the actions of EOP-01, Standard Post Trip Actions", and implemented EOP-02, "Reactor Trip Recovery" after diagnosing an uncomplicated trip. Upon completion of the Reactor Trip Recovery procedure, the unit was maintained in a stable, Mode 3 condition for post trip review and event investigation.

CAUSE OF THE EVENT The cause of this event was cognitive personnel error by a utility non-licensed operator who failed to correctly implement a procedural step during performance of a turbine overspeed trip surveillance. The operator was momentarily distracted by placing a valve locking device in a secure position, and did not close the valve as directed by the procedure.

YSXS OF THE EVENT This event is reportable under the requirements of 10 CFR 50.73.a.2.iv, as "any event that resulted in a manual or automatic action of any Engineered Safety Feature."

The closure of the Main Turbine Governor and Intercept valves caused a rapid reduction in secondary steam flow. The effect of the reduction in secondary steam demand was an increase in SG pressure and temperature, and RCS temperature and pressure. Increasing RCS pressure resulted in an uncomplicated Reactor trip on high pressurizer pressure as designed.

An investigation performed after the event revealed that the calibration on the 1A Main Feedwater Regulating Valve (FCV-9011) electro-pneumatic transducer (E/P) had drifted, so that the feedwater flow control valve did not close fully as expected on the plant trip. This caused the 1A Steam Generator level to increase above the normal value to the high level trip setpoint for the Main Feedwater Pump. Closing the Main Feedwater Block valve secured the flow to the 1A SG from FCV-9011, stabilizing SG level.

This event is bounded by section 15.2.7 of the St. Lucie Unit 1 Updated Final Safety Analysis Report (UFSAR) "Loss of External Electrical Load or Turbine Stop Valve Closure." This section describes a rapid, large reduction of power demand on the reactor while operating at full power. The UFSAR states, "When the turbine stop/control valve closes, the steam flow is terminated, causing the secondary system temperature and pressure to increase. The primary-to-secondary heat transfer decreases as secondary system temperature increases. If the reactor is not tripped when the turbine is tripped,. .the reactor will trip on high pressurizer pressure, reducing the primary heat source."

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NRC FORII 366A (5-92)

NRC FORll 366A U.S. NUCLEAR REGULATORY CQHIISS ION APPROVED BZ QNI NO. 3150-0104 (5-92) EXP I RES 5/31/95 ESTIHATED BURDEN PER RESPOHSE TO CONPLY MITH THIS IHFORNATION COLLECTION REQUEST: 50.0 HRS.

FORMARD COHHENTS REGARDING BURDEN ESTIHATE TO THE

'LICENSEE EVENT REPORT (LER) INFORHATION AND RECORDS NANAGENEHT BRANCH (HHBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY CONHI SS ION, MASHINGTQI, DC 20555-0001 AND TO THE PAPERMORK REDUCTIOH PROJECT (3140-0104), OF FICE OF HANAGEHENT AND BlmGET MASHINGTQl DC 20503.

FACILITY lUWE 1 DOCKET NQIBER 2 LER NWBER 6 PAGE 3 YEAR SEQUENT IAL REVISION St. Lucie Unit 1 05000335 95 003 4OF 5 TEXT If mor s c is r ired use additions co ies of N C Form 366 (17)

ANALYSIS OP THE EVENT cont,i ued In addition to the above, UFSAR section 15.2.7, states that, "The mitigative features of the pressurizer spray, pressurizer relief valves (PORV), and the Steam Bypass System are assumed not to function so as to exacerbate the calculated pressurization of the primary system. The purpose,...is to demonstrate that the primary safety relief capability is sufficient to limit primary pressure to less than 110% of the design pressure (2750 psia), and to demonstrate that the secondary safety relief capacity is sufficient to limit secondary pressure to less than 110/ of the design pressure (1100 psia)."

During this event, the PORVs (EIIS:AB) functioned properly to limit primary pressure to 2430 psia, so that the Pressurizer code safety valves (EIIS:AB) were not challenged. The SG code safeties (EIIS:SB) limited SG pressure to 1023 psia and SBCS functioned as designed. This event is less limiting than that described in UFSAR section 15.2.7. The health and safety of the public were not affected by this event.

CORRECTIVE ACTIONS

1) Operations personnel involved with this event were counseled on the importance of applying self-checking principles.

2) The surveillance procedure for conducting this test, OP 1/2-0030150, "Secondary Plant Operating Checks and Tests" will be changed to incorporate format improvements, and to include additional verification that critical steps have been completed,

3) Plant Staff will review other load threatening surveillances to determine if additional procedural changes or precautions are necessary to minimize the potential for personnel error.

4) A technical subcommittee was formed to evaluate this event for generic implications and provide additional corrective actions to prevent reoccurrence.

5) Site management held a trip review meeting, attended by personnel from Operations, Maintenance, Training, Engineering, Technical staff, and senior Nuclear Division management to examine this event.

The meeting was video taped to assure that lessons learned are available to all Operations personnel.

6) Instrument and Control (I/C) and System Engineers calibrated the 1A Main Feedwater Regulating Valve E/P transducer prior to unit startup. The Main Feedwater RegUlating valve positioning components affecting this event are being evaluated for additional corrective actions.

7) This Event will be included into Operations training for both licensed and non-licensed Operations personnel.

HRC- FORH 366A (5-92)

NRC FORM 366A U.S. NUCLENI REGULATORY CQHIISSIQI APPROVED BY QS NO. 3150-0104 5-92) EXPIRES 5/31/95 ESTINATED BURDEN PER RESPONSE TO CQ(PLY MITH THIS IN FORHAT10N COLLECTION REQUEST: 50.0 KRS.

FORMARD CQINENTS REGARDING BURDEN ESTIMATE TO THE LICENSEE EVENT REPORT (LER) INFORMATION AMD RECORDS MANAGEMENT BRANCH (HMBB 7714), U.S. HUCLEAR REGULATORY CQBIISSION, llASHINGTON, DC 20555-0001~ AND TO THE PAPERMORk TEXT CONTINUATION REDUCTION PROJECT (3140-0104) ~ OFFICE OF HANAGEHEMT AMD BIIGET, MASHINGTON, DC 20503.

FACILITY NU% 1 DO(XET NMBER 2 LER NNBER 6 PAGE 3 YEAR SEQUENTIAL REVISIOH St. Lucie Unit 1 05000335 95 pp3 p 5 OF 5 EXT If mor s ce is ired use edditione co ies of MRC Form 366A (17)

ADDITIONAL NPORMATION il

' n I nifi ti No component failures were identified for this event.

Pr vi mil rEv n LER 389/86-002 describes a Reactor trip initiated by loss of load during Turbine overspeed testing due to cognitive personnel error.

HRC FORM