ML17228B282

LER 95-003-01:on 950708,automatic Reactor Trip Occurred During Overspeed Surveillance Testing Due to Personnel Error.Counselled Operations Personnel Involved W/Event & Made Procedure changes.W/950923 Ltr

ML17228B282
Person / Time
Site:	Saint Lucie
Issue date:	09/23/1995
From:	Benken E, Sager D FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-95-269, LER-95-003, LER-95-3, NUDOCS 9510020133
Download: ML17228B282 (8)

Text

PRIORITY (ACCELERATED RIDS PROCESSING)

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9510020133 DOC.DATE: 95/09/23 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 95-003-01:on 950708,automatic reactor trip occurred during overspeed surveillance testing due to personnel error. Counselled Operations personnel involved w/event &

made procedure changes.W/950923 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 PD2-1 PD 1 1 NORRIS,J 1 1 INTERNAL: ACRS 1 1 REOD/SPD B 2 2 AEOD/S PD/RRAB 1 1 LE CENTER+ 1 1 NRR/DE/ECGB 1 1 NRR/DE/EEL'B 1 1 NRR/DE/EMEB 1 1 NRR/DISP/PIPB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRPM/PECB 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 POORE,W. 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 NOTE TO ALL "RIDS" RECIP1ENTS:

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

Fiorida Power & Light Company, P.O. Box 128, Fort Pierce, FL 34954.0128 September 23, 1995 L-95-269 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 Revision 1 Date of Event: July 8, 1995 utomatic Reactor Tri Durin Turbine Overs eed Surveillance Testin 'due to Personnel Error The attached Licensee Event Report is being submitted pursuant to the requirements of 10 CFR 50.73 to provide an update on the subject event.

Very truly yours, JAB' D. A. er Vice sident St. Lucre Plant DAS/EJB Attachment cc: Stewart D. Ebneter, Regional Administrator, USNRC Region II Senior Resident Xnspector, USNRC, St. Lucie Plant

/ 4 ~

9510020133 950923 PDR ADOCK 05000335 8 PDR an FPL Group company

NRC FORM 366 S NUCLEAR REGULATORY C(NMISS I ON PROVED BY (NB KO. 3150-0104 (5-92) EXPIRES 05/31/95 ESTIMATED BURDEH PER RESPONSE TO COMPLY KITH LICENSEE EVENT REPORT (LER) THIS INFORHATIOM COLLECTIOH REQUEST: 50.0 HRS.

FORNARD COMHENTS REGARDING BURDEH ESTIMATE TO THE IHFORMATIOH AMD RECORDS MAMAGEMEHT BRANCH (HMBB 7714), U.S. NUCLEAR REGULATORY COMHISSIOM ~

(See reverse for required nwher of digits/characters for each block) MASHIMGTOH, DC 20555-0001 AND TO THE PAPERIJORK REDUCTION PROJECT (31(0-0104), OFFICE OF MANAGEHENT AMD BUDGET NASKINGTOM DC 20503.

FACILITY NAHE (1) DOCKET NNBER (2) PAGE (3)

St. Lucie Unit 1 05000335 1 OF 5 TITLE (4) Automatic Reactor Trip During Turbine Overspeed Surveillance Testing due to Personnel Error.

EYEMT DATE 5 LER NNBER 6 REPORT DATE 7 OTHER FACILITIES INVOLVED 8 SEQUENTIAL REVISIOM FACILITY NAME DOCKET KLNBER MONTH DAY YEAR YEAR HOKTH DAY YEAR NUMBER KUMBER N/A FACILITY NAME 07 08 95 95 003 09 23 DOCKET KUMBER N/A OPERATING THIS REPORT IS SUBMITTED PURSUAMT TO THE REQUIREMENTS OF 10 CFR  : Check one or mor e 11 IKX)E (9) 20.402(b) 20.405(c) 50.73(e)(2)(iv) 73.71(b)

POKER 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 NAME Edwin J. Benken, Licensing Engineer TELEPHONE NUHBER (Include Area Code)

(407) 468-4248 C(NPLETE ONE LIME FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 13 CAUSE SYSTEH REPORTABLE REPORTABLE COMPONENT HANUFACTURER CAUSE SYSTEH COMPONENT MANUFACTURER TO MPRDS TO MPRDS RV D243 SUPPLEMENTAL REPORT EXPECTED 14 MONTH DAY YEAR EXPECTED YES SUBHI S SION (If yes, coaplete EXPECTED SUBMISSIOK DAlE). X KO DATE. (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 s'ingle-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.

MRC FORM ( )

RRC FORH 366A U.S NUCLEAR REGULATORY CQNISSION APPROVED BY QQ NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COHPLY IIITH THIS INFORHAT IOM COLLECTION REOUEST: 50.0 HRS.

FORNARD COMMENTS REGARDING BURDEN EST IHATE TO THE LICENSEE EVENT REPORT (LER) INFORHATIOM AMD RECORDS HANAGEHENT BRANCH (HMBB TEXT CONTINUATION 7714), UoS. HUCLEAR REGULATORY COHHISSION, NASHINGTON, DC 20555-0001 AND TO THE PAPERMORK REDUCI'ION PROJECT (3130-0104), OFFICE OF HAMAGEHENT AMD BUDGET MASHINGTON DC 20503.

FACILITY NAME 1 DOCKET NNBER 2 LER NWBER 6 PAGE 3 YEAR SEQUENTIAL REVISION St. Lucie Unit 1 20F 05000335 95 003 5 TEXl'f more s ce is r uired use additional co ies of NRC Form 366A ('17)

DESCRIPTION OF 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 test 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 />. The maximum RCS pressure reached during this event was approximately 2430 psia. The maximum secondary pressure reached was approximately 1023 psia.

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

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.

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.

HRC- FORH 366A (5 92)

QRC FORM 366A U.S. NUCLEAR REGULATORY CONIISSION APPROVED BZ QQ NO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPOHSE TO COMPLY NITH THIS INFORMATION COLLECT ION REQUEST: 50.0 HRS.

FORIIARD COMMENTS REGARDING BURDEN ESTIMATE TO THE LICENSEE EVENT REPORT (LER) INFORMATION AND RECORDS MANAGEMEHT BRANCH (HNBB TEXT CONTINUATION 7714), U.S. NUCLEAR REGULATORY COMMISSIOH, NASHIHGTON, DC 20555-0001 AND TO THE PAPERNORK REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGET UASHINGTOH DC 20503.

FACILITY NAME 1 DOCKET NNBER 2 LER NOSER 6 PAGE 3 SEOUENT I AL REVISION YEAR St. Lucia Unit 1 05000335 95 003 3 OF 5 TEXT If more s ace is r uired use additionat co ies of NRC Form 366A (17)

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.

ANALYSIS 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. Lucia 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."

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)."

HRC- FORH 366A (5-92)

IRC FORM 366A U.S. NUCLEAR REGULATORY CQIIISSION APPROVED BY QHI HO. 3150-0104 (5-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY MITH THIS INFORMATION COLLECT IOH REQUEST: 50.0 HRS.

FORHARD COMMENTS REGARDIHG BURDEN ESTIMATE TO THE LICENSEE EVENT REPORT (LER) INFORMATION AHD RECORDS MANAGEMENT BRANCH (HHBB TEXT CONTINUATION 7?14), U.S. NUCLEAR REGULATORY COMMISSION, l!ASHIHGTOH, DC 20555-0001 AHD TO THE PAPERNORK REDUCTIOH PROJECT (3140-0104), OFFICE OF MANAGEMENT AHD BUDGET IIASHIHGTON DC 20503.

FACILITY NAME 1 DOCKET HINBER 2 LER HIHIBER 6 PAGE 3

'EQUENT I AL REVISION St. Lucie Unit 1 05000335 003 4OF 5 TEXT If more s ce is r ired Use additional co ies of HRC Form 366A (17)

ANALYSIS OP THE EVENT continued During this event, the maximum primary pressure reached was approximately 2430 psia, which is below the Pressurizer code safety valve setpoint of 2500 psia. During the initial review of plant data from the reactor trip, it was determined that the PORVs (EIIS:AB) functioned properly to limit primary pressure. This conclusion was based on a review of the primary pressure response and PORV acoustic flow data, and supported by a Quench Tank pressure increase seen during the transient. In August of 1995, the PORV main valves were found to be inoperable due to improper mechanical assembly (Reference LER 335-95-005-00). Subsequent testing, inspection and analysis showed that the PORV main valves most probably did not open during this event. The (SG) code safeties (EIIS:SB) operated to limit SG pressure to 1023 psia and the SBCS functioned as designed. These systems, in conjunction with the Reactor trip, functioned to limit primary system pressure. 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.

t

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- FORM 366A (5-92)

IiRC FORM 366A .S. NUCLEAR RECULATORY CQIIISSION PROVED BY 0$ NO. 3150-0104 (5-92) EXPIRES 5/31/95, ESTIMATED BURDEN PER RESPONSE TO COMPLY MITH THIS INFORMATION COLLECTION REQUEST: 50 ~ 0 HRS.

FORMARD COMMENTS REGARDING BURDEH ESTIMATE TO THE LICENSEE EVENT REPORT (LER) IHFORMATION AND RECORDS MANAGEMENT BRANCH (MHBB 7714), U.S. NUCLEAR REGULATORY COMMISSIOH, MASHIHGTOH, DC 20555-0001 AND TO THE PAPERMORK TEXT CONTINUATION REDUCTION PROJECT (3140-0104), OFFICE OF MANAGEMENT AND BUDGETS MASHINGTOH, DC 20503.

FACILITY NAME 1 DOCKET klNBER 2 LER NNBER 6 PACE 3 YEAR SEQUENT IAL REVISION St. Lucie Unit 1 05000335 95 003 5OF5 TEXT If more s ce is r uired use edditionsI co ies of NRC Form 366A (17)

ADDITIONAL INFORMATION F ile m onent Iden ifi i n Manufacturer: Dresser Ind. Vlv 5 Inst Div /Ashcroft Model Number: 31533VX-30 Device: Pressurizer PORV valve Pr vi u imil r Even s LER 389/86-002 describes a Reactor trip initiated by loss of load during Turbine overspeed testing due to cognitive personnel error.

'I NRC- FORM A ( )