ML20046D403

Monthly Operating Rept for Jul 1993 for Pilgrim Nuclear Power Station.W/930813 Ltr

ML20046D403
Person / Time
Site:	Pilgrim
Issue date:	07/31/1993
From:	Kraft E, Munro W BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BECO-LTR-93-098, BECO-LTR-93-98, NUDOCS 9308190183
Download: ML20046D403 (8)

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. $II BOSTON EDISON Piigrim Nuclear Po*u Station Rocky Hill Road Plymouth, Massachusetts 02300 E. S. Kraft, Jr.

Vice Presdent Nuclear Operations ano stanon onecto' August 13, 1993 BEco Ltr. #93-098 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 License No. DPR-35 Docket No. 50-29:-

July 1993 Monthly Report In accordance with PNPS Technical Specification 6.9. A.2, a copy of the Operational Status Summary for Pilgrim Nuclear Power Station is attached for your information and planning. Should you have any questions concerning this report please contact me directly.

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E. S. Kra Jr ) .

RAG /bal Attachment cc: Mr. Thomas T. Martin Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. R. B. Eaton Div. of Reactor Projects I/II Office of NRR - USNRC One White Flint North - Mail Stop 14D1 11555 Rockville' Pike Rockville, MD 20852 Senior Resident Inspector l l 9308190163 930731 3 J l

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OPERATING DATA REPORT -

i DOCKET NO. 50-293 f DATE August 13, 1993 )

COMPLE72D BY: W. Munro )

TELEPHONE (508) 747-8047 OPERATING STATUS NOTES I

1. Unit Name Pilgrim I i

.2. Reporting Period July 1993 )

3. Licensed Thermal Power (MWt) 12.2.E j

4. Nameplate Rating (Gross MWe) 678  !

5. Design Electrical Rating (Net MWe) SEE

6. Maximum Dependable Capacity (Gross MWe) EEE  :

7. Maximum Dependable Capacity (Net MWe) E7Q

8. If Changes Occur in Capacity Ratings (Item Number 3 Through 7) Since Last Report, Give Reasons: .

None .)

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9. Power Level To Which Restricted, If Any (Net MWe): None l

-10. Reasons For Restrictions, If Any: N/A j 1

This Month Yr-to-Date Cumulative i

11. Hours In Reporting Period 744.0 5087.0 180959.0

12. Number of Hours Reactor Was Critical 661.0 3551.8 109410.4 ,

13. Reactor Reserve Shutdown Hours 0.0 0.0 -0.0 l

14. Hours Generator On-Line 636.6 3439.8 105346.2

.15. Unit Reserve Shutdown Hours 0.0 0.0 0.0

-16 Gross Thermal Energy Generated (MWH) 1219200.0 6506472.0 184515288.0  ;

17. Gross Electrical Energy Generated (MWH) 416420.0 2236750.'O 62370744.0 .[

18. Net Electrical Energy Generated (MWH) 400422.0 2150867.0 599437S),Q

19. Unit Service Factor 85.6 67.6 58.2  ;

20. Unit Availablility Factor 85.6 67.6 58.2 l 21.. Unit Capacity Factor (Using MDC Net) 80.3 63.1 49.4 i

22. Unit Capacity Factor (Using DER Net) 82.2 64.6 50.6 .

23. Unit Forced Outage Rate 0.0 4.5 12.1 )

24. Shutdown Scheduled Over Next 6 Months (Type, Date, and Duration of Each): None

25. If Shut Down At End of Report Period,' Estimated Date of Startup Unit Doeratino.  ;

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-293 UNIT Pilcrim I DATE Aucust 13. 1993 COMPLETED BY: W. Munro TELEPHONE (508) 747-8047 MONTH July 1993 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe Net) (MWe-Net) 1 655 17 598 2 659 18 593 3 659 19 656 4 661 20 659 5 661 21 659 6 660 22 37 7 657 23 0 8 660 24 0 9 659 25 0 i

10 659 26 34 11 657 27 465 ,

12 658 28 661 i i

13 657 29 662 14 651 30 611 15 651 31 534 16 650 This fonnat lists the average daily unit power level in MWe-Net for each day in the reporting month, computed to the nearest whole megawatt.

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. BOSTON EDISON COMPANY I

, PILGRIM NUCLEAR POWER STATION  ;

DOCKET NO. 50-293 ,

OPERATIONAL SUFFARY FOR JULY 1993 i

The unit started the reporting period at approximately 100 percent core thermal  ;

power (CTP) where it was essentially maintained until 7-17-93 when power was reduced j to approximately 50 percent CTP to facilitate a backwash of the main condenser.  ;

Following the backwash, reactor power was increased to 100 percent on 7-18-93. l Reactor power was maintained at 100 percent CTP until 7-22-93 when a manual reactor j shutdown was initiated to investigate unidentified leakage in the drywell. At 0352 l hours on 7-22-93 the turbine generator was manually tripped off line. While repairs j were ongoing to repair a leaking weld in an undervessel drain line, installation of l the NRC mandated reactor water level instrumentation modification was also  ;

completed. The unit was made critical at 2310 hours0.0267 days <br />0.642 hours <br />0.00382 weeks <br />8.78955e-4 months <br /> on 7-25-93 and power was >

increased until reactor pressure was at 1000 psi. Final acceptance testing of the l reactor level modification was completed and the unit was synchronized to the grid  ;

at 1518 hours0.0176 days <br />0.422 hours <br />0.00251 weeks <br />5.77599e-4 months <br /> on 7-26-93. On 7-27-93 reactor power was reduced to approximately 65 [

percent CTP after experiencing "A* Feedwater Regulation Valve oscillations. i Following adjustments to the booster relay, reactor power was increased and 100 percent CTP was achieved on 7-28-93 at 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br />. Power was maintained at approximately 100 percent CTP until 2020 hours0.0234 days <br />0.561 hours <br />0.00334 weeks <br />7.6861e-4 months <br /> on 7-30-93 when a power reduction to '

50 percent CTP commenced to perf onm maintenance in the condenser bay, and facilitate  ;

a main condenser backwash. Following condenser bay maintenance and the backwash, }

reactor power was increased and at 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> on 7-31-93 the unit attained 100 f percent CTP where it was maintained for the remainder of the reporting period. j f

f SAFETY RELIEF VALVE CHALLENGES $

MONTH OF JULY 1993 j Requirement: NUREG-0737 T.A.P. II.K.3.3 .

i There were no safety relief valve challenges during the reporting period.  ;

I An SRV challenge is defined as anytime an SRV has received a signal to operate via l reactor pressure, auto signal (ADS) or control switch (manual). Ref. BECo ltr. #81- l 01 date 01/05/81.

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REFUELING IFFORMATION "The foll'owing refueling information is included in the Monthly Report as requested  !

in an NRC letter to BECo, dated January 18, 1978:  !

For your convenience, the information supplied has been enumerated so that each number corresponds to equivalent notation utilized in the request.  !

1. The name of this facility is Pilgrim Nuclear Power Station, Docket Number 50-293.

2. Scheduled date for next refueling shutdown: April 1, 1995

3. Scheduled date for restart following next refueling: June 5, 1995 ,

4. Due to their similarity, requests 4, 5, & 6 are responded to collectively under #6.

5. See #6.  !

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6. The new fuel loaded during the 1993 refueling outage was of the same design as loaded in the previous refueling outage and consisted of 140 assemblies.

7. (a) There are 580 fuel assemblies in the core.

(b) There are 1629 fuel assemblies in the spent fuel pool.

8. (a) The station is presently licensed to store 2320 spent fuel assemblies.

The actual usable spent fuel storage capacity is 2320 fuel assemblies. >

(b) The planned spent fuel storage capacity is 2320 fuel assemblies.

9. With present spent fuel in storage, the spent fuel pool now has the capacity  !

to accommodate an additional 691 fuel assemblies.

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4 MOWrH 3ULY. 1993 ,,

PILGRIM NUCLEAR POWER STATION MAJOR SAFETY RELATED MAINTENANCE

• CORRECTIVE ACTION TO ASSOCIATED-SYSTEM' COMPONENT MALFUNCTION CAUSE MAINTENANCE PREVEttr RECURRENCE LER, High-- Pressure Alarm function Missing jumper Installed (Refer to associated 93-016 00 Pressure Switch- inoperable in alann panel . jumper. LER) sto be issued)

Coolant 9090/ Alarm during Performed Injection performance monthly (HPCI) of monthly surveillance System surveillance satisfactorily.

Procedure 8.M.2-2.7 l (PR 93.9331)

High HPCI full HPCI' System Foreign HPCI full (Refer to asscciated 93-015-00:

Pressure flow failed monthly materials- flow restric- LER)

Coolant restricting- operability plugged a ring orifice Injection orifice test. System portion of the eas removed, (HPCI) declared restricting cleaned and System inoperable due orifice reinstalled.

to low flow in the full Performed rate. flow test monthly (PR93.9308) pipeline. surveillance satisfactorily.

Reac'Jor RCIC Primary Initial failed Missing (Refer to associated 93-007-01

. Core Turbine Containmc:? attempts to open electrical LER) (update of Isolation Steam- Isolation MO-1301-16 were jumper was root cause &

Cooling Supply Control System caused by a installed and corrective (RCIC) Valve Gp 5 isola- missing MO-1301-16 was actions)-

System MO-1301-16 tion due.to electrical jumper satisfactorily high steam flow that bypasses tested.

while attemp- .the torque switch ting to open in the opening RCIC turbine circuit.

steam flow supply valve, MO-1301-16 (PR-93.9094)

(PR-93.9213)

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MONTH JULY. 1993 .

PIIGRIM NUCLEAR POWER STATION ,

i MAJOR SAFETY RELATED MAINTENANCE CORRECTIVE ACTION TO ASSOCIATED .

MAINTENANCE PREVENT RECURRENCE LER '

SYSTEM COMPONEffP MALFUNCTION CAUSE Weld repaired. N/A 93-018-00 Reactor Valve Weld leak Insufficient HO-261-65 upstream of fusion during Liquid penetrant (to be issued)

Water '

examination and 4

2 Cleanup (weld valve HO-261-65 welding process.

12-BC-15) causing hydrostatic test i (RWCU) were successfully System leakage into drywell sumps. performed. >

(PR93.9335)

High HPCI eteam HPCI steam Surveillance Instrumenta Revise Procedure 93-017-00 supp.ly val- valves isolated procedure tion and Cont- 8.M.2-1.5.10 to (to be issued.

Pressure reflect implementa-2301-4 and during not revised to rol personnel Coolant tion of Plant Design Injection 2301-5. performance of reflect RFO-9 secured the surveillance design change test. Opera- Change PDC 91-75.

(HPCI) procedure (PDC 91-75) tions personnel j System 8.M.2+1.5.10, reset the "HPCI Vacuum isolation and Breaker Valve reopened valves Testing" 2301-4 and 2301-5.

(PR93.9332)

Reactor High Closed the Refer to associated 93-019-00 Residual RHR Group 3 Throttling isolation while Pressure 1000-28A LER (to be issued .

Heat Removal Valve performing Switches valve and reset MO-1001-28A Procedure PS-261-23A !. B the isolation.

(EHR)

System 2.2.19 actuated due to Performed Proc-

" Residual Heat a pressure edure 2.2.19 Removal". transient successfully.

1 (PR93.9334) - resulting from throttling the 1001-28A valve.

Dissolved non- Implemented Reactor Water Level N/A Instr- Reactor Periodic reac-tor water level condensable gases Plant Design System reliability ment water level in the reactor Change 93-24, has been enhanced by and reference spiking during pressure vessel Reactor Water addition'of Plant

' Control legs. controlled Design Change 93-24.

System shutdowns. (RPV) water level Level Refer-reference leg ence Leg Back-coming out of fill System.

solution during i

RPV depressurization.

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UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO: }0-293 '.

DOCKET NO: 50-293 .

NAME: Pilorim I ,

DATE: Aucunt 11. 1993 J COMPLETED BY: W. Munro

• TELEPHONE: (508) 747-8047 REPORT MONTH July 1993 r METHOD OF LICENSE CAUSE & CORRECTIVE DURATION SHUTTING EVENT SYSTEM COMPONENT ACTION TO PREVENT NO. DATE TYPE (HOURS) REASON DOWN REACTOR REPORT # CODE CODE RECURRENCE 1 2 3 4 5 1

07 7/22/93 'S 107.4 B 1 93-018-00 IK PSP Repair of 2 inch undervessel drain line weld which ,

was causing leakage into the drywell. ,

- Implementation of of Plant Design Change (PDC 93-24)

" Reactor Water Level Reference Leg Back-fill System."

1 2 2 3 4&5 F-FORCED A-Equip-Failure F-Admin 1-Manual Exhibit F & H S-SCHED B-Main or Test G-Oper Error 2-Manual Scram Instructions for C-Refueling .

H-Other- 3-Auto Scram Preparations of

'D-Regulatory Restriction 4-Continued Data Entry Sheet E Operator Training 5-Reduced Load Licensee Event Report

& License Examination 9-Other (LER) File (NUREG-1022)

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