Text

Monthly Operating Rept for Dec 1982
	ML20064G435
Person / Time
Site:	LaSalle
Issue date:	01/06/1983
From:	Ullrich J; COMMONWEALTH EDISON CO.
To:	;
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ML20064G433	List: ML20064G430; ML20064G435;
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	NUDOCS 8301110496
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, LASALLE NUCLEAR POWER STATION l

UNIT 1 HONTHLY PERFORMANCE REPORT DECEMBER, 1982 4

COMMONWEALTH EDISON COMPANY -

I NRC DOCKET NO. 050-373 LICENSE NO. NPF-11 1

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a TABLE OF CONTENTS

1. INTRODUCTI ON ll.

SUMMARY

OF OPERATING EXPERIENCE FOR UNIT ONE Ill. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A. Amendments to Facility License or Technical Specifications B. Facility or Procedure Changes Requiring NRC Approval C. Tests and Experiments Requiring NRC Approval D. Corrective Maintenance of Safety Related Equipnent IV. LICENSEE EVENT REPORTS V. DATA TABULATIONS A. Operating Data Report B. Average Daily Unit Power Level C. Unit Shutdowns and Power Reductions VI. UNIQUE REPORTING REQUIREMENTS A. Main Steani Relief Valve Operations B. ECCS System Outages C. Off-Site Dose Calculation Manual Changes D. Major Changes to Radioactive Waste Treatment System E. Changes to the Process Control Program l

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1. INTRODUCT I ON The LaSalle Nuclear Power Station Unit One is a Boiling Water Reactor with a designed electrical output of 1078 MWe net, located in Marseilles, Illinois. The Station is owned by Commonwealth Edison Company. The Architect / Engineer was Sargent & Lundy, and the primary construction contractor was Commonwealth Edison Company.

The condenser cooling method is a closed cycle cooling pond.

The plant is subject to License Number NPF-11, issued on April 17, 1982. The date of initial criticality was June 21, 1982. The unit has not commenced commercial generation of power.

This report was compiled by John Ullrich, telephone number (815)357-6761, extension 481.

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II.

SUMMARY

OF UNIT OPERATING EXPERIENCE FOR UNIT ONE December 1 The unit started the reporting period at approximately 46% Reactor Power. At 1246 the reactor scrammed on low water. level due to EAP on the "A" TDRFP running low. The reactor was critical for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 46 minutes on December 1.

December 2-18 The reactor went critical at 1044 hours0.0121 days 0.29 hours 0.00173 weeks 3.97242e-4 months on December 2, 1982. At 1905 the main generator was synchronized to the grid and power increased to 20%. At 2234 the reactor scrammed for STP-31 " Loss of Offsite Power Test." This outage was extended due to the "A" RR Pump Discharge Valve sticking closed.

December 19-31 The reactor went critical, with the "A" Recirculation Loop out of service, at 1420 hours0.0164 days 0.394 hours 0.00235 weeks 5.4031e-4 months on December 19, 1982. At 1730 on December 20 the main generator was synchronized to the grid and load increased to 170 MWe, At 2125 on December 22, load was increased to 435 MWe, At 2045 on December 30 a normal unit ' shutdown was commenced due to an " Unusual Event" of the "B" RHR Pump being inoperable. At 1440 on December 31, 1982, the reactor was shutdown.

Ill. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS AND SAFETY RELATED MAINTEN ANCE A. Amendments to Facility License or Technical Specifications.

AMENDMENT 10 This Amendment consists of a deletion of License Condition 2.C. (17) and changes the Technical Specifications as a result of hardware modifications to satisfy License Condition 2.C. (17).

AMENDMENT 11 This Amendment authorizes the operation of LaSalle County Station, Unit 1 with one recirculation loop out of service for the first fuel cycle only. Also, the Amendment adds a license condition limiting a power level to 50 percent of full power or less during the operation with one recirculation loop out of service.

AMENDMENT 12 The regulation 10CFR 50.44 requires that all reactors with Mark i or Mark 11 containment must be inerted 6 months after initial criticality. LaSalle Unit I has a Mark ll containment and its initial criticality was performed on June 21, 1982. It is presently completing its startup program and Commonwealth Edison requested an exemption from this regulation to avoid any potential negative impact on the startup program. The exemption is authorized until either the required 100 percent rated thermal power trip test has been performed or the reactor has operated for 120 effective full power days, whichever is first.

Ill. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS AND SAFETY RELATED MAINTENANCE (Continued)

B. Facility or Procedure Changes Requiring NRC Approval.

Modification M-1-82-049 - This modification changed the permissive for opening the LPCI and LPCS injection valves to the condition that reactor pressure is less than 500 psig. The interlock prevents the opening of the injection valves when reactor pressure is above the design value of the low pressure piping.

Installation and testing of this modification was completed on 12/13/82.

C. Tests and Experiments Requiring NRC Approval.

There were no tests or experiments requiring NRC approval during the reporting period.

D. Corrective Maintenance of Safety Related Equipment.

The following tables present a summary of safety-related maintenance completed on Unit One during the reported period. -

The headings indicated in this summary include: Work Request Numbers, LER Numbers, Component Name, Cause of Malfunctions, Results and Ef fects on Safe Operation, and Corrective Action.

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LER C0t1PONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION VCRK REQUEST ON SAFE OPERATIO!!

1TR-Cr.037, Chart Lcw speed chart drive Drive gear j amming & Replaced motor L17963 --

Recorder .nctor broken chart won't advance Tornado Da ter Da per in wrong position I croper da per indication Moved da per for L18L37 --

proper indication L13L33 --

1 A R:iR Testable Valve 1e'a'ks Valve leaks Repacked valve Check Valve assembly Alarm frequently comes Installed new relay L18611 , ,

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up without actually losinc

' . continuity to squib

APRM Sad: ,eter A?RM readings on back Replaced reter L13333 --

panel do not agree with APRM readings at corputer program Main Steam Lir.e Sad nctor - Valve will not open from Replaced motor, L19335 --

Outboard Drain Valve the control roon

] reset limits L19624 -- Suppression Pcc1 Low speed chart drive Chart does not d ri ve on Replaced motor, Te perature Recorder notor broken low speed ,

L19737 --

Recorder 1321-R614 Recorder out-of- Tailpipa recoeratures Calibrated recorder calibraticn appear low Ccctrol Roon Ener- Da-per has a bent shaft Da per broken Replaced shaft and L19737 32-151/03L-0 gency Makeup Train bearings Outlet Da per Valve 1321-F016 Torque switch contacts Valve will not seal in Cleaned torque L206L7 32-153/03L-0 d i rt y closed switch contacts L20715 -, Control Roon A/C Brcken wi re on audible Audible alarm doesn't Repaired wire Sys t em S OA Amcm i a alarn switch work Detector L20713 --

Control Room A/C Locse wire Flow light not lit Reconnected wire Syste : B OA A ron i a Detector

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IV. . LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for LaSalle Nuclear Power Station, Unit One, occurring during the reporting period, December 1. to December.31,1982. This information is provided pursuant to the reportable occurrence reporting requirements as set forth in section 6.6.B.1 and 6.6.B.2 of the Technical Specifications.

Licensee Event Report Number Date Title of Occurrence 82-166/03L-0 11/26/82 Simultaneous Failure of IPL15J

& IPL75J 82-167/03L-0 12/1/82 Division 11 Post LOCA Honitor 82-168/03L-0 12/1/82 RCIC Steam Line High Flow / Low Pressure isolation Instrument Sensing Line Leak 82-169/03L-0 12/1/82 Administrative Control to LAP 240-1 82-170/03L-0 12/14/82 2A D/G Room Temperature Below Tech. Spec. Limit 82-171/03L-0 12/19/82 Rod Worth Minimizer 82-172/03L-0 12/20/82 Loss of Division ll Level Instrumentation 82-173/03L-0 12/21/82 Failure of Containment isolation Valve to Close 82-174/03L-0 12/19/92 lRM B Detector Cable Unplugged 82-175/03L-0 12/24/82 Drywell Vacuum Breaker inoperable Position Indication 82-176/03L-0 12/28/82 "B" RHR Pump High Vibration 82-177/03L-0 12/30/82 1E12-F024A Failed to Close 82-178/03L-0 12/31/82 ADS Valves Low Pressure 82-179/03L-0 12/24/82 Lake Blowdown Flow Indication

P V. DATA TABULATIONS .

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The following data tabulations are presented in this report:

A. Operating Data Report B. Average Daily Unit Power Level i

! C.- Unit Shutdowns and Power Reductions r

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-[LYP 550*7 ATTAC10fENT D R2vislo'n 2 fc November 13,.1979 J. .- : 9 >

' OPERATING DATA REPORT

- OCCKET NO. 050-373 UNIT LaSalle 1 DATE I/6/83

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COMPLETED BY John Ullrich TELEPHONE (815)357-6761x4}1 OPER ATING STATUS .

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1. REPORTING PERICO: De cembe r . 1982 GROSS. HOURS IN R EPORTING PERICO:

uax. OEpENo. CArACiTY tuWeNein: 0 l

2. CURRENTLY AUTHORIZED POWER LEVELiMWti: 50%

OESIGN ELECTRICAL MATING (MW, Net): 1078 i 0

- 3. POWER LEVEL TO WHICH RESTRICTED (!F ANY) (MWeNet):

4. REASONS FOR RESTRICTION ilF ANY): Due to single loop operation TwiS MONTH YR TO CATE ',CUMULATIV E

s. vuueER op wouRs REACTOR WAS CRmCAt. ...............

312.9 2747.4 2747.4 0 -

0 0

s. REACTOR RESERVE SHUTDOWN HOURS . . . . . . . . . . . . . . . . . . .

7. wouRs G EN ER AToR ON U N E . . . . . . . . . . . . . . . . . . . . . . . . . .

266.2 1857.6 18'57.6 0 0 0

- a. UNIT RESERVE SHUTDOWN HOURS ......................

- s. Gross THERMAL ENERGY GENER ATED (MWM) ........... ..

366497 2140579 2140579 lo. GnoSS ELECTRICAL ENERGY GENERATED IMWH) . . . . . . . . ...

84188 520399 520199 71121 460775 46077c

11. NET ELECTRICALENERG GENER ATED (MWH) ..............

NA NA

12. R EACTOR SE RvlCE F ACTOR . . . . . . . . . .. . . . . . . . . . . . . . . . . .

NA NA

13. REACTOR AVAILAstLITY F ACTOR . . . . . . . . . . . . . . . . . . . . . . .

NA NA NA

14. UNIT S E RVIC E F ACTO R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

NA NA NA

15. UNIT AV AILA tluTY F ACTOR . . . . . . . . . . . . . . . . . . . . . . . . .

- NA NA NA l

16. UNIT CAPACITY FACTOR (Using MDC) .....................

17. UNIT CAPACITY FACTOR (Uses Design MWel . . . . . . ..........

10. UNIT F O RCE D CUTAG E R AT E , . . . . . . . . . . . . . . . . . . . . . . . . .

NA NA NA 1s. . SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE DATE. AND OUR ATION OF EACHl:

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20. is SHuf ooWN Ar ENo OF REPORT PERIOD. ESTIMATE 0 0 ATE OF STARTUP: I/2k/83 6 21. UNITS IN TEST STATUS tPRIOR TO COMMERCIAL CPERATION): FORECAST ACHIEVED INmAL CRmCAuTy 6/21/82 INITI AL E LECTRICITY 9/4/82 CoMMERCI AL OPERATION h/1/81 e

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1 , November 13, 1979 l

. ATTACHMENT A AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 050-373 i l UNIT LaSalle 1 DATE 1/6/83 .

COMPLETED BY Johri Ullrich TELEPHONE (815) 357-6.761 x481 MONTH De cembe r , 1982 DAY AVERAGE DAll.Y POWER LEVEL . DAY AVERAGE DAILY POWER LEVEL (MWe-Ne t) (MWe-Ne t)

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4. -17 20.

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6. -11 22. 367 7 -14 23 389-I

8. -14 24. 397

,9 -15 25 354

10. -14 26. 280-

11. -14 27 237 1

12. -14 28. 255 l 13 -14 '29 280

, 14. -14 30. 200 1

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LTP-300-7 Revision 2 November 13, 1979 6 . ,

- f ATTACHMENT B **

UNIT SHUTDOWNS AND POWER REDUCTIONS .

DOCKET NO. 050-373

UNIT NAME Lasalle 1 December, 1982 DATE 1/6/83 l REPORT HONTH ,

COMPLETED BY _Inhn til l ri ch TELEPHONE (815)357-6761 X481 METHOD OF SHUTTING DOWN .

TYPE THE REACTOR OR F: FORCED DURATION REDUCING P0k'ER (2)- CORRECTIVE ACTIONS / COMMENTS ,

DATE S: SCHEDULED (HOURS) REASON (1)

HO.

Reactor Scram on low water p A 3 8 12/1/82 F 30.3 due to EAP on "A" TDRFP running low.

I 8 3 Reactor scram due to " Loss of i 9 12/2/82 F 427.0 Of fsite Power Test ," STP 31.

This outage was extended by i the "A" RR Pump Discharge j Valve sticking closed. p 1 Normal shutdown due to 10 12/31/82 F 20.5 D

" Unusual Event" due to "B" RHR Pump being inop. ,

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VI. UNIQUE. REPORTING REQUIREMENTS A. Main Steam. Relief Valve Operations for Unit i VALVES NO. & TYPE PLANT DATE ACTUATION- CONDITION REASON ACTUATQ 12/21/82 .IB21-F013M 2 Manual ,37% Power LST 79-1 12/21/82 1821-F013C 1 Manual 37% Power LST 79-1 12/21/82 1821-F013G 1 Manual 37% Power LST 79-1 12/21/82 IB21-F013H 1 Manual 37% Power LST 79-1 12/21/82 1821-F013R 1 Manua l 37% Power LST 79-1 12/22/82 IB21-F013E 1 Manual 43% Power LST 79-1 12/22/82 1821-F013C 4 Manual 43% Power. LST 79-1 12/22/82 1821-F013G 4 Manual 43% Power- LST 79-1 12/22/82 1821-F013H 4 Manual 43% Power LST 79-1 12/22/82 1821-F013M 2 Manual 43% Power LST 79-1 12/22/82 1821-F013R 4 Manual '43% Power LST 79-1 12/23/82 1821-F013F 2 Manual *AS% Power LST 79-1 12/23/82 1821-F013G 2 Manual ~46% Power LST 79-1 12/23/82 1821-F013H 2 Manual ~45% Power LST 79-1 12/23/82 1B21-F013M 6 Manual ~45% Power LST 79-1 12/23/82 1821-F013R 2 Manual +45% Power LST 79-1 12/24/82 1821-F013C 4 Manual ^48% Power LST 79-1 12/24/82 1821-F013G 2 Manual ~48% Power LST 79-1 12/24/82 1821-F013H 2 Manual ^48% Power LST 79-1 12/24/82 IB21-F013M 2 Manual ~48% Power LST 79-1 12/24/82 IB21-F013R 2 Manual ~48% Power LST 79-1 12/26/82 IB21-F013C 5 Manual 48% Power LST 79-1 12/26/82 1821-F013G 4 Manual 48% Power LST 79-1 12/26/82 1821-F013M 7 Manual 48% Power LST 79-1 12/26/82 1B21-F013R 1 Manual 48% Power LST 79-1 l

12/27/82 IB21-F013C 14 Manual 45% Power LST 79-1 12/27/82 1821-F013G 10 Manual 45% Power LST 79-1 12/27/82 IB21-F013H 4 Manual 45% Power LST 79-1 12/27/82 1821-F013M 3 Marual 45% Power LST 79-1 12/27/82 1821-F013R 6 Manual 45% Power LST 79-1 12/28/82 1821-F013C 5 Manual 46% Power LST 79-1 12/28/82 1821-F013G 5 Manual 46% Power LST 79-1 12/29/82 1821-F013C 5 Manual .~49% Power LST 79-1 12/29/82 1821-F013G 2 Manual ~49% Power LST 79-1 12/29/82 1821-F013H 1 Manual ~49% Power LST 79-1 12/29/82 1821-F013M 1 Manual +49% Power LST 79-1 12/29/82 1821-F013R 1 Manual ~49% Power LST 79-1

VI. UNIQUE REPORTING REQUIREMENTS (Continued)

VALVES NO. & TYPE PLANT DATE ACTUATED ACTUATI ON CONDITI ON REASON 12/30/82 1821-F013C 7 Hanual ~49% Power LST 79-1 12/30/82 1821-F013G 7 Hanual ~49% Power LST 79-1 12/30/82 IB21-F013H 7 Hanual ~49% Power LST 79-1 12/30/82 I B21 -F013R 7 Hanual #49% Power LST 79-1 12/30/82 1821 -F013H 4 Manual ~49% Power LST 79-1 B. ECCS Systems Outages Outage No. Equipnent Purpose Of Outage 1-1902-82 "B" RHR Pump To repair pump 1-1909-82 RHR Full Flow Test Valve Valve would not close C. Off-Site Dose Calculation Manual There were no changes to the Off-Site Dose Calculations Manual during this reporting period.

D. Radioactive Waste Treatment System There were no changes to the Radioactive Waste Treatment System i

during this reporting period.

E. Process Control Program The Process Control Program will be revised to change reference ASTH C150-1980 to ASTH C150. This will allow the use of the most current revision.

ML20064G435

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