Text

Monthly Operating Rept for Jan 1984
	ML20080U033
Person / Time
Site:	LaSalle
Issue date:	01/31/1984
From:	Diederich G, Lin D; COMMONWEALTH EDISON CO.
To:	; NRC OFFICE OF ADMINISTRATION (ADM), NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
	NUDOCS 8403020138
	Download: ML20080U033 (31)
	v • d • e

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ATTACHMENT C LTP-300-7 Revision 3 March 1, 1983 7

OPERATING DATA REPORT DOCKET NO. 050-373 UNIT LaSalle One DATE Feb. 15. 1984 COMPLETED BY Diane L. Lin TELEPHONE (815)357-6761 OPERATING STATUS

1. REPORTING PERIOD: January 1984 GROSS HOURS IN REPORTING PERIOD: 744

2. CURRENTLY AUTHORIZED POWER LEVEL (MWL):100% MAK DEPEND CAPACITY (MWo-Net): 1036 DESIGN ELECTRICAL RATING (MWe-Net):1078

3. POWER LEVEL TO WHICH RESTRICTED (IF ANY) (MWe-Net): _

4. REASONS FOR RESTRICTION (IP ANY):

THIS MONTH YR TO DATE CUMULATIVE *,

5 NUMBER OF HOURS REACTOR WAS CRITICAL 544.0 544.0 544.0

6. REACIOR RESERVE S51UTDOWN HOURS 182.1 182.1 182.1

7. HOURS GENERATOR ON LINE 46_7. 5 467.5 4_67_.5

8. UNIT RESERVE SHUTDOWN IIOURS 1.0 1.0 1.0

9. GROSS THERMAL ENERGY GENERATED (MWil) 1099670 1099670 1099670

10. GROSS ELEC. ENERGY GENERATED (MWH) 347622 347622 347672

11. NET ELEC. ENERGY GENERATED (MWH) 321874 327874 327874

12. REACTOR SERVICE FACTOR 13.1% 13.1% 73.1%

13. REACTOR AVAILABILITY FACTOR 97.6% 97.6% 97.6%

14. UNIT SERVICE FACTOR 62_.8% 62.8% 62. 8%.

15. UNIT AVAILABILITY FACTOR 63.0% 63.0% 63.0%

16. UNIT CAPACITY FACTOR (USING MDC) 42.5% 42.5% 42.5%

17. UNIT CAPACITY FACTOR (USING DESIGN MWe) 40.9% 40.9% 40.9%

18. UNIT FORCED OUTAGE RATE 26.6% 26.6% 2_6.6%

19. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH) 2/29/84 Leak rate testing, modifications on FW Heaters installed, condenser inspection.

20. IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:

21. UNITS IN TEST STATUS (PRIOR TO COMMERCIAL OPERATION):

FORKCAST ACHIEVED INITIAL CRITICALITY 6/21/82 INITIAL ELECTRICITY 9/04/82 COMMERCIAL OPERATION 1/1/84

Started over because LaSalle One became conumercial 1/1/84.

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LTP-300-7 Rr.visicn 3 March 1, 1983 6

ATTACHMENT B AVEMAGE DAILY UNIT POWER LEVEL DOCKET NO: 050-373 UNIT: LASALLE ONE DATE: Feb. 15, 1984 COMPLETED BY: DIANE L. LIN TELEPHONE: (815) 357-6761 MONTH January 1984 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Not) (MWe-Met)

. 1. __ , 0 17. 0

2. 43 18. 0-3 '4 19. 0

4. 0 20. 0

5. 0 21, 12

6. 98 22, 344

7. 6 23. 745

8. 503 24. 726

9. 701 25. 915

10. 732 26. -924

11. 762 27. 916

12. 809 28. 900

13. 732 29. 909

14. 8 30. 905

.15. 513 31. 920 16 . _. 453 INSTRUCTIONS On this form list the average daily unit power level in MWe-Net for each day in the reporting month. Compute'to the nearest whole megawatt.

Theso figures will be-used to plot a graph for each reporting month. Note

- that when maximum dependable capacity is used for the not electrical rating of the unit there may be occasions when the daily average power level exceeds the

100% 110e for the restricted power level line. In such cases the average daily unit power output sheet should be footnoted to explain the appacent anomaly..

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R;vicion 3 March 1, 1983 ~

9 (Final)

ATTACHMENT E UNIT SHUTDOtGIS AND POWER REDUCTIONS DOCKET NO. 050-373 UNIT NAME LaSalle One DATE Feb. 15. 1984 REPORT MONTH JANUARY 1984 COMPLETED BY Disne L. Lin TELEPHONE (815)357-6761 METHOD OF TYPE SHUTTING DOWN F: FORCED DURATION THE REACTOR OR- CORRECTIVE NO. DATE S: SCHEDULED (HOURS) REASON (1) REDUCING POWER ACTIONS /COletENTS 42 11/3/83* S 34.7 B 4 Opened oil circuit breakers 9-10 and 10-11 for STP-27.

1 1/3/84 S 72.6 B 2 Manually tripped turb-ine,then manually scrammed Ex to pre-pare for sequence change for scram time testing.

2 1/6/84 F 23.5 A- 3 Rx scram due to low Ex water level resulting 1

from the "A" TDRFP trip on high seal injection temperature T

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, LTP-3OO-7 -

, Revision 3 March 1, 1983 9 (Final)

ATTACHMENT E UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 050-373 UNIT NAME LaSalle One DATE Veb. 15. 1984 REPORT MONTH JANUARY 1984 COMPLETED BY Diane L. Lin CONTINUED TELEPHONE (815)357-6761 t

METHOD OF TYPE SHUTTING DOWN F: FORCED DURATION THE REACTOR OR. CORRECTIVE NO. DATE S: SCHEDULED (HOURS) REAMON (1) REDUCING PCMhR ACTIONS /C9MMENTS 3 1/13/84 F 22.7 A 3 Ex scram on main turbine trip due to a generator protective trip caused by the main generator pot fuse device opening up 4 1A6/84 F 123.0 A 3 Rx scram caused by steam in-pinsemer.t on the "C" Cond-i enser boot seal which fail-ed and caused a loss of con-denser vacuum.

CRafsr to the November 1983 report DOCUMENT 0044r i

VI. UNIQUE REPORTING REQUIREMENTS A. Main Steam Relief Valve Operations for Unit 1.

Relief valve operaticas during the reporting period are summaelzed in the following Table. The table included information as to which relief valve was actuated, how it was activated and the circumstances resulting in its actuation.

Valves No & Type Plant Description Date Actuated Actuations Conditions of Events 1/1/84 1821-F013U 5 Manual O psig LOS-MS-E2

< 1/16/84 1821-F013E 1 Manual 1000 psis To control pressure after scram .

1/16/84 1821-F013E 1 Manual 963 psig To control pressure after scram 1/16/84 1821-F013F- 1' Manual 990 psig To control pressure after

- scram 1/16/84 1821-F013F 1 Manual 963 psig To control pressure after scram B. ECCS Systems Outages There were no ECCS System outages during this reporting period.

C. Off-site Dose Calculation Manual The following changes were made to the Off-Sita Dose Calcualation Manual.

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1. Pane 2.1-17. Section 2.1.2.2 The adult is considered the limiting person and the 10 cfr 20 limits are based on the inhalation pathway.

2. Pane 2.1-18. Equation 2.18 The seasonal adjustment factor, K, is set equal to zero because compliance is demonstrated for the inhalation pathway and does not include the milk pathway.

3. Page 2.2-4 The change in the definition of MPC was maae to address unknown nuclides as well as known nuclides

4. Pane 4-1 Corrected page number.

5. Page 4.1-1 Additional clarification added to Table 7.2-1 is the Aquatic Environment Dose Parameters.

6. Page 4.1-3 Corrected page numbsr.

7. Page 8.1-4 Section 8.1.3 Changes made to describe the mid and high range detector systems on the Plant Vent WRGM.

8. Pages 8.1-5 and 8.1-6. Section 8.1.4 Changes made to describe the mid and high range detector systems on the SGTS WRGM.

The changes shown in Revision 5. February 1983, and Revision 10, Sootember 1983 are changes in the administration of the program. The accuracy or reliability of dose calculations and setpoints are not reduced. The subject changes were reviewed and found acceptat'le by the Station Onsite Review and Investigative Furetion. The '

-Onsite Review was OSR83-64 and dated November 15, 1983.

D. Radioactive Waste Treatment System l

There were no changes to the Radioactive Waste Treatment System l during this reporting period.

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Revision E Dati: Sept. 8,1980 Page 4 ATTACNMENT A.

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.LASALLE COUNTT STATION ASSIGNMG T OF SU8 JECT,FOR ,

l QMSITE. REVIEW ANO. INVESTIGNf10N l

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ASS 1.GNMENT REVIEW NUMBER E 3-lo Y d .

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NAME-WISC IPt.lNE DATE FIN 0lNGS AND RECOMMENOATIONS ARE TO SE REPORTED:

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]f Tu DISCIPLINES: ,

NPPT - Nuctsar Power Plant Technology R0 .9eactor Operations "

RE - Reactor Engineering

- Radiation Protection and Control

R&C ist - Instrumentation and Control MsES - Mechanical and Electrical Systems

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REVISION 5

,,,j FEBRUARY 19 8 3 Fy Heat Fraction (days /kg)

The fraction of the animal's daily intade of radionu.clide i which appears in each' kilogram of flesh. See Table 7.1-4.

t, Slaughter to consumption Time (hr)

The time from slaughter consumption, see Table 7.1-2.

2.1.2.1.2 Inhalation + Food Pathways Dose, Calendar Year (Pour Consecutive Quarters) 3.17 x 10-8 x 10 6 R, DFA g$, /

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l 2.1.2.2 10 CFR 20 Release Rate Limit The maximum dose rate to an organ of an adult from all radio- l nuclides, radioactive materials in particulate form, and radio-nuclides other-than noble gases with half-lives greater than 8 days shall be limited to the values given by the equations l which follow. For purposes of demonstrating compliance with the Technical Specifications, the dose to the adult fron the inhalation pathway shall be considered limiting. i l

l 2.1-17 l

REVISION 5 FEBRUARY 1983 6 + +

10 R, DFA g$3 ( X/Q) , Qis + ( VQ) y Q gy (WQ)g O gg i .

,K DFI t$, U Cf < 1500 mrem /yr (2.18)

K Seasonal Adjustment Factor K is a seasonal adjustment factor to account for nongrazing. For purposes of demonstrating '

technical compliance for the inhalation pathway, K = 0 throughout the year.

Cf Milk concentration (pCi/ liter)

The' concentration of radionuclide i in milk.

(2.19)

_ Cf=F g Cf Wg exp (-Ag t) M Cf Feed Concentration (pCi/kg)

The concentration of radionuclide i in feed.

Cf=dixr -

1 - exp (- AEi t,)-

A (2.20)

Yv Ei (Note that this assumes feed to be 100%

pasture grass.)

dg Deposition Rate (pCi/m x hr) 6 +

dg = 3600 x 10 x Og , (D/Q) , + Qgy (D/Q)y Q ig (D/OI g (2.21) f.

e 2.1-18

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REVISION 5 FEBRUARY 1983 <

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Cf Concentration in the Discharge Tank (L.i/ml)

The concentration of radionuclide i in the (radwaste discharge or other similar) tank. ,

I F Flow Rate, Radwasts Discherge (ft /sec)

The flow rate of radwaste from the discharge tank to the initial dilution stream.

d Flow Rate, Initial Dilution Stream F (ft /sec)

The flow rate of the in'itial dilution stream which carries the radionuclides to the un-restricted area boundary (e .g . , the blow-down from cooling tower or lake or the circulating cooling water flow).

MPC 1 Maximum Permissible Concentration (VCi/ mil E . The maximum permissible concentration of nuclide i l (or unknown nuclide) in water in the unrestricted area (see Table 7.1-10; or 10 CFR 20, Appendix B, Table II, Column 2 including Note 3.c).

2.2.3 10 CFR 20 Maximum Permissible Concentrations at the Nearest Surface Water Supply The quantity of radionuclides, excluding tritium and dissolved or. entrained noble gases, in outdoor tanks without overflow l pipes connected to other storage tanks shall be limited to l

! ensure that-in the case of an' overflow, the. annual average con-centrati~on of radioactivity in the potable water of the nearest surface water supply is less.than the 10 CFR 20, Appendix B, Table II, Column'2 limits.

! The annual average concentration of each radionuclide in the potable D

e water of the nearest surface water supply is calculated as follows:

'# F C" = Ct t M[ exp (- Agxto) x j(gp (2.29) l YO Y l

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_ . _ . __,_ _ _ _ 2.2 _4,_. _ _ _ . . _ _ . _ _ _ . _ _ - _ . _ _ _ . _ . _ . _ . . . _ . _ _ _

REVISION 5 FEBRUARY 1983 4.0 AQUATIC TRANSPORT AND DOSE MODELS TABLE OF CONTENTS PAGE

- 4.0 AQUATIC TRANSPORT AND OOSE MODELS 4.1-1 4.1 AQUATIC TRANSPORT . 4.1-1 4.1.1 River Model 4.1-1 4.1.2 . Lake Michigan Model 4.1-1 4.1.3 Symbols Used in Section 4.1 4.1-3 l 4.2 AOUATIC DOSE MODEL 4.2-1 4.2.1 Symbols Used in Section 4.2 4.2-2 4.3 AQUATIC TRANSPORT DURING TANK OVERFLCW CONDITIONS 4.3-1 4.3.1 River Model 4.3-1 4.3.2 Lake Michigan Model 4.3-1 4.3.3 Symbols Used in Section 4.3 4.3-2 y..

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REVISION 5 FEBRUARY 1983 4.0 AQUATIC TRANSPORT AND DOSE MODELS 4.1 AOUATIC TR.5N_ SPORT Dose.via the aquatic pathway is discussed in Section 2.2.

Two dilution factors are considered; F, the flow of the

, receiving body of water; and 1/M, an additional dilution factor.

4.1.1 River Mode 2.

For purposes of calculating the drinking water dose from liquid effluents discharged into a river, it is assumed that total mixing of the discharge in the river flow (F") occurs prior to consumption. No additional dilution is assumed to occur; thus 1/M" equals 1.0. The river flow is taken as the long-term ,

(generally 10 years) average. The nearest potable water intakes

. on the receiving bodies of water are described in a footnote to Table 7.2-1.

For the-fish consumption pathway, a near-field dilution flow F I is used; 1/MI = 1.0.

4.1.2 Lake Michigan Model i

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For purposes of calculating doEt from liquid effluents dis- -

charged to Lake Michigan, it is assumed that the concen'tra-tion of radioactivity is diluted initially in the condenser cooling water of flow (FC ) and then by an additional factor 1/M" of.60 prio'irto consumption as potable water. The dilution factor of 60 is the product of the initial entrainment dilution (factor of 10); the plume dilution.(factor of 3 over approximately 9-1 mile); and the current direction frequency (annual average i .

factor of 2).

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4.1-1

REVISION 5 FEBRUARY 1983 L

For the fish ingestion pathway only, it is assumed the radio-activity is diluted fully in a hypothetical river of flow F;I 1/MI = 1.0. To determine F f , it was assumed that the near ,

shore lake current (which can vary in width from 2 to 10 miles) l constitutes a " river" 5 miles wide, 50 feet deep (the avsrage lake depth from shore to. 5 miles near Zion), and flows at the offshore, measured average speed of 0.2 mile per hour. This j results in FI = 4.0 x 105 ft 3 /sec.

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REVISION 5 FEBRUARY 1983

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4.1.3 Symbols Used In Section 4.1 SYMBOL NAME W I T, F Flow df the Receiving Body of Water 1/M Additional Dilution Factor F" Average Flow Rate (ft /sec)

(Drinking Water Pathway) 1/M" Additional Dilution Factor (Drinking Water Pathway)

I F Near-Field Flow Rate (ft /sec)

(Fish Ingestion Pathway) v f

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Additional Dilution Factor (Fish Ingestion Pathway)

F c Average Flow of the (gal / min)

Condenser Cooling Water During the Period of Discharge 4.1-3

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(,; LA SALLE REVISION 10 SEPTEMBER 1983 i . The mid- and high-range detection systems consists of solid-state UdTe (C1) detectors, shielded sample chambers, and pre-ampliers.- Signals fpom the three d.etection systems are processed by a microprocessor which also controls the system pumps and

~

monite s process stream and sample flowrates. The individual detection system outputs and other system parameters are displayed on a digital readout and control module. A three-pen recorder is utilized to record the individual detection system results in pCi/cm 3. The detection system whose output is indicative of the existing release activity is converted by the micropro-cessor to pCi/sec utilizing the existing process stream flowrate and recorded on a single-pen recorder. This .pci/see value

. . . is also compared to an operator-entered alarm point.

The' recorders and digital readout and control module are located

. '# in the main control room. The sample conditioning skid, detec-tion skid, and microprocessor are located in the auxiliary building on the 796 ft 6 in. elevation. Power is suppl!.ed

to this monitor from Division 1 power.

l

- Detector efficiencies are initially determined by calibration with Xe-133 gas. Once operational, efficiency factors will be based on monitor response and isotopic analysis data.

The alarm setpoint for this monitor will be selected to ensure that the combined release rate of the station vent stack and l SGrs' stack does not exceed the most conservative release limit l

l . determined from Equations 8.1 and 8.3 by setting the alarm t-b point at or.below one-half the release limit. .

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D' 8.1.4 Standby Gas Treatment Stack Monitor I

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Release:of radioactivity from the standby gas treatment system (SGrs) stack is monitored by one of three SGrS monitoring systems.

8.1-4 . . _ _ , _

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LA SALLE REVISION 10

( SEPTEMBER 1983 I

Two of the systems consist of a beta sensitive scintillation detector for particulate; a beta sensitive scintillation detector for low-range noble gas; a beta sensitive scintillation detector for high-range noble gap; and a gamma. sensitive scintillation detcetor for iodine. Provisions are made for system inlet and outlet grab. samples.

The monitoring system uses a~ microprocessor to analyze the data from the beta and gamma scintillation detectors. This micro-processor performs-background subtraction and compares the radi-ation values against operator entered alarm-limits. A four-pen strip chart recorder records the monitoring system output.

Alarms are located in che main control room.

Power is supplied to this monitor subsystem from Division ~2 power. The equipment for each monitoring channel is skid modnted

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.and located on the 786 ft 6 in. elevation in the auxiliary building.

-The third ScrS monitor (OPLD2J ) utilizes an isokinetic probe to sample the effluent stream prior to discharge into the atmosphere.

The offline. monitor consists of three detection systems. Gas flow through the system 1; prcvided by vacuum pumps; one for the low-range detection system and one for l

I the mi6- and high-range detection systems. A sample conditioning skid,. upstream of the detection system, filters particulate and iodine and provides for collection of particulate and iodine grab samples.

The low-rango detection system consists of a beta scintillation

detector, a shielded sampling chamber, and a preamplifier.

The mid- and high-range detection systems consist of solid-4

-state CdTe (Cl) detectors, shielded sample chambers, and f prearplifiers. Signals from the three detection systems

.are processed by a microprocessor which also controls the system pumps'and monitors process stream and sample flowrates.

_ . . _ _ _ _ _ _ _ _ . _ . . _ _ _ . _- - - - -8.1-5..____.___.-__----------

LA SALLE REVISION 10

\s , SEPTEMBER 1983 The individual detection system outputs and other system para-meters are displayed on a digital readout and control module.

A three-pen recorder is utilized to record the individual detec-tion system results id'uci/cm3 . The' detection system whose output is indicative of'the existing release activity is con-verted by the microprocessor to pCi/see utilizing the existing .

process. stream flowrate and recorded on a single-pen recorder.

This yci/sec value is also compared to an operator-entered alarm point.

The recorders and digital readout and control module are located in the main control room. The sample conditioning skid, detec-

ticm skid, and microprocessor are located in the auxiliary bailding on the 796 ft 6 in. elevation. Power is supplied to this monit'or from Division 2 power.

-Detector efficiencies are initially determined by calibration with Xa-133 gas. Once operational, efficiency factors will be based on monitor response and isotopic analysis data.

The alarm setpoint for this monitor will be selected to ensure that the combined release rate of the station vent stack and SGTS stack does not exceed the most conservative release limit t

determined from Equations 8.1 and 8.3 by setting the alarm point at or below one-half the release limit.

l 8.1.5 SJAE Off-Gas Monitors The : steam jet air ejector (SJAE) monitor subsystem continually measures and records the ga.nma radiation in the off-gas as it 9

e-is drawn from the main condenser by the steam jet air ejectors before it passes through the holdup line and carbon beds enroute

(_ to the station vent etack.

8.1-6

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... CN a Commonwrith Edison J 1.aSalle County Nuclear Station

_._ s RuralRoute 41, Box 220 v

J Marseilles, Illinois 61341

/ Telechone 815/357-6761 J

February 15, 1984 Di rector, Of fice of Management- information and Program Control United States Nuclear Regulatory Commission i Washington, D.C. 20555 ATTN: Document Control Desk Gentlemen:

Enclosed for your information is the monthly performance report covering LaSalle County Nuclear Power Station, Unit One, for the period covering January 1 through January 31, 1984.

Very truly yours, j

J. Diederich

' Station Superintendent LaSalle County Station GJD/DLL/bej Enclosure

-xc: J. G. Keppler NRC, Region lli NRC Resident inspector LaSalle

. Gary Wright 111. Dept. of Nuclear Safety

.D. P. Galle Ceco D. L. Farrar ' CECO INPO Records Center Ron A. Johnson,.PlP Coordinator SNED W.R. Jackson, GE Resident

/ J

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LASALLE NUCLEAR POWER STATION UNIT 1 MONTHLY PERFORMANCE REPORT JANUARY 1984 COMMONWEALTH EDISON COMPANY NRC DOCKET NO. 050-373 LICENSE NO. NPP-11 I

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TABLE OF CONTENTS I. INTRODUCTION II.

SUMMARY

OF OPERATING EXPERIENCE FOR UNIT ONE III. PLANT OR PROCEDURE CHANCES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A. Amendments to Facility Licence or Technical Specifications B. Facility or Procedure Changes Requiring NRC Approval C. Tests and Experiments Requiring NRC Approval D. Corrective Maintenance of Safety Related Equipment IF. LICENSEE EVENT REf0RTS 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 Steam Relief Valve Operations

B. ECCS System Outages C. Off-Site Dose Calculation Manual Changes D. Major Changes to Radioactive Waste Treatment System a

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I. INTF0 DUCTION The LaSalle Nuclear Power Station Unit One is a Boiling Water Reactor with a designed electrical output of 1078 MWe net, located in Marseilles.

l Illinois. The Station is owned by Commonwealth Edison Company. The Architect /Ragineer was Sargent & Lundy, and the primary construction contractor was Commonwealth Edison Compsny.

.The condenser cooling method is-a closed cycle cooling pond. The plant is subjset to License Number NPF-11, issued on' April' 17, 1982. The date of initial 1

criticality was June 21, 1982. The unit commenced commercial generation of power on January 1, 1984.

This repc-t was compiled by Diane L. Lin, telephone

' number (815)357-6761, extension 499.

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' JANUARY 1 'The unit: was s;iutdown due to's ventilat. ion problem in'thi[dr'ywell and thi eds1*acement of cables

.wh'ich were'damagid by excessive heat its the drywell.

~

. .s JANUARY 1-4 Thereactorwentcrit$ cal.at1736hourson January'1. At 1040. hours'~on January 2 the main' generator was.syr.chronize'd to:the b id andflocJed. At.

1630 hours0.0189 days 0.453 hours 0.0027 weeks 6.20215e-4 months 'on' January 7 reactor power'was 22%. On January 3 at 0431~h'ours the main turbine was manually tripped. 'At c537 hours on January 3 the reactor was manually scrammed for a sequence change in order to perform scram time testing. The reactor was critical.

for 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months and 1 minute.

[ANUARY 5-6 The reactor went critical at 0500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months on

[ January 5. At 0510 hours0.0059 days 0.142 hours 8.43254e-4 weeks 1.94055e-4 months on Janut.ry 6 the main

. generator was synchronized to the grid and loaded. At

.1525 hours0.0177 days 0.424 hours 0.00252 weeks 5.802625e-4 months on-January 6 reactor powcr was 25%. .On January 6 at 2210 hours0.0256 days 0.614 hours 0.00365 weeks 8.40905e-4 months the reactor scrasuned due to low reactor water level resulting from the 'A' TDEFP trip on high seal injection temperature. The reactor was celtical for al: hours and 10 minutes.

JANUARY 7-13~ The reactor went critical at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months on January 7. -At 2138 hours0.0247 days 0.594 hours 0.00354 weeks 8.13509e-4 months on January 7 the main generator was synchronized to the grid and loaded. At 2200 hours0.0255 days 0.611 hours 0.00364 weeks 8.371e-4 months on January 7 reactor power was 25% At 1500 hours0.0174 days 0.417 hours 0.00248 weeks 5.7075e-4 months on January 8 reactor power was 64%. At 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months on January 12 reactor power was 84%. On January 13 at 2115 hours0.0245 days 0.588 hours 0.0035 weeks 8.047575e-4 months the reactor scrammed on main turbine trip due to a generator protective. trip caused by the s.aln

-generator pot fuse device opening up. The reactor was critical for 149 hours0.00172 days 0.0414 hours 2.463624e-4 weeks 5.66945e-5 months and 15 minutes.

l JANUARY 14-19 The reactor went critical at 1310 hours0.0152 days 0.364 hours 0.00217 weeks 4.98455e-4 months on January 14. At 1950 hours0.0226 days 0.542 hours 0.00322 weeks 7.41975e-4 months on January 14 the main generator was synchronized to the grid and' loaded. At[

2300 hours0.0266 days 0.639 hours 0.0038 weeks 8.7515e-4 months on January 14 reactor, power was 23%. At 0720-hours on January 15 reactor power was 43%. .At 1500 .

hours.on January 15 reactor power was'67%.1 On January

. 16 at 1525 hours0.0177 days 0.424 hours 0.00252 weeks 5.802625e-4 months the resctor scrammed due.to the "C" condenser boot seal whict. failed and iub' sequent loss of condenser vacuum. The reactor was critical for 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months '

and 15 minutes.-

p JANUARY 20-31 The~ reactor.went critical'at 2040 hours0.0236 days 0.567 hours 0.00337 weeks 7.7622e-4 months on January 20. .At 1823 hours0.0211 days 0.506 hours 0.00301 weeks 6.936515e-4 months on_ January 21 the main '

generator was synchronized to the r, rid and loaded. . At!

1530 hours0.0177 days 0.425 hours 0.00253 weeks 5.82165e-4 months on January 22 reactor power was 68%. At 3500

^

hours on January 25 reactor power was,86%. The reactor was critical for'267 hours0.00309 days 0.0742 hours 4.414683e-4 weeks 1.015935e-4 months and 20 minutes. .

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.III. PLANT OR PROCEDURE CHANCES, TESTS, EXPERIMENTS AND SAFETY RELATED MAINTENANCE.

A. . Amendments to Facility License or Technical Specifications.

Amendment 15- This amendment revised the control rod drive coupling specification, 3.1.3.6, include Action Statement "c" "The provisions of specificiation 3.0.4 are not applicable". The same change was added to the control rod operability specification, 3.1.3.1, in Action Statement b.3.

B. Facility or Procedure Changes Requiring NRC Approval.

There were no facility or procedure changes requiring NRC

, approval'during the reporting period.

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 Effects on Safe Operation, and Corrective Action .

Document 0043r

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.LTP-300-7) ,

!ATTACIGIENT A- Revision 3 ~ '

March 1, 1C83.

CORRECTIVE MAINTENANCE OF ,

'5 ,-

SAFETY RELATED EQUIPMENT l 1

-WORK. REQUEST' LER COMPONENT- CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION L20584 83-143/ RCIC. Inject- High air Temperature Physical damage to the Replaced-cables 01T-0 tion Testable cables Check Valves

, L29888 RHR "B" LPCI High drywell temperature Possible damage to limit Replaced cable and limit Testable switches and cables switch Check and By-pass Valves 4

L29891 -

RCIC Test- High drywell temperature Possible damage to limit Repulled cable able check switch and cable Bypass Valve 4

L31885 -

Aux. Electric Bad Pase This fuse powers dampers Replaced fuse l Equip. Room OVE07YA,OVE08YA,OVE09YA j HVAC Dampers i

L31125 -

RHR Valve . Limit out-of-adjustment Valve drives too far into Reset valve motor limits lE12-F003B seat a

i DOCUMENT 0044r

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L g LTP-300-7 ATTACHMENT.A Cont'd Revicion 3 '

MacIh 1, 1983

. CORRECTIVE. MAINTENANCE OF 5

, SAFETY litELATED EQUIPMENT s

I WORK REQUEST LER COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION

~~

'L31232 Reactor Press- Switch out-of-adjustment Switch tripped at 942 psig Recalibrated switch ure Switch .Rx pressure instead of 1043 for Low Vac- psig uum Bypass Logic Div.A-2 L31252 -

HPCS Dischar- Transmitter out-of-ad- Gauge is pegged high Recalibrated transmitter se Pres Gauge justment L31288 -

VE Return Limit switch out-of-ad- Damper shows dual indication Reset limit switch Suction Damper justment when full closed L31316 -

Fuel Oil Tran- Bad 0-rings Valve does not open Replaced 0-r'ngs .

sfer Pump Dis-.

charge Soleno--

id Valve DOCUMENT 0044r J

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LTP-300-7 ATTACHMENT A Revicion 3 .

Marth 1.'1983 CORRECTIVE MAINTENANCE OF 5 -.

SAFETY RELATED EQUIPMENT 1 I I I I WORK FEQUEST LER COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION-ON SAFE OPERATION-I I ~l i I L31352 Inboard MSIV Incorrectly wired- Switch indicates closed Rewired limit switch per Limit Switch when valve is open drawing L31358 "B" VE Mixing Linkage arm disconnected Wont' operate properly Reconnected linkage arm Damper from damper L31437 SBGT Radiat- Defective display unit Cannot read display unit Replaced display unit lation Monit readings or s

L31666 "B" Rod Block Relay loose in its Channel does not bypass Reseated K9 relay Monitor Chan socket automatically when an edge nel rod is selected L31906 Containment Bad chart drive motor Recorder does not drive Replaced chart drive Temperature chart in slow speed motor Suppression Pool Water Temp. Record-er L32007 Squib Valve Meter lamp burned out No light on apron Replaced lamp and holder Continuity assembly Alarm DOCUMENT 0044r

LTP-300-7:

ATTACHMENT A Cont'd. Revicion 3. -

. March l 'i, 1983; CORRECTIVE MAINTENANCE OF 5 ',M SAFETY RELATED EQUIPMENT WORK REQUEST LER COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE OPERATION l

L32196' - APRM Bad DC amplifier card. APRM output' remains down- Replaced DC rday11fier Channel "A" scale with Rood input card

, L32199 --

"B" ADS "S" Bad lamp driver card Low pressure alarm is up on Replaced lamp driver Accumulator ESF alarm panel and won't card clear L32030 --

"B" LPCI low' Switch out-of-tolerance Alarm comes up for no ap- Calib.' pressure switch Header Press- parent reason ure Switch 4

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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, January 1 through January 31, 1984. 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 83-153/01T-0 12/27/83 "A" DFP 83-154/03L-0 12/14/83 Turbine Stop Valve RPS Limit Switch Setpoint exceeded L.C.O.

. 83-155/03L-0 12/29/83 Minimum Flow Valve Thermal overload Bypass Logic 83-156/03L-0 12/29/83 RPS Limit Switch Exceeds L.C.O.

84-001-00 1/23/84 Radwaste Discharge Process Rad Monitor i

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V .~ DATA TABULATIONS The following data tabulations are presented in thia report:

.A. Operating Data Report B. Average Daily Unit Power Leywl C. Unit Shutdowns and Power Reductions b

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ML20080U033

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SUMMARY

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