ML20215C708

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Monthly Operating Repts for Nov 1986
ML20215C708
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 11/30/1986
From: Diederich G, Peters J
COMMONWEALTH EDISON CO.
To:
NRC OFFICE OF ADMINISTRATION (ADM), NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
NUDOCS 8612150344
Download: ML20215C708 (45)
v • d • e


Text

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LASALLE NUCLEAR POWER STATION i UNIT 1 l

' MONTHLY PERFORMANCE REPORT l

NOVEMBER 1986 l l COMMONWEALTH EDISON COMPANY l

l NRC DOCKET NO. 050-373 LICENSE NO. NPF-ll 8612150344BhIh73 ADOCK O PDR PDR R

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TABLE OF CONTENTS I. INTRODUCTION II. MONTHLY REPORT FOR UNIT ONE A. Stannary of Operating Experience B. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE

1. Amendments to Facility License or Technical Specifications
2. Facility or Procedure Changes Requiring NRC Approval
3. Tests and Experiments Requiring NRC Approval
4. Corrective Maintenance of Safety Related Equipment
5. Completed Safety Related Modifications.

C. LICENSEE EVENT REPORTS D. DATA TABULATIONS

1. Operating Data Report
2. Average Daily Unit Power Level
3. Unit Shutdowns and Power Reductions E. UNIQUE REPORTING REQUIREMENTS
1. Main Steam Relief Valve Operations
2. ECCS System Outages
3. Off-Site Dose Calculation Manual Changes
4. Major Changes to Radioactive Waste Treatment System
5. Indications of Failed Fuel Elements i

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o I. INTRODUCTION 4

The LaSalle County Nuclear Power Station is a two-unit facility owned by comanonwealth Edison Company and located near Marseilles, Illinois. Each unit is a Boiling Water Reactor with a designed net electrical output of 1078 Megawatts. Waste heat is rejected to a man-made cooling pond using the Illinois River for make-up and blowdown. The architect-engineer was Sargent and Lundy and the primary construction contractor was Commonwealth Edison Company.

Unit one was issued operating license number NPF-ll on April 17, 1982. Initial criticality was achieved on June 21, 1982 and commercial power operation was commenced on January 1, 1984.

This report was compiled by James P, Peters, telephone number (815)357-6761 extension 325.

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D II. MONTHLY REPORT FOR UNIT ONE 4

A. SUfetARY OF CPERATING EXPERIENCE FOR UNIT ONE November 1-31 November 1, 0000 Hours. The Unit entered November with the Reactor critical and on-line at 84.2% Power (980 MWe).

November 1, 1230 Hours. Began Increasing Power 10 MWe/hr. to 92.4%

(1075 MWe).

November 2, 0250 Hours. Began Decreasing Power 50 MWe to 87.7% (1020 MWe) for Control Rod Manipulations.

November 4, 1980 Hours. Began Decreasing Power 25 MWe to 85.9% (1000 MWe) due to Feedwater Heater 14B Trouble.

November 4, 1842 Hours. Began Increasing Power to 89% (1040 MWe).

November 4, 1930 Hours. Reactor Power Decreased to 86.8% (1010 MWe) due to Feedwater Hester 14B Trouble.

November 4, 1931 Hours. Began Increasing Power 10 MWe/hr. to 88.6%

(1035 MWe).

November 5, 0905 Hours. Began Decreasing Power to 86.8% (1010 MWe).

November 5, 1020 Hours. Began Increasing Power 10 MWe/hr to 91.5%

(1065 MWe).

November 5, 1755 Hours. Reactor Power Decreased to 85.9% (1000 MWe) due to Feedwater Heater 14B Trouble.

November 5, 1826 Hours. Began Increasing Power 10 MWe/hr to 92.8%

(1080 MWe).

November 8, 0400 Hours. Began Decreasing Power to 86.8% (1010 MWe) due to Control Rod Manipulations.

November 8, 0430 Hours. Began Increasing Power 5.8 MWe/hr to 91.5%

(1065 MWe).

November 15, 0255 Hours. Began Decreasing Power to 82.5% (960 MWe) for Monthly TCV and MSIV Surveillance and Control Rod Manipulations.

November 15, 0530 Hours. Began Increasing Power 7 MWe/hr to 92.8%

(1080 MWe).

November 20, 1310 Hours. Reactor Power Decreased to 45.9% (535 MWe) due to Feedwater Heater 14B Trouble.

November 20, 1415 Hours. Began Increasing Power (Fast Ramp) to 85.1%

(990 MWe) then at 10 MWe/hr to 92.8% (1080 MWe).

November 21, 0405 Hours. Began Decreasing Power to 62.7% (730 MWe) to Test Control Rod Drive 10-47.

November 22, 1600 Hours. Began Increasing Power 7 MWe/hr to 75.6%

(880 MWe).

November 23, 0550 Hours. Reactor Power Decreased to 58.4% (680 MWe) due to shutdown of "D" CD/CB Pump.

November 24, 2035 Hours. Began Decreasing Power to 49.9% (580 MWe) for Friction Testing of Controil Rod Drive 10-47.

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II. MONTHLY REPORT FOR UNIT ONE A: SUlttARY OF OPERATING EXPERIENCE FOR UNIT ONE November 1-31 (Continued)

November 24, 2205 Hours. Began Increasing Power 15 MWe/hr to 90.2%

(1050 MWe).

November 29, 2240 Hours. Began Decreasing Power 50 MWe/hr to 85.9%

(1000 MWe) due to control Rod Manipulations.

November 29, 0045 Hours. Began Increasing Power 20 MWe/hr to 89.4%

(1040 MWe).

November 30, 2400 Hours. Reactor and Generator on line and holding at 88.9% Power (1033 MWe).

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B. PLANT OR PROCEDURE CHANGES, TESTS,. EXPERIMIDITS AND SAFETY RELATED MAINTENANCE.

1. Amendments to facility license or Technical Specification.

4 Amendment 47 - This revises the LaSalle County Station Unit #1 Technical Specifications to incorporate certain

. changes made to the Administrative Controls Section of the Technical Specifications.

2. Facility or procedure changes requiring NRC approval.

There were no facility or procedural changes requiring NRC approval during this reporting period.

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3. Tests and Experiments requiring NRC approval.

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

4. Corrective maintenance of safety related equipment.

The following table (Table 1) presents a summary of safety-related maintenance completed on Unit One during the reporting period. The headings indicated in this summary include: Work Request number, Component Name, Cause of Malfunction, Results and Effects on Safe Operation, and Corrective Action.

5. Completed Safety Related Modifications.

The following Table (Table 2) presents a list of completed modifications during this reporting period. Each entry will have a short synopsis explaining details involved with each modification.

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TABLE 1 CORRECTIVE MAINTENANCE OF -

SAFETY RELATED EQUIPMENT UNIT #1 WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE PLANT OPERATION L4~1282 "A" VC Flow Switch Failed Relay Relay at Panel was cycling Replaced Relay.

Relay, OPSY-VC003X excessively, blackening glass.

L59413 "A" VC Flow Switch Failed Relay Relay at Panel was cycling Replaced Relay.

Relay, OPSY-VC006X excessively, blackening glass.

L62921 "B" MSL Radiation Bad indicating device, Continuous alarms in the Replaced Indicating Monitor ID18-K610B ID18-K610B Control Room. Device, ID18-K610B.

L62930 "B" CRD Charging Excessive Static Build-up Monitor Indicates 1020 PSIG Cleaned and sprayed anti-water scram monitor when other 3 indicate static solu '.9 to free ICll-N616D 1200 PSIG. up motor.

L62945 HCU-30-ll Excessive Leakage Must charge HCU 2-3 times Replaced valve.

Instrument Stop Valve per shift.

ICll-D001-lli L63116 "A" MSL Radiation Setpoints determined Occasional upscale alarms. Performed setpoint Monitor 1D18-K610A using too low a background change per LRP-1820-15.

setpoint.

L63124 Div-II S/P and D/W Bad Pinion Gear and Paper does not advance. Replaced Pinion Gear Temperature Recorder clutch assembly, and clutch assembl*/

ITR-CM038.

L63276 Div-II S/P and D/W Loose Connections on Temperatures range 6*-10*F. Tightened loose con-Temperature Recorders Range Card below ITR-CM037. nection and replaced ITR-CM038 Simplifier Assembly.

L63283 "B" VC Train Ammonia Zero drifted low causing Detector could not be Readjusted zero drift.

Detector OXY-VCl25B master fault. calibrated.

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TABLE 1 CORRECTIVE MAINTENANCE O ' ,

SAFETY RELATED EQUIPMENT UNIT #1 WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE PLANT OPERATION L63307 Div-I D/W and S/P Bad Cervo Motor. Temperatures are consistently Replaced Servo-motor.

Temperature Recorders higher. and recalibrated.

ITR--CM037.

L63369 "A" VC Train Ammonia Fallea WISA Pump. Ammonia Detector spiking is Replaced Wisa Pump and Detector OXY-VC125A tripping logic, readjusted.

L63518 Rod 10-4~1 Insert Vent Bad Valve Internals Valve is stuck 1/2 - 1 turn Replaced valve internals.

Valve ICll-F101. open.

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TABLE 2

. COMPLETED SAFETY RELATED MODIFICATIONS i

MODIFICATION NUMBER: A brief synopsis of incorporated modification objectives with final design resolution. Also, state reviewed or enreviewed safety questions.

The following list below has all the Safety Related Modifications completed from January 1, 1986 through November 30, 1986.

M-1-0-82-074 - This Modification installs check valves in Service Air Piping to prevent contamination of SA System due to discharge pressure in the CP and FC Systems. This is per NRC IE Information Notice #79-08. There was no unreviewed safety question.

M-1-0-82-132 - The purpose of this modification is to meet License condition 2.C.(21)(c) that requires all control and monitoring instruments for the "0" Diesel Generator be removed from the engine skid, except for the instruments that are qualified for this location. There was no unreviewed safety question.

M-1-0-83-020 - Replace the existing Control Room 6 ft. 5 in. x 10 ft. 2-1/2 in. bullet resistant door with a 6 ft. 5 in. x 8 ft. O in.

U.L. listed combination high power rifle, bullet resistant and 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rated door and frame. The height of the existing door frame will be reduced by adding a block wall above door frame. There was no unreviewed safety question.

M-1-0-83-026 - This Modification involves modifying the lube oil system on the "0" Diesel Generator in order to provide a consistent supply of oil to the turbo-charger and crankshaft in anticipation of an emergency start. There was no unreviewed safety question.

M-1-0-84-056 - This Modification will install noise suppressors to reduce electronic noise emmission that occurs during a source check or repositioning of flow control valve on WGRM units. There was no unreviewed safety questions.

M-1-1-80-001 - This Modification consists of reducing the water rod lower end plug shank length and replacing the upper expansion spring with a modified design spring. There was no unreviewed safety questivn.

M-1-1-82-080 - This Modification calls for installation of containment flood level instrumentation for a parameter already analyzed.

There was no unreviewed safety question.

M-1-1-82-Il9 - This modification will replace the existing drywell chiller drain flow instrumentation. The drain line from the primary containment chillers will be crosstied with a common flow loop monitoring this condensation for compliance with the 5 gpm unidentified leakage. There was no unreviewed safety question.

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TABLE 2 COMPLETED SAFETY RELATED MODIFICATIONS

- M-1-1-82-242 - The Modification calls for the relocation of the inboard and outboard MSIV - Leakage Control System dilution air flow transmitters. These transmitters will be locally relocated above their instrument line process taps to insure adequate line drainage. There was no unreviewed safety question.

j M-1-1-82-263 - This Modification installs on the scram discharge volume, diverse and redundant level instrumentation and redundant vent and drain valves. This will satisfy NURBG-0803 and Licensing condition #14. There was no unreviewed safety question.

i M-1-1-82-284 - This Modification provides for the addition of the degraded voltage protection scheme to the ESP Buses 141Y, 142Y and i 143. This provides additional protection for safety related i equipment by preventing a failure caused by sustained.

operation at low grid voltage. There was no unreviewed Safety Question.

M-1-1-82-290 - This modification installs redundant fault protection on low voltage electrical penetrations passing through primary containment. In the event of a single breaker failure the mechanical integrity of the penetration will then be maintained. There was no unreviewed safety question.

I M-1-1-82-292 - This Modification involves the replacement of the 6 in. x 4 in, reducer and weldolet on the HPCS minimum flow line with a 6 in. sweepolet to strengthen the piping where flow induced vibration causes high stress. There was no unreviewed safety question.

M-1-1-82-305 - This Modification installs 4 pressure transmitters which operate independently, on the CRD charging water header to the HCU's. This trip will result in a reactor scram when a l low charging water heuder pressure exists. There was no unreviewed safety question.

M-1-1-82-310 - The purpose of this modification is to meet License conditions 2.C.(21)(c) that requires all control and monitoring instruments for the "lA" Diesel Generator be removed from the engine skid, except for the instruments that l are qualified for this location. There was no unreviewed safety question.

M-1-1-82-319 - The purpose of this modification is to meet license condition i 2.C.(21)(c) that requires all control and monitoring instruments for the "lB" Diesel Generator be removed from the engine skid, except for the instruments that are qualified for this location. There was no unreviewed safety question.

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TABLE 2 COMPLETED SAFETY RELATED MODIFICATIONS 1

4 M-1-1-83-018 - This Modification adds display instrumentation and alarms to Unit 1 Control Room for each class 1E 125 VDC and 250 VDC 2 Subsystem, per SER 8.3.1.2 and License Condition 2.C.(22).

There was no unreviewed safety question.

M-1-1-83-024 - This Modification installs additional shims in pipe whip restraints. The additional shims are necessary to reduce the

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gap between the piping and whip restraint ring while the piping is in the hot position. There was no unreviewed j safety question.

i M-1-1-83-081 -- Modify shielded airlock entries G-9-A, V-10 and J-13 to comply with required U.L. Fire Rating per NRC Commitment.

There was no unreviewed safety question.

M-1-1-83-089 - This Modification calls for the addition of new pressure switches and relays for Low RHR Pump Discharge Pressure Alarms and ESF Status. There is now a separate annunciator for high and low annunciators. There was no unreviewed safety question.

M-1-1-83-119 - This Modification will remove the existing Bailey 553 transmitter and replaces it with an improved Rosemount 1153GB Transmitter. This will eliminate the Zero-Shift probless and provides a correct pressure indication. There was no unreviewed safety question.

i M-1-1-83-129 - This Modification involves modifying the lube oil system on the "lA" Diesel Generator in order to provide a consistent supply of oil to the turbo-charger and crankshaft in anticipation of an emergency start. There was no unreviewed safety question.

M-1-1-83-135 - This Modification shortens the "lA" RHR Pump Shaft and

Discharge Column due to vibration problems experienced caused

} by the extended length of the pump shaft and discharge i column. There was no unreviewed safety question.

i M-1-1-84-003 - This Modification removes the following testable check valves; lE21-F327A-C, lE51-F354, IE51-F355, IE22-F354 and lE21-P333. These bypass valves are used solely for testing and do not serve a safety function. There was no unreviewed safety question.

4 M-1-1-84-012 - This Modification replaces the MSIV-Leakage control Blowers j Bakelite fan blades in the siemens blower motor with an aluminium substitute. The new fan blade is environmentally qualified. There was no unreviewed safety question.

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TATLE 2 COMPLETED SAFETY RELATED MODIFICATIONS M-1-1-84-015 - Install Control Room Alarms providing indication of any of the 8 MSIV's or 4 TSV's are "NOT FULL OPEN". This will provide trip position of relays during MSIV and TSV surveillances. There was no unreviewed safety question.

M-1-1-84-026 - This Modification replaces four 26 in. limitorque valves on the purge lines; four 26 in, limitorque valves on the vent i lines and two 8 in. limitorque valves on the tie lines with appropriate wafer stop valves from Clow corporation. These new valves are operated by high torque Bettis actuators.

There was no unreviewed safety question.

M-1-1-84-036 - This Modification eliminates the need for manual initiation of ADS to assure adequate core cooling for those transients not associated with high drywell pressure. It will provide ADS Manual inhit-it switch and high drywell pressure bypass timer. There was no unreveiwed safety question.

M-1-1-84-047 - This Modification routes a new cable from thermal element in the Suppression Pool to a temperature indicator at the remote shutdown panel. This cableing will separate indication feeds to two indicators. There was no unreviewed safety question.

M-1-1-84-052 - This Modification removes all woodruff keys that engages the worm shaft to the worm gear. This will prevent manual positioning of the air operated control valves or forced open against the spring pressure, defeating an isolation signal.

There was no unreviewed safety question.

M1-1-84-053 - This Modification provided the replacement of interlocks in

electrical contractors for the 12 VDC powered valves at MCC l 121Y. This replacement is required in order to comply with Environmental Qualification Constraints. There was no f unreviewed safety question.

l M-1-1-84-059 - This Modification installed 16 branch connection fittings on l the control Rod Drive Scram Discharge Volume Headers to allow l

the system to be hydrolyzed occasionally to prevent contamination buildup. There was no unreviewed safety j question.

M-1-1-84-066 - This Modification provided the installation of a thermowell in the RHR Shutdown Cooling Common Suction Header. The Thermocouple wil be connected to a recorder with a high l temperature alarm. There was no unreviewed safety question.

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M-1-1-84-067 - This Modification reused the cable and associated wire l terminations in main control room panel for standby liquid control system to maintain at least 6 inches of separation

! between circuits of Train "A" and "B". There was no j unreviewed safety question.

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TABLE 2 COMPLETED SAFETY RELATED MODIFICATIONS s

M-1-1-84-072 - For pressure switches related to Group I - III, replace the l present pressure switch with a new switch from Static-O-Ring. This is required for compliance with 10CFR50.49, Environmental Qualification. There was no unreviewed safety question.

M-1-1-84-084 - This modification replaces various snubbers with struts due to revised thermal modes of various systems. There was no unreviewed safety question.

M-1-1-84-088 - This Modification split-up high and low pressure alarms on the LPCS Pump Discharge Pressure. The high pressure alarm uses the original switch and the low pressure uses a new switch and new annunciator window. There was no unreviewed safety questions.

M-1-1-84-090 - This Modification implements the replacement of motors in limitorque valve actuators having B insulation with equivalent motors having RH insulation. This replacement is required to meet environmental qualification criteria per NUREG 0588. There was no unreviewed safety question.

M-1-1-84-091 - This Modification implements the replacement of presently installed Barton differential pressure switches with new switches from SOR, Inc. The SOR switches are environmentally qualified for service in pertinent plant areas; their installation is requied for compliance with 10CFR50.49.

There was no unreviewed safety question.

M-1-1-84-092 - The Modification implements the installation or conductor seal assemblies on four limit switches mounted on each of eight MSIVs, as required for compliance with 10CFR50.49.

There was no unreviewed safety question.

M-1-1-84-093 - This Modification implements the replacement of seals and switch mechanisnm in level controllers manufactured by Magnetrol International, as required for compliance with 10CFR50.49. There was no unreviewed safety question.

M-1-1-84-104 - This Modification was implemented to reroute and physically separate the scram contactor wiring for scram contactors K14A thru H to assure that the line and load side wiring do not come in contact with each other. There was no unreviewed safety question.

M-1-1-84-105 - This Modification calls for the addition of a control box to the circuits for solenoids 11N100 and IIN101. The purpose of the control box is to reduce the coil holding voltage so that the heat build-up is also reduced. There was no unreviewed safety question.

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TABLE 2 COMPLETED SAFETY RELATED MODIFICATIONS M-1-1-84-106 - This Modification provides for replacing certain Barton 760 instruments with environmentally qualified units, as required for compliance with 10CFR50.49. There was no unreviewed safety question.

M-1-1-84-123 - This Modification was implemented to remove the energy absorbing material (EAM) from certain pipe whip restraints no longer required for plant safety. The EAM is required for testing. There was no unreviewed safety question.

M-1-1-84-124 - This Modification was implemented to reduce the end and side clearances in five battery racks in order to bring the racks into a seismically qualified configuration. There was no unreviewed safety question.

M-1-1-84-125 - This Modification was implemented in conjunction with and for the same purpose as M-1-1-84-124. There was no unreviewed safety question.

M-1-1-84-126 - This Modification was implemented in conjunction with and for the same purpose as M-1-1-84-124. There was no unreviewed safety question.

M-1-1-85-011 - This Modification was implemented to install new flow elements and transmitters to detect inboard MSIV leakage.

This modification is required for compliance with 10CFR50.49. There was no unresolved safety question.

M-1-1-85-026 - This Modification changed the Unit 1 SBGT electrical heater breaker trip settings. The new heater draws more current (per design) than the one it recently replaced. There was no unreviewed safety question.

M-1-1-85-028 - This Modification was implemented to relocate the process Radiation Monitoring Tap locations on the RHR service water piping in order to allow complete sampling of the RHR WS on Unit 1. Previous design did not allow RHR Pump seal cooling water to be sampled. There was no unreviewed safety question.

M-1-1-85-037 - This Modification was implemented to rework spent fuel pool piping to support the replacement of values on adjacent piping systems. There was no unreviewed safety question.

M-1-1-85-039 - This Modification was implemented to replace the HTX series Asco Solenoid Valves on the MSIV actuators with equivalent NP series solenoid valves. The modification will improve the equipment qualification record associated with the installed hardware and will implement corrective actives recommended by ASCO (IE Information Notice 85-17). There was no unreviewed safety question.

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TABLE 2 COMPLETED SAFETY RELATED MODIFICATIONS o

M-1-1-85-045 - This Modification removed portions of the soft seat from the Inboard Feedwater check valves in order to eliminate seat leakage resulting from degradation of the soft seat material. There was no unreviewed safety question.

M-1-1-86-003 - This Modification cut the packing leak-off line adjacent to valve 1812-F050A and installed a threaded coupling. This allows the valve to be removed for maintenance without 1

continually cutting / rewelding the leak-off line. There was no unreviewed safety question.

M-1-1-86-031 - This Modification was implemented to remove a snubber from HPCS subsystem lHP-02. The snubber is no longer required to support the piping analysis. There was no unreviewed safety question.

M-1-1-86-032 - This Modification removed a snubber from HPCS subsystem 1HP-06. The snubber is not required to support the current piping analysis. There was no unreviewed safety question.

M-1-1-86-033 - This Modification removed a snubber M-1302-22-102S from HPCS subsystem 1HP-A1. The snubber is not requied per the current piping analysis. There was no unreviewed safety question.

M-1-1-86-037 - This Modification was implemented to delete lines 1RT51 A/B/C (valve leak-off lines) due to snubber failures. Leakage is not directed to the drywell equipment drain sump. There was

no unreviewed safety question.

M-1-1-86-040 - This Modification removed a snubber 1RH40-1506S from subsystem 1RH-04 in order to prevent possible unnecessary i stresses on line IRH40BA-12". There was no unreviewed safe ty

question.

M-1-1-86-042 - This Modification removed a snubber 1R44-1539S/1543S/1544S/

1551S/1554S from subsystem 1RH-08. The snubber is not required per the current piping analysis. There was no unreviewed safety question.

! M-1-1-86-046 - This Modification removed three snubbers 1RHB4-1007S/1008S/

1011S from subsystem 1RH-71 in order to prevent possible unnecessary stresses on line IRHB4AB. There was no unreviewed safety question.

M-1-1-86-047 - This Modification was implement to delete valve leak-off lines 1RH61 AA/BA/CA because of snubber failures. Leakage is now accomodated by the drywell equipment drain sump. There was no unreviewed safety question.

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TABLE 2 COMPLETED SAFETY RELATED MODIFICATIONS M-1-1-86-048 - THis Modification removed a snubber M-1302-30-52 from subsystem 1RH-A7. The snubber is not required per the current piping analysis. There was no unreviewed safety question.

M-1-1-86-049 - This Modification removed a snubber M-1302-22-110 from subsystem 1RH-64 to prevent possible unnecessary stresses on line IRH11BA. There was no unreviewed safety question.

M-1-1-86-050 - This Modification removed two Reactor Recirculation System Snubbers 1RR17-1003S/1005S replaced a fourth RR snubber with a strut, and deleted a RB system snubber because of snubber failures. The snubbers affected are not required per the current piping analysis. There was no unreviewed safety question.

M-1-1-86-051 - Snubbers M-1302-24-148/149/152 failed LTS-500-14, snubber Functional Testing. These snubbers are removed due to Design Evaluation demonstrating the specific snubbers do not affect the ability of the system to meet its design service. There was no unreviewed safety analysis.

! M-1-1-86-052 - Snubber RR59-409s failed LTS-500-14, snubber Functional Testing. This snubber is removed due to Design Evaluation demonstrating this snubber does not affect the ability of the

system to meet its design service. There was no unreviewed safety analysis.

i l M-1-1-86-053 - Snubber M-1302-24-103 filed LTS-500-14, Snubber Functional Testing. This snubber is removed due to Design Evaluation

demonstrating this snubber does not affect the ability of the system to meet its design service. There was no unreviewed safety analysis.

M-1-1-86-057 - This Modification changes the support arrangment (lRI-A3),

thus creating adequate thermal flexibility and brings the piping arrangement within code limits. There was no unreviewed safety analysis, i

M-1-1-86-058 - This Modification changed the support arrangement (lRH-E5),

thus creating adequate thermal flexibility and brings the

piping arrangement within code limits. There was no unreviewed safety analysis.

I i M-1-1-86-059 - This Modification installed a double block vent to line IRH03BA-12 to function as a high point vent line. These l vents are to improve pipe venting reducing the potential of water hammer and possible snubber damage during operation.

l There was no unreviewed safety analysis.

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TABLE 2 COMPLETED SAFETY RELATED MODIFICATIONS M-1-1-86-060 - This Modification installed a double block vent to line IRH03BB-12 to function as a high point vent line. These vents are to improve pipe venting reducing the potential of i

water hammer and possible snubber damage during operation.

There was no unreviewed safety analysis.

M-1-1-86-062 - This Modification installed a double block vent to line IRH34A-6 to function as a high point vent line, these vents are to improve pipe venting reducing the potential of water hammer and possible snubber damage during operation. There was no unreviewed safety analysis.

.M-1-1-86-066 - This Modification added RWCU piping flanges to the inlet and outlet of relief valve 1G33-2001/062 to allow for its removal for testing. There was no unreviewed safety analysis.

M-1-1-86-067 - This Modification changed the thermal overload heater relay FH28 with a FH35, due to its corresponding valve limitorque motor being oversized as bolt. There was no unreviewed safety analysis.

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C. LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for LaSalle Nuclear Power Station, Unit One, logged during the reporting period, November 1, through November 30, 1986. This information is provided pursuant to the reportable occurrence reporting requirements as set forth in 10CFR 50.73.

Licensee Event Report Number Date Title of Occurrence 86-040-00 11/18/86 "A" VC Ammonia Detector Trip due to degragation of the WISA Sample Pump.

86-041-00 11/12/86 Reactor Water Cleanup (RWCU) system isolated due to spurious actuations due to voltage oscillation when HFA Relay actuated during surveillance, i

i Document 0043r/0005r I

D. DATA TABULATIOblS The following data tabulations are presented in this report:

1. Operating Data Report
2. Average Daily Unit Power Level
3. Unit Shutdowns and Power Reductions J

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Document 0043r/0005r

1.* OPERATING DATA REPORT DOCKET NO. 050-373 UNIT LaSalle One DATE December 10, 1986 COMPLETED BY James P. Peters TELEPHONE (815)357-6761 OPERATING STATUS

1. REPORTING PERIOD: November, 1986 GROSS HOURS IN REPORTING PERIOD: 720
2. CURRENTLY AUTHORIZED POWER LEVEL (MWt):3323 MAX DEPEND CAPACITY (MWe-Net): 1036 DESIGN ELECTRICAL RATING (MWe-Net): 1078
3. POWER LEVEL TO WHICH RESTRICTED (IF ANY) (MWe-Net): 1050
4. REASONS FOR RESTRICTION (IF ANY): Administrative THIS MONTH YR TO DATE CUMULATIVE 5 NUMBER OF HOURS REACTOR WAS CRITICAL 720 1704.65 13743.65
6. REACTOR RESERVE SHUTDOWN HOURS 0.0 0.0 1642.0
7. HOURS GENERATOR ON LINE 720 1648.62 13290.62
8. UNIT RESERVE SHUTDOWN HOURS 0.0 0.0 0.0
9. GROSS THERMAL ENERGY GENERATED (MWH) 2188608 4514184 36727834
10. GROSS ELEC. ENERGY GENERATED (MWH) 740392 1501591 12000985
11. NET ELEC. ENERGY GENERATED (MWH) 715062 1353908 11358365
12. REACTOR SERVICE FACTOR 100% 21.3% 53.7%
13. REACTOR AVAILABILITY FACTOR 100% 21.3% 60.1%
14. UNIT SERVICE FACTOR 100% 20.5% 51.9%
15. UNIT AVAILABILITY FACTOR 100% 20.5% 51.9%
16. UNIT CAPACITY FACTOR (USING MDC) 95.8% 16.3% 42.9%
17. UNIT CAPACITY FACTOR (USING DESIGN MWe) 92.1% 23.5% 41.2%
18. UNIT FORCED OUTAGE RATE 0.0% 0.0% 15.6%
19. SHUTDOWNS CCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH)

NONE.

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

N/A.

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2,. AVERAGE DAILY UNIT POWER LEVEL DOCKET NO: 050-373 UNIT: LASALLE ONE DATE: December 10, 1986 COMPLETED BY: James P. Peters TELEPHONE: (815) 357-6761 MONTH: November, 1986 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)

1. 1011 17. 1050
2. 1023 18. 1046 3 1018 19. 1047
4. 1012 20. 945
5. 994 21. 1006
6. 1017 22, 859
7. 1038 23. 771
8. 1029 24. 755
9. 1033 25. 870
10. 1034 26. 1010
11. 1034 27. 1007
12. 1035 28. 1010
13. 1037 29. 999 14, 1038 30. 999
15. 1020 31. N/A
16. 1049 Document 0043r/0005r
3. UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 050-373' UNIT NAME Lase 11e One DATE December 10. 1986 REPORT MONTH NOVEMBER. 1986 COMPLETED BY James P. Peters TELEPHONE (815)357-6761 t

! METHOD OF TYPE SHUTTING DOWN F: FORCED DURATION THE REACTOR OR CORRECTIVE NO. DATE S: SCHEDULED (HOURS) REASON REDUCING POWER ACTIONS /COfMENTS I There were no unit shutdowns or power reductions during this reporting period.

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E. UNIQUE REPORTING REQUIREMENTS

1. Safety / Relief valve operations for Unit One.

VALVES NO & TYPE PLANT DESCRIPTION DATE ACTUATED ACTUATION CONDITION OF EVENT There were no Safety Relief Valves Operated for Unit One during this reporting period.

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2. ECCS Sy0tems OutEges The following outages were taken on ECCS Systems during the reporting period.

OUTAGE NO. BOUIPMENT PURPOSE OF OUTAGE 0-215-86 "0" DG "B" Air System Repairs Compressor 0-235-86 "0" DG "B" Refrigerated Prevent Overload Dryer of Compressor.

1-1582-86 "B" DG Air Replacer High Pressure Compressor. Valve Awsembly, i

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3. Off-Site Dose Calculation Manual There were no changes to the ODCM during this reporting period.
4. Radioactive Waste Treatment Systems.

Reference attached letter dated November 18, 1986.

5. Indications of Failed Fuel Elements There were no indications of Failed Fuel Elements during this reporting period.

Document 0043r/0005r

November 18, 1986 t

BACKGROUND i

During the past two years there have been many operational difficulties associated with Stock Equipment System used to solidify radweste. Now that most of the difficulties have been resolved radwaste solidification is proceeding in a timely manner. However during the interim a significant backlog of radwaste has accumulated hindering efficient operation of the's system as a whole. Additionally, vendors can offer a more cost effective ;

means for solidifying and dewartering various wastes. -

For the above reasons it has been decided to employ Westinghouse Hittman

!- Nuclear Inc. to dewater the contents of 1A/lB/2A/2B Phase Separator Tanks, the URC Sludge Tank, the Spent Resin Tank and the Waste Sludge Tank. The dewatering and subsequent transportation of waste off-site is expected to j begin Nov. 7, 1986. The contract with Westinghouse Hittman (Vendor) will run l through March 31, 1987.

Our Stock system has been modified to allow interfacing with the vendor's equipment. The changes, installed under Temporary System Change 1-1921-86, constitute major changes to the radwaste system and Licensee - initiated substitutions to the PCP END needing review under T.S. 6.9.1. and 6.7.2, j respectively.

i The contractor will be using his own equipment, personnel and procedures

to perform the dewatering. The dewatering will be performed in accordance with a NRC approved Topical Report (STD-R-05-011) and Process Control Program (STD-PCP-03-003) which meets the requirements of 10CPR20/50/61/71 and 100.

2 The vendor's procedures have been approved by the on Site review and

! Investigative Function (#86-50 and #86-51) for implementation at LaSalle j Station.

MODIFICATION DETAILS All vendor equipment is to be maintained within the Radwaste Truck Bay.

The dewatering will take place under both operating and health physics l supervision. Entry level status of the Truck bay will be upgraded at the discretion of the Health Physics department.

An interface with the existing loop transfer lines is needed to transfer waste to the vendor's equipment. Flanges existing on several lines and off l

two new " tees" all installed under Temporary System Change 1-1921-86 will be i opened and manual isolation valves will be flanged in-line. The various sludge lines are collected in a common header with another manual outlet isolation valve.

l DOCUMENT ID 2593r/

s An air supply, uncontaminated water supply and electrical hook-up are presently available in the truck bay.

No radioactive liquid or gaseous releases are expected. During the dewatering process air samples are taken at the outlet of the vacuum pump to determine if any notrelease of radioactive material occurs. If there is any release of radioactive materals, the outlet vent will be filtered and/or re-routed directly to the Radwaste Ventilation System, under the direction of the Tech Staff Engineer, resulting in no release.

Increased exposures to individuals in the unrestricted area and to the general population are not expected.

Exposure to plant personnel is expected to decrease 80% (35 man-rem) as a result of using a vendor's system. Actual exposure to plant personnel is monitored by the health physics department as. standard procedure.

Any comments or questions involving the above program should be addressed to John Arand, x-5*10.

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1 LASALLE NUCLEAR POWER STATION UNIT 2 ,

MONTHLY PERPORMANCE REPORT NOVEMBER, 1986 CONT 0NWFALTH EDISON COMPANY i

NRC DOCKET NO. 050-374 LICENSE NO. NPF-18 1

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DOCUMENT ID 0036r/000Sr

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,. TABLE OF CONTENTS I. INTRODUCTION II. MONTHLY REPORT FOR UNIT TWO A. Summary of Operating Experience B. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE

1. Amendments to Facility License or Technical Specifications
2. Facility or Procedure Changes Requiring NRC Approval
3. Tests and Experiments Requiring NRC Approval
4. Corrective Maintenance of Safety Related Equipment
5. Completed Safety Related Modifications C. LICENSEE EVENT REPORTS D. DATA TABULATIONS
1. Operating Data Report
2. Average Daily Unit Power Level
3. Unit Shutdowns and Power Reductions E. UNIQUE REPORTING REQUIREMENTS
1. Safety / Relief Valve Operaticns
2. ECCS System Outages
3. Off-Site Dose Calculation Manual Changes
4. Major Changes to Radioactive Waste Treatment System
5. Indications of Failed Fuel Elements DOCUMENT ID 0036r/0005r

,'I. INTRODUCTION The LaSalle County Nuclear Power Station is a two-unit facility owned by Commonwealth Edison Company and 1ccated near Marseilles, Illinois. Each unit is a Boiling Water Reactor with a designed net electrical output of 1078 Megawatts. Waste heat is rejected to a man-made cooling pond using the Illinois River for make-up and blowdown. The architect-engineer was Sargent and Lundy and the primary construction contractor was Commonwealth Edison Company.

Unit two was issued operating license number NPF-18 on December 16, 1983. Initial criticality was achieved on March 10, 1984 and commercial power operation was conunenced on June 19, 1984.

This report was compiled by James P. Peters, telephone number (815)357-6761 extension 325.

DOCUMENT ID 0036r/0005r

II. MONTHLY REPORT FOR UNIT TWO A ., ,

  • SUpstARY OF OPERATING EXPERIENCE FOR UNIT TWO November 1-31 November 1, 0000 Hours. The Unit Entered November with the Reactor Critical and On-Line at 83.5% (975 MWe).

November 7, 0602 Hours. Reactor Power Decreased to 79.1% (920 MWe) due to Control Rod 42-47 Drifting In.

November 7, 2330 Hours. Began Decreasing Power to 49% (570 MWe) due to Work on Control Rod Drive 42-47.

November 8, 0500 Hours. Began Increasing Power at 80 MWe/hr to 66%

(770 We).

November 8, 1435 Hours. Began Increasing Power at 10 MWe/hr to 85.1%

(990 MWe).

November 13, 1600 Hours. Began Decreasing Power to 77.4% (900 MWe) to investigate TCV Oscillations.

November 15, 1035 Hours. Began Decreasing Power to 72.2% (840 MWe) due to TCV oscillations.

November 30, 2400 Hours. Reactor and Generator on-line and coasting down in Power at 73.5% (808 MWe).

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B. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS AND SAFETY RELATED

,, MAINTENANCE.

1. Amendments to facility license or Technical Specification.

Amendment 29 - This revises the LaSalle County Station Unit #2 Technical Specifications to incorporate certain changes made to the administrative controls section of the Technical Specifications.

2. Facility or procedure changes requiring NRC approval.

There were no facility or procedure change requiring NRC approval during the reporting period.

3. Tests and experiments requiring NRC approval.

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

4. Corrective Maintenance of Safety Related Equipment.

The following table (Table 1) presents a summary of safety-related maintenance completed on Unit Two during the reporting period. The headings indicated in this sununary include: Work Request number, Component Name, cause of malfunction, results and effects on safe operation, and corrective action.

5. Completed Safety Related Modifications.

The following table (Table 2) presents a list of completed modifications during this reporting period. Each entry will have a short synopsis explaining details involved with each modification.

LOCUMENT ID 0036r/0005r

I TABLE 1

, CORRECTIVE MAINTENANCE OF .

SAFETY RELATED BQUIPMENT .

Unit #2 WORK REQUEST COMPONENT CAUSE OF MALFUNCTION RESULTS AND EFFECTS CORRECTIVE ACTION ON SAFE PLANT OPERATION L63294 LPRM-56-41D Bad Auxiliary Card LPRM is hard upscale and Replaced Auxiliary Card in Unit. bypassed. and Recalibrated.

L63311 DIV-1, POST LOCA Out of Calibration Contradictory Indication Reca11brated control Room. Instruments.

L63382 "C" MSL Radiation Loose Spring across Drawer giving 1/2 Scram Removed Spring.

! Monitor 2D18-K610C Terminal Strip. and Isolations.

DOCUMENT ID 0036r/000Sr

TABLE 2 COMPLETED SAFETY RELATED MODIFICATIONS MODIFICATION NUMBER: A brief Synopsis of Incorporated Modification objectives with final design resolution. Also, state reviewed or unreviewed safety questions.

The following list below has all the Safety Regulated Modifications completed from January 1, 1986, through November 30, 1986.

M-1-2-85-033 - This Modification changed the Unit #2 SBGT electric heater breaker thermal and magnetic trip settings. The heaters were recently installed and draw more current per design. There was no unreviewed safety questions.

M-1-2-85-062 - This Modification changed the "2B" diesel generator magnetic setting setpoint from 6 to 6.5 per extensive testing results.

There was no unreviewed safety questions.

M-1-2-85-069 - This Modification changed the thermal overload heater relay FH28 with a FH35, due to its corresponding valve limitorque motor being oversized as built. There was no unreviewed safety question.

M-1-2-86-035 - This Modification changed the support arrangement (2RI-A3), thus creating adequate thermal flexibility and brings the piping arrangement within code limits. There was no unreviewed safety question.

DOCUMENT ID 0036r/0005r

..C. LICENSEE EVENT REPORTS The following is a tabular sununary of all licensee event reports for LaSalle Nuclear Power Station, Unit Two, logged during the reporting period, Noventer 1 through November 30, 1986. This information is provided pursuant to the reportable occurrence reporting requirements as set forth in 10CFR 50.~13.

Licensee Event Report Number Date Title of Occurrence There were no Licensee Event Reports submitted during this reporting period.

DOCUMENT ID 0036r/0005r

D. DATA TABULATIONS The following data tabulations are presented in this report:

1. Operating Data Report
2. Average Daily Unit Power Level

'. Unit Shutdowns and Power Reductions 2

DOCUMSNT ID 0036r/0005r

1. ObERATING DATA REPORT DOCKET NO. 050-374

. UNIT LaSalle Two DATE December 10, 1986 COMPLETED BY James P. Peters TELEPHONE (815)357-6761 OPERATING STATUS

1. REPORTING PERIOD: November, 1986 GROSS HOURS IN REPORTING PERIOD: 720
2. CURRENTLY AUTHORIZED POWER LEVEL (MWt):3323 MAX DEPEND CAPACITY (MWe-Net): 1036 DESIGN ELECTRICAL RATING (MWe-Net): 1078
3. POWER LEVEL TO WHICH RESTRICTED (IF ANY) (MWe-Net): 982
4. REASONS FOR RESTRICTION (IF ANY): Administrative THIS MONTH YR TO DATE CUMULATIVE 5 NUMBER OF HOURS REACTOR WAS CRITICAL 720 5870.05 11259.45
6. REACTOR RESERVE SHUTDOWN HOURS 0.0 29.83 29.83
7. HOURS GENERATOR ON LINE 720 5790.62 11026.92
8. UNIT RESERVE SHUTDOWN HOURS 0.0 0.0 0.0
9. GROSS THERMAL ENERGY GENERATED (MWH) 1886136 16215696 31724248
10. GROSS ELEC. ENERGY GENERATED (MWH) 625703 5360249 10470318
11. NET HLEC. ENERGY GENERATED (MWH) 601895 _5150239 9973454
12. REACTOR SERVICE FACTOR 100% 73.2% 60.6%
13. REACTOR AVAILABILITY FACTOR 100% 73.6% 60.7%
14. UNIT SERVICE FACTOR 100% 72.2% 59.4%
15. UNIT AVAILABILITY FACTOR 100% 72.2% 59.4%
16. UNIT CAPACITY FACTOR (USING MDC) 80.7% 62.1% 51.8%
17. UNIT CAPACITY FACTOR (USING DESIGN MWe) 77.5% 59.6% 49.8%
18. UNIT FORCED OUTAGE RATE 0.0% 27.8% 27.6%
19. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, AND DURATION OF EACH):

The first refueling outage is scheduled to begin January 3, 1987.

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

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2. AVERAGE DAILY UNIT POWER LEVEL DOCKET NO: 050-374 UNIT: LASALLE TWO DATE: December 10, 1986 COMPLETED BY: James P. Peters TELEPHONE: (815) 357-6761 MONTH: October 1986 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)
1. 929 17. 809
2. 918 18. 806 3 913 19. 802
4. 910 20. 798
5. 904 21. 794
6. 903 22. 790
7. 882 23. 785
8. 691 24. 779
9. 859 25. 776 10 . _, 943 26. 771
11. 935 27. 766
12. 930 28. 763
13. 906 29. 759 14, 860 30. 755
15. 832 31. N/A
16. 811 DOCUMENT ID 0036r/0005r
3. UNIT SHUTDOWNS AND POWER CEDUCTIONS DOCKET NO. 050-374 UNIT NAfW LaSalle %

DATE December 10. L986 REPORT MONTH NOVEMBER. 1986 COMPLETED BY James Peters TELEPHONE (815)357-6761 METHOD OF TYPE SHUTTING DOWN F: FORCED DURATION THE REACTOR OR CORRECTIVE NO. DATE S: SCHEDULED (HOURS) REASON REDUCING POWER ACTIONS /COPWENTS Th2re were no unit shutdowns or power reductions during this reporting period.

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, E. UNIQUE REPORTING RE(gUIREMENTS

1. Safety / Relief Valve Operaticns for Unit Two.

DATE VALVES NO & TYPE PLANT DESCRIPTION ACTUATED ACTUATIONS CONDITIM OF EVENT There were 1.o safety relief valves operated for Unit #2 during this reporting period.

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, 2. ECCS Sy2tems Outage 2 o The following outages were taken on ECCS Systems during the reporting period.

OUTAGE NO. BOUIPMENT PURPOSE OF OUTAGE 2-626-86 RCIC Turb. Steam Supply Change Lubrication Oil.

Valve 2E51-F045 2-628-86 "B" RHR Service Water Valve Repair Valve.

2E12-F332D.

2-635-86 2E12-F047A Troubleshooting.

2-636-86 RCIC System Repair Pump Suction Relief Valve 2-638-86 "B" DG Change Crankcase Pressure Switch.

2-639-86 2VG01C Work Request L47422.

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DOCUMENT ID 0036r/0005r 1

3. Off-Sito Dosa Calculation Manuni

,' There were no changes to the ODCM during this reporting period.

4. Radioactive Waste Treatment Systems.

Reference Attached letter dated November 18, 1986.

5. Indications of Failed Fuel Elements.

There were no indications of failed fuel elements during this reporting period.

l DOCUMENT ID 0036r/0005r

e o November 18, 1986 BACKGROUND During the past two years there have been many operational difficulties associated with Stock Equipment System used to solidify radwaste. Now that most of the difficulties have been resolved radwaste solidification is proceeding in a timely manner. However during the interim a significant backlog of radwaste has accumulated hindering efficient operation of the system as a whole. Additionally, vendors can offer a more cost effective means for solidifying and dewartering various wastes.

For the above reasons it has been decided to employ Westinghouse Hittman

> Nuclear Inc. to dewater the contents of 1A/lB/2A/2B Phase Separator Tanks, the URC Sludge Tank, the Spent Resin Tank and the waste Sludge Tank. The dewatering and subsequent transportation of waste off-site is expected to begin Nov. 7, 1986. The contract with Westinghouse Hittman (Vendor) will run through March 31, 1987.

Our Stock system has been modified to allow interfacing with the vendor's equipment. The changes, installed under Temporary System Change 1-1921-86, constitute major changes to the radwaste system and Licensee - initiated substitutions to the PCP EMD needing review under T.S. 6.9.1. and 6.7.2, respectively.

The contractor will be using his own equipment, personnel and procedures to perform the dewatering. The dewatering will be performed in accordance with a NRC approved Topical Report (STD-R-05-Oll) and Process Control Program (STD-PCP-03-003) which meets the requirements of 10CFR20/50/61/71 and 100.

The vendor's procedures have been approved by the On Site review and Investigative Function (#86-50 and #86-51) for Laplementation at LaSalle Station.

MODIFICATION DETAILS All vendor equipment is to be maintained within the Radwaste Truck Bay,

.i The dewatering will take place under both operating and health physics supervision. Entry level status of the Truck bay will be upgraded at the discretion of the Health Physics department.

An interface with the existing loop transfer lines is needed to transfer waste to the vendor's equipment. Flanges existing on several lines and off two new " tees" all installed under Temporary System Change 1-1921-86 will be opened and manual isolation valves will be flanged in-line. The various sludge lines are collected in a common header with another manual outlet isolation valve.

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e An air supply, uncontaminated water supply and electrical hook-up are presently available in the truck. bay.

No radioactive liquid or gaseous releases are expected. During the dewatering process air samples are taken at the outlet of the vacuum pump to determine if any notrelease of radioactive material occurs. If there is any release of radioactive materals, the outlet vent will be filtered and/or re-routed directly to the Radwaste Ventilation System, under the direction of the Tech Staff Engineer, resulting in no release.

Increased exposures to individuals in the unrestricted area and to the general population are not expected.

Exposure to plant personnel is expected to decrease 80% (35 man-rem) as a result of using a vendor's system. Actual exposure to plant personnel is monitored by the health physics department as standard procedure.

Any comments or questions involving the above progran should be addressed to John Arand, x-5'70.

DOCUMENT ID 2593r/

O)e Commonwe:lth r Rural Route #1, Box 220 Edison LaSalls County Nuclear Station

[ _%' / Marseilles, Illinois 61341

,g Telephone 815/357 6761 December 8, 1986 Director, Office of Management Information and Program Control United States Nuclear Regulatory Conunission Washington, D.C. 20555 ATTN: Document Control Desk Gentlemen:

Enclosed for your information is the monthly performance report covering LaSalle County Nuclear Power Station for the period November 1, 1986 through November 31, 1986.

Very truly yours, CL

'G.' E. D ederich S ation Manager LaSalle County Station GJD/JPP/jdp Enclosure xc: J. G. Keppler, NRC, Region III l

NRC Resident Inspector LaSalle Gary Wright, Ill. Dept. of Nuclear Safety A. Bournia, NRR Project Manager D. P. Galle, CECO D. L. Farrar, CECO INPO Records Center L. J. Anastasia, PIP Coordinator SNED M. A. Ortin, GE Resident H. E. Bliss, Nuclear Fuel Services Manager l

C. F. Dillon, Senior Financial Coordinator, LaSalle Central File f/f Document 0043r/0005r )\ I l

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