ML20235F888

Monthly Operating Repts for Jan 1989 for Quad-Cities Nuclear Power Station Units 1 & 2

ML20235F888
Person / Time
Site:	Quad Cities
Issue date:	01/31/1989
From:	Deelsnyder L, Robey R COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
RAR-89-04, RAR-89-4, NUDOCS 8902230043
Download: ML20235F888 (26)

Text

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QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2.

MONTHLY PERFORMANCE REPORT-JANUARY, 1989 COMMONWEALTH EDISON COMPANY AND IONA-ILLINOIS GAS & ELECTRIC COMPANY NRC DOCKET NOS. 50-254 AND 50-265 LICENSE NOS. DPR-29 AND DPR-30 l

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l-TABLE OF CONTENTS I. Introduction II. Summary of Operating Experience A. Unit One B. Unit Two III. 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 Equipment 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 Steam Relief Valve Operations B. Control Rod Drive Scram Timing Data VII. Refueling Information VIII. Glossary 0027H/0061Z-

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i I. INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Boiling Water Reactors, each with a Maximum Dependable Capacity of 769 MWe Net, located in Cordova, Illinois. The Station is jointly owned by Commonwealth Edison Company and Iowa-Illinois Gas & Electric Company. The Nuclear Steam Supply-Systems are General Electric Company Bolling Water Reactors. The Architect / Engineer was Sargent & Lundy, Incorporated, and the primary construction contractor was United Engineers & Constructors. The Mississippi River is the condenser cooling water source. The plant is subject to license numbers DPR-29 and DPR-30, issued October 1, 1971, and March 21, 1972, respectively; pursuant to Docket Numbers 50-254 and 50-265. The date of initial Reactor criticalities for Units One and Two, respectively were October 18, 1971, and April 26, 1972. Commercial generation of power began on February 18, 1973 for Unit One and March 10, 1973 for Unit Two.

This report was compiled by Lynne Deelsnyder and Verna Koselka, telephone number 309-654-2241, extensions 2185 and 2240.

0027H/0061Z

l II.

SUMMARY

OF 0PERATING EXPERIENCE 1

A.' Unit One Unit' One began the month of' January operating in Economic Generation Control (EGC). On January 2, at 0305 hours0.00353 days <br />0.0847 hours <br />5.042989e-4 weeks <br />1.160525e-4 months <br />, the unit was taken off EGC to reverse condenser flow. A load increase to 800 MWe was taken with recirculation pumps. At 0420 hours0.00486 days <br />0.117 hours <br />6.944444e-4 weeks <br />1.5981e-4 months <br />, the unit was placed in EGC. The unit remained in EGC with minor interruptions to perform routine surveillance until January 9. At 0820 hours0.00949 days <br />0.228 hours <br />0.00136 weeks <br />3.1201e-4 months <br />, the unit was taken off EGC and power increase to full load was taken. Power levels were held constant while Traversing Incore Probe Sets were performed. On January 10, at 2035 hours0.0236 days <br />0.565 hours <br />0.00336 weeks <br />7.743175e-4 months <br />, all sur-veillances were'successfully completed and the unit was placed in EGC. The unit remained in EGC until January 13. At 2225 hours0.0258 days <br />0.618 hours <br />0.00368 weeks <br />8.466125e-4 months <br />, a power reduction.

to 300 MWe was taken at the request of the Chicago Load Dispatcher. . Power levels were held constant until January 15. At 0555 hours0.00642 days <br />0.154 hours <br />9.176587e-4 weeks <br />2.111775e-4 months <br />, the Chicago Load Dispatcher requested full power. At 0950. hours, 822 MWe was achieved. Power levels were held constant until January 16. At 1618 hours0.0187 days <br />0.449 hours <br />0.00268 weeks <br />6.15649e-4 months <br />, power levels were adjusted and the unit was placed in ECC. The unit remained in EGC until January 21 when the unit was taken off EGC and a power reduction to 500 MWe-was taken to perform a deep / shallow exchange with control rods. On January 22, at 0635 hours0.00735 days <br />0.176 hours <br />0.00105 weeks <br />2.416175e-4 months <br />, a load ~ increase to 780 MWe was taken and at 1750 hours0.0203 days <br />0.486 hours <br />0.00289 weeks <br />6.65875e-4 months <br /> another load increase to full power was taken. At 1820 hours0.0211 days <br />0.506 hours <br />0.00301 weeks <br />6.9251e-4 months <br />, a power reduction to 800 MWe was taken due to high level alarms on Local Power Range Monitor 32-41B. Power levels were held constant while an Average Power Range Monitor Heat Balance surveillance was performed. At 0815 hours0.00943 days <br />0.226 hours <br />0.00135 weeks <br />3.101075e-4 months <br />, the surveillance was successfully completed and power levels were held constant until January 24. At 1915 hours0.0222 days <br />0.532 hours <br />0.00317 weeks <br />7.286575e-4 months <br />, power levels were adjusted and the unit was placed in EGC. The unit remained in EGC until January 27. At 0034 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />, EGC was tripped for core flow calculations. At 1153 hours0.0133 days <br />0.32 hours <br />0.00191 weeks <br />4.387165e-4 months <br />, calculations were completed and the unit was placed in EGC. At 1657 hours0.0192 days <br />0.46 hours <br />0.00274 weeks <br />6.304885e-4 months <br />, the unit was taken off EGC and a load increase to full power was taken with control rods. At 1725 hours0.02 days <br />0.479 hours <br />0.00285 weeks <br />6.563625e-4 months <br />, 822 MWe was achieved. Full power was held until January 28.

At 0108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br />, power levels were adjusted and the unit was placed in EGC.

For the remainder of the month, the unit remained in EGC or operated near full power.

B. Unit Two Unit Two began the month of January operating in Economic Generation Control (EGC). On January 2, at 2310 hours0.0267 days <br />0.642 hours <br />0.00382 weeks <br />8.78955e-4 months <br />, the unit was taken off EGC, and a power reduction to 400 MWe was taken to perform a control rod sequence exchange.

At 0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br />, on January 3, this was successfully completed, and a load increase to full power was begun with control rods. 810 MWe was achieved at 0727 hours0.00841 days <br />0.202 hours <br />0.0012 weeks <br />2.766235e-4 months <br />, and power levels were held constant until January 4 when

ThaversingIncoreProbeSetswereperformed. At 1814 hours0.021 days <br />0.504 hours <br />0.003 weeks <br />6.90227e-4 months <br />, all surveillance were completed and.the unit was placed in EGC. The unit remained in EGC' until January 5. At 0943 hours0.0109 days <br />0.262 hours <br />0.00156 weeks <br />3.588115e-4 months <br />, the unit was taken off EGC to perform a test on the turbine #1 combined intermediate valve. At 1438 hours0.0166 days <br />0.399 hours <br />0.00238 weeks <br />5.47159e-4 months <br />, the testing was completed and the unit was placed in EGC. At 1505 hours0.0174 days <br />0.418 hours <br />0.00249 weeks <br />5.726525e-4 months <br />, the unit was taken off.EGC and another test was performed on the #1 CIV. At 2215 hours0.0256 days <br />0.615 hours <br />0.00366 weeks <br />8.428075e-4 months <br />, the testing was completed, and the unit was placed in EGC. On January 6, at 2200 hours0.0255 days <br />0.611 hours <br />0.00364 weeks <br />8.371e-4 months <br />, EGC was tripped. A power reduction 317 MWe was taken while repairs were made on the #1 combined intermediate valve due to continuous problems. A drywell entry was also made at this time to add oil to the 2B recirculation pump. On January 7, at 0935 hours0.0108 days <br />0.26 hours <br />0.00155 weeks <br />3.557675e-4 months <br />, maintenance was completed on the valve, and an ascent to full power was begun with control rods. At 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br />, 813 MWe was achieved. On January 8, at 1905 hours0.022 days <br />0.529 hours <br />0.00315 weeks <br />7.248525e-4 months <br />, the unit was placed in EGC. The unit remained in EGC until January 9. At 1111 hours0.0129 days <br />0.309 hours <br />0.00184 weeks <br />4.227355e-4 months <br />, the unit was taken off EGC to perform routine surveillance.

At 1708 hours0.0198 days <br />0.474 hours <br />0.00282 weeks <br />6.49894e-4 months <br />, all surveillance were successfully completed, and-the unit was placed in ECC. On January 10, at 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />, the. unit was taken off EGC, and a power reduction to 600 MWe was taken at the request of the Chicago Load Dispatcher. At 0630 hours0.00729 days <br />0.175 hours <br />0.00104 weeks <br />2.39715e-4 months <br />, a load increase was taken per the Load Dispatcher, and the unit was placed in EGC. On January 11, at 0004 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, EGC was tripped and a power reduction to 463 MWe was taken at.the request of-the Load Dispatcher. At 0735 hours0.00851 days <br />0.204 hours <br />0.00122 weeks <br />2.796675e-4 months <br />, a load increase to full power was j taken and at 2341 hours0.0271 days <br />0.65 hours <br />0.00387 weeks <br />8.907505e-4 months <br />, the unit was placed in EGC. On January 12, at 0325 hours0.00376 days <br />0.0903 hours <br />5.373677e-4 weeks <br />1.236625e-4 months <br />, a power reduction to 545 MWe was taken at the request of the Load Dispatcher. At 0530 hours0.00613 days <br />0.147 hours <br />8.763227e-4 weeks <br />2.01665e-4 months <br />, a load increase to full power was taken and at 2035 hours0.0236 days <br />0.565 hours <br />0.00336 weeks <br />7.743175e-4 months <br />, the unit was placed in EGC. From January 13 thru the end of the month, normal operational activities and surveillance occurred while-the unit operated in EGC or remained near full power.

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III. PLANT OR PROCEDURE ~ CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A. Amendments to Facility License or Technical Specifications' There were'no Amendments to the Facility License or Technical =

Specifications forsthe reporting period.

B. . Facility or Procedure Changes Requiring NRC Approval There were no Facility or Procedure changes requiring NRC approval for the reporting period.

C. Tests and Experiments Requiring NRC Approval There were no Tests or Experiments requiring NRC approval:for l the reporting period.

D. Corrective Maintenance of Safety Related Equipment The following represents a tabular summary of the major safety related maintenance performed on Units One and.Two during the reporting period. This summary includes the following: -Work Request Numbers, Licensee Event Report Numbers, Components, Cause of Malfunctions, Results and Effects on Safe Operation, and Action Taken to Prevent Repetition.

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UNIT 1 MAINTENANCE

SUMMARY

WORK REQUEST No.: Q59815 LER NUMBER: NA COMPONENT: System 5700 - Line 1-10119D-3/8"-0 is broken off in union at cooler.

1A RHR S.W. Pump Cubicle Cooler.

CAUSE OF MALFUNCTION: The cause was due to a 3/8 inch pipe nipple on the high side drain line which had failed and separated from a pipe union. This failure was due to corrosion and wear of the pipe threads aggravated by metal fatigue caused by line vibration during pump operation.

RESULTS & EFFECTS ON SAFE OPERATION: The safety significance of this event was minimal because all testing required by Technical Specifications was completed to verify the remaining three RHR service water pumps operable. This was acceptable per Technical Specifications.

ACTION TAKEN TO PREVENT REPETITION: To prevent reoccurrence, the Technical Staff visually inspected the same pipe nipples on the remaining seven RHR service water pump coolers. They also inspected the pipe nipples on the Diesel Generator cooling water pump room coolers which are subject to the same environmental.

conditions. None of the nipples inspected showed signs of imminent failure and were deemed to be acceptable.

WORK REQUEST NO.: Q60207 LER NUMBER: 87-016 COMPONENT: System 2300 - HPCI steam exhaust check valve leaking more than acceptable. Replace EPN 1-2301-45.

CAUSE OF MALFUNCTION: The cause of the failure was determined to be steam errosion of the seating material.

RESULTS & EFFECTS ON SAFE OPERATION: This event has minimal safety consequences since the total leakage determined by local leak rate testing does not represent a probable leakage from the primary containment during accident condition.

ACTION TAKEN TO PREVENT REPETITION: To prevent reoccurrence, the valve was replaced with a Marlin valve (serial no. 3769). The Marlin valve differs in that the Marlin design protects the seating material against steam impingement to a greater degree than the old valve. The station feels that the Marlin valve should improve the LLRT performance of this valve. l

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• D- ' WORK REQUEST No : ;Q60828 .

.LER NUMBER: 87-016

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COMPONENT: . System 1600 - Nitrogen make-up supply valve leaks excessively.

Valve 1-1601-58.

CAUSE OF MALFUNCTION: -Cause of malfunction was due to;a dirty seat ~ that

. was in need of lapping.- This condition was due to the build-up of foreign material on the seat over many years of operation.

- RESULTS'& EFFECTS ON SAFE OPERATION: This event has minimal safety conse-quences since the total. leakage determined by local leak rate testing does not represent a probable leakage from the primary containment during accident condition.

ACTION TAKEN TO PREVENT REPETITION: The valve was repacked and the seating surfaces lapped. Air operator was also rebuilt. This Js the first. incident involving a high leakage rate for these valves since plant began operation.

It is not considered a chronic problem. No further action is required.

WORK REQUEST NO.: Q61063 LER NUMBER: N/A

- COMPONENT: System 0201 - Weld overlay weld 02M-S4 and surface finish. "M" reactor jet pump riser EPN 1-0201M-12A.

CAUSE OF MALFUNCTION: The cause of the crack indications is postulated to be Intergranular Stress Corrosion Cracking (IGSCC). The materials used fn the original piping and fitting are regular Grade Type 304 stainless steel which is known to be susceptible to stress corrosion cracking based on plant operating history.

RESULTS & EFFECTS ON SAFE OPERATION: The effects on safe operation are mini-mal because crack indications of this type tend to propogate at a slow rate.

Therefore, a 100 percent throughwall crack would be easily detected using existing primary containment leakage monitoring systems before a complete failure would occur.

ACTION TAKEN TO PREVENT REPETITION: The weld was repaired with " full structural"

. design overlay. The overlay was then surface finished to permit application of the EPRI technique for overlay ultrasonic examination. The overlay was found to have sound overlay metal of sufficient thickness to meet the full structural design criteria.

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• WORK REQUEST No.: Q61212 LER NUMBER: 87-016 1 COMPONENT: System 1000 - RHR drywell spray isolation valve leaking excessively.

Repair EPN M0-1-1001-26A.

CAUSE OF MALFUNCTION: The disc and seat were discovered to be corroded and scratched.

RESULTS & EFFECTS ON SAFE OPERATION: This event has minimal safety consequences since the total leakage determined by local leak rate testing does not represent-a probable leakage from the primary containment during accident condition.

ACTION TAKEN TO PREVENT REPETITION: The valve seat was cleaned and lapped, and the disc wa.c ground. Subsequent leak rate tests also failed. Work Request Q62372 was initiated to replace the spring cartridge cap with a like-for-like replacement. The valve was tested again and produced results similar to the "as found" test. The station felt that no further action could be taken until the next refueling outage. This valve will be disassembled and inspected during the next Unit 1 refueling outage to further attempt to discover the cause of the increased leakage.

WORK REQUEST No.: Q61355, Q61356, Q61357, Q61359 and Q61360 LER NUMBER: N/A COMPONENT: System 1400 - Weld overlay weld numbers 14B-S8, 14B-F2, 14A-F2, lAA-S9 and 14A-F11. Core spray line (A and B loop).

CAUSE OF MALFUNCTION: The cause of the crack indications is postulated to be Intergranular Stress Corrosion Cracking (IGSCC). The materials used in the original piping and fitting are regular Grade Type 304 stainless steel which is known to be susceptible to stress corrosion cracking based on plant operating history.

RESULTS & EFFECTS ON SAFE OPERATION: The effects on safe operation are minimal because crack indications of this type tend to propogate at a slow rate.

Therefore, a 100 percent throughwall crack would be easily detected using existing primary containment leakage monitoring systems before a complete failure would occur.

ACTION TAKEN TO PREVENT REPETITION: The weld was repaired with a " leak barrier" design overlay. The overlay was then surface finished to perform ultrasonic bonding inspection. The weld was found to have sound overlay material.

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' WORK REQUEST No.: Q61560 LER NUMBER: N/A

-COMPONENT: System 201 - Weld overlay weld #02F-S3. "F" Reactor Jet Pump Riser.

CAUSE OF MALFUNCTION: The cause of'the crack indications is-postulated to be Intergranular Stress Corrosion Cracking (IGSCC). The materials used in the original piping and fitting are regular Grade Type 304 stainless steel

.which is known to be susceptible to stress corrosion cracking based'on plant operating-history.

RESULTS & EFFECTS ON SAFE OPERATION: The effects on safe operation are mini-mal because crack indications of this type tend to propogate at a slow rate.

Therefore, a 100 percent throughwall crack would be easily detected using existing primary containment leakage monitoring systems before a complete failure would occur.

ACTION TAKEN TO PREVENT REPETITION: The weld was repaired with " full structural" design overlay. The overlay was then surface finished to permit application of the EPRI technique for overlay ultrasonic examination. The overlay was found to.have sound overlay metal of sufficient r',.ekness to meet the full structural design criteria.

WORK REQUEST NO.: Q61561

'LER NUMBER: N/A COMPONENT: System 201 - Weld overlay weld #02B-F1. Reactor Recire. System Ring Header (B Loop).

CAUSE OF MALFUNCTION: The cause of the crack indications is postulated to be Intergranular Stress Corrosion Cracking (IGSCC). The materials used in the original piping and fitting are regular Grade Type 304 stainless steel which is known to be susceptible to stress corrosion cracking based on plant operating history.

RESULTS & EFFECTS ON SAFE OPERATION: The effects on safe operation are mini-mal because crack indications of this type tend to propogate at a slow rate.

Therefore, a 100 percent throughwall crack would be easily detected using existing primary containment leakage monitoring systems before a complete failure would occur.

ACTION TAKFN TO PREVENT REPETITION: The weld was repaired with a " leak barrier" design overlay. The overlay was then surface finished to perform ultrasonic-bonding inspection. The weld was found to have sound overlay material.

' WORK REQUEST NO.: Q61562, Q61563 1

LER NUMBER: N/A COMPONENT: System 201 - Weld overlay weld #02C-S3 ("C" Reactor Jet Pump Riser) and #02E-S3 ("E" Reactor Jet Pump Riser).

CAUSE OF MALFUNCTION: The cause of the crack indications is postulated to be Intergranular Stress Corrosion Cracking (IGSCC). The materials used in the original piping and fitting are regular Grade Type 304 stainless steel which is known to be susceptible to stress corrosion cracking based on plant operating history.

RESULTS & EFFECTS ON SAFE OPERATION: The effects on safe operation are mini-mal because crack indications of this type tend to propogate at a slow rate.

Therefore,'a 100 percent throughwall crack would be easily detected using existing primary containment leakage monitoring systems before a complete failure would occur.

ACTION TAKEN TO PREVENT REPETITION: The weld was repaired with a " leak barrier" design overlay. The overlay was then Lurface finished to perform ultrasonic bonding inspection. The weld was found to have sound overlay material.

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UNIT 2 MAINTENANCE

SUMMARY

WORK REQUEST NO.: Q58082 l.

LER NUMBER: 87-07 COMPONENT: System 7800 - Trip tested MCC 28-3 main feed breaker. (2-7800-28-3).

CAUSE OF MALFUNCTION: The apparent cause was attributed to an electrical storm.

It is suspected that a bolt of lightning hit a meteorological tower northeast of'the plant which is fed from MCC 28-3. The power surge backfed to Bus 28 causing the main breaker for that bus to trip.

RESULTS & EFFECTS ON SAFE OPERATION: The effects on safe operation were minimal because the RWCU system isolation which occurred does not effect the safe operation of the reactor.

ACTION TAKEN TO PRE 7ENT REPETITION: The station electrical engineering depart-ment evaluated the trip setting coordination of the main feed circuit breakers for MCC 28-3 and Bus 28. The results indicated that the coordination was proper.

The MCC 28-3 field breaker was trip tested under this work order and proper operation resulted.

WORK REQUEST No.: Q58085 LER NUMBER: 87-07 COMPONENT: System 7100 - Trip tested Bus 28 main field breaker.

CAUSE OF MALFUNCTION: The apparent cause was attributed to an electrical storm.

It is suspected that a bolt of lightning hit a meteorological tower northeast of the plant which is fed from MCC 28-3. The power surge backfed to Bus 28 causing the main breaker for that bus to trip.

RESULTS & EFFECTS ON SAFE OPERATION: The effects on safe operation were minimal because the RWCU system isolation which occurred does not effect the safe operation of the reactor.

ACTION TAKEN TO PREVENT REPETITION: The station electrical engineering depart-ment evaluated the trip setting coordination of the main feed circuit breakers for MCC 28-3 and Bus 28. The results indicated that the coordination was proper.

The Bus 28 main field breaker was successfully trip tested under this work order.

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IV. LICENSEE EVENT REPORTS l l

The-following is a tabular summary of allLlicensee event. reports.for-

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Quad-Cities Units One and Two occurring during the reporting' period,.

pursuant to the reportable occurrence reporting requirements as set forth in sections 6.6.B.1. and 6.6.B.2. of the Technical Specifications.

-UNIT 1 Licensee Event Report Number Date Title of decurrence 89-001 1-06-89 RCIC Inop, 1-1301-48; valve 89-002 1-09-89 Total Combined LLRT Interval for A0;1-203-1A Exceeded.3.25 Times the Specified Surveillance >

Interval UNIT 2 There were no Licensee Event Reports for the month of January, 1989 for Unit 2.

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

A. Operating Data Report B. Average Daily Unit Power Level C. Unit Shutdowns and Power Reductions 1

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APPENOlX C OPERATING DATA REPOR7 -

DOCKET NO. 50-254 UNIT One DATg February 8, 1989' COMPLETEO SY Lynne Deelsnyder TELEPHONg 309-654-2241

_l OrtitATING STAM 0000 010189 2400 013189 GROSS HOURS IN REPORTING PERICO-

1. REPORTING PERICO:

2511 MAX. DEPENO. CAPACITY tesNo.fiset: 769

2. CURRENTLY AUTHORIEED POWER LEVEL (Munt: W1 DESIGN ELECTRICAL RATING (efNo.Ned:

N/A

3. POWER LEVEL TO WHICN REETRICTED (IP ANYI (emus Neti:

4 REASONS POR RESTRICTION (IP ANY1:

THIS MONTN YR TO DATE CutfULATIVE 744.0 118286.2-S. NuteSER OF HOURS REACTOR WAS CRITICAL . . . . . . . . . . . . . . 744. 0 0.0 0.0 3421.9_

E. REACTOR RESERVE SMUTOOWN MOURE . . . . . . . . . . . . . . . . . . .

. 744.0 744.0 114403.2

7. MOURS OENERATOR ON UNE . . . . . . . . . . . . . . . . . . . . . . .

0.0 0.0 -909.2 E. UNIT RESERVE SMUT 00WN MOURS . . . . . . . . . . . . . . . . . . . . . .

1734065 243424144

9. GROSS THERMAL ENERGY GENERATED (MWHI ............. 1734065

....... 566230 566230 78923843

10. GROSE ELECTRICAL ENERGY GENERATED (MWHI . . . .

542719 74107041

11. NET ELECTRICAL ENERGY GENERATED (MWHI ............. 542719 1 0.0 100.0 80.7

12. r EACTOR SERVICE P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . . .

100.0 100.0 83.0

13. REACTOR AV AILASILITY P ACTOR . . . . . . . . . . . . . . . .

0.0 100.0 78.0

14. UNIT SERVICE P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

100.0 78.6

15. UNIT AV AILASILITY P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . 100. 0 94.9 94.9 6 5. 7 __

18. UNIT CAPACITY P ACTOR tusinE MCCI....................

92.5 92.5 64.1

17. UNIT CAPACITY P ACTOR (UsanE Demic MWel . . . . . . . . . . . . . . . . .

oo o.o s, iE. UNif PORC50 0uTAGE RATE . . . . . . . . . . . . . . . . . . . ....

19. SHUTOOWNS SCHEDULE 0 CVER NEXT E MONTHS ITYPE. DATE. ANO OURATION OF EACM):

20. IF SMUT 00WN AT END OF REPORT PERIOD, ESTIMATED DATE OP STARTUP:

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

PCRECAST ACMfEVED INITI AL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERA 7800f l

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1 APPENOtX C  ;

. OPERATING DATA REPORT l I

OOCKET NO. _ 50-265 UNIT Tun OATg February 8, 1989

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COMPT.ETED BY Lynne Deelsnyder i TELEPHONg 309-654-2241 l ]

OMAATINGSTATUS 0000 010189

'2400 013189 744

1. REPORTiesG PER100: GROGE MOURG IN REPORTINO PEnl00:

max. OEptNo. CAPACITY (MWo.Ned. 769 S. CURRENTLY AUTHORISEO POWER LEVEL (impfg 2511

/09 00880N ELECTRICAL RATING (MWs.Ned:

N/A

3. POWER LEVEL TO WMcCH REETRICTED llP ANY) IMWo.Ned:

4. REASONE POR RESTRICTION (IP ANYh THIS MONTN YR 70 OATE CUMULATfv6

'744.0 111693.9-

8. NUMEGR OF MOURE REACTOR WAE CRITICAL . . . . . . . . . . . . . . 744. 0 0.0 0.0 2985.8_

E. REACTOR REEERVE EMUTOOWN MOURE . . . . . . . . . . . . . . . . . . .

744.0 108475.7

f. MOURE GENERATOR ON LINE . . . . . . . . . . . . . . . . . . . . . . . . . 7 4 4. 0 0.0 0.0 707.9 E. UNIT RESERVE SMUTOOWN MOURE . . . . . . . . . . . . . . . . . . . . . .

1720714 232630987

9. GROGE THERMAL ENEROY GENERATED (MWHI ............. 1720714 563232 74496703

10. GROGE ELECTRICAL ENERGY GENERATED (MWHI . . . . . . . . ... 563232 540033 540033 70276610

11. NET ELECTRICAL ENERGY GENERATED (MWM) ............

100.0 76.6

12. r EACTOR SERVICE P ACTOR . . . . . . . . . . . . . . . . . . . . . .

... 100.0 100.0 ,

100.0 78.7

13. REACTOR AV AILAEILITY P ACTOR . . . . . . . . . . . . . . . . . . . . .

100.0 74.4

14. UNIT SERVICE P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 0. 0 100.0 100.0 74.9

15. UNIT AV AILAglLITY P ACTOR . . . . . . . . . . . . . . . . . . . . . . . .

94.4 62.7

18. UNIT CAPACITY P ACTOR (Using MOC) .................... 94.4 61.1 92.0 92.0

17. UNIT CAPACITY PACTOR (Umag Osmion MWel . . . . . . . . . . . . . . . .

0.0 0.O 8.4

18. UNIT PORCEO OUTAGE RATE . . . . . . . . . . . . . . . . . .

19. SMUT 00WNS SCHEDULE 0 CVER NEXT E MONTHS (TYPE, CATE. AND OURATION OF EACMI:

20. IP SMUT DOWN AT END OP REPORT PERICO. ESTfMATED DATE OF STARTUP:

PORECAST ACHIEVED

21. UNITS IN TEST STATUS (PRIOR TO COMMERCI AL OPERATIONI:

INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERAT1006 1.164 1

l. _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO, 50-254 l 1

UNIT- one i DATE February 6, 1989 COMPLETED BY Lvnne Deelsnyder TELEPHONE 309-654-2241 MONTH January, 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL' (MWe Net) (MWe-Net) 1 723 37 750 2 '744 q, 751 3 728 gg 737 4 742 748 3

5 731 21 749 [

8 749 648 -

22 714 772 7

8 740 24 773 g 756 739 3

10- . 792 3 763 743 p 752 39 12 749 28 760 13 729 760 29 14 308 30 752 15 637 33 765 16 767 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.

These figures will be used to plut a graph for cach reporting month. Note that when maximum dependable capacityis used f or the net electrical rating of the unit, there may be occasions when the daily average power level exceeds the 100'# line (or the restricted power level line). In such cases, the average daily unit power output sheet should be footnoted to explain the apparent anomaly.

l.16-8 4

[ . .. ..

t APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-265 UNIT Two DATE Feburary 6, 1989 COMPLETED BY Lynne Deelsnyder TELEPHONE 309-654-2241 1

s

+

MONTH January, 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe Net) (MWe-Net) 1 715 37 740 2 726 gg 735 3 686 3, 726 4 771 20 732 5

754 749 21 s

8 728 22 712 7 496 23 732 ,

8 770 24 732 g 724 25 742 ,

10- 731 26 757 693 737 33 27 12 730 28 746 706 739 13 29 14 728 y 741 15 735 31 767 16 727 INSTRUCTIONS On this form, list the average daily unit power level in MWe Net for each day in the reporting month. Compute to the neaiest whole megawatt.

These figures will be used to plot a graph for cach reporting month. Note that when maximum dependable capacityis used for lhe net electrical rating of the unit, there may be occasions when the daily average power level exceeds the 100'4 line (or the restrwted power level line). In such cases, the average daily unit power output sheet should be footnoted tu explain the apparent anomaly.

1.16-8

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VI. UNIQUE REPORTING REQUIREMENTS The following items are included in this report based en prior commitments to the commission:

A. Main Steam Relief Valve Operations There were no Main Steam Relief Valve Operations for the reporting period.

B. Control Rod Drive Scram Timing Data for Units One and Two There was no Control Rod Drive Scram Timing Data for Units One and Two for the reporting period.

0027H/00612 I

VII. REFUELING INFORMATION The'following information about future reloads at Quad-Cities Station was requested in a January 26, 1978,. licensing memorandum (78-24) from D..E.

O'Brien to C. Reed, et al., titled "Dresden, Quad-Cities, and Zion Station--NRC Request for Refueling Information",. dated January 18. 1978.

1 0027H/0061Z

t. _ . _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ _ - _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ - _ _ -

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I.- QTP 300-S32 Revision 1 QUAD-CITIES REFUELING Harch 1978 INFORMATION REQUEST I- *

1. Unit: 01 Reload: 9 Cycle: 10

2. Scheduled date for next refueling shutdown: 9-9-89

.3 Scheduled date for restart following refueling: 12-11-89

4. Will refueling or resumption of operation thereaf ter require a tec'nnical specification change or other license amendment:

NOT AS YET DETERMINED.

-5 Scheduled date(s) for submitting proposed IIcensing action and supporting information:

JUNE 10, 1989

6. Important licensing considerations associated with refueling, e.g., new or

' different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:

.NONE AT PRESENT TIME.

7 The number of fuel assemblies.

a. Number of assemblies in core: 724

b. Number of assemblies in spent fuel pool: 1773

8. The present IIcensed spent fuci pool storage capacity.and the size of any increase in licensed storage capacity that has been requested or is planned in number of fuel assemblies:

a. Licensed storage capactty for spent fuel: 3657

b. Planned increase in licensed storage: 0 9 The projected date of the last refueling that can be discharged to the spent fuel pool asseming the present Ilcensed capacity: 2008 XPPROVED 4

APR 2 01978 )

o. c. o. s. a. I a

l

______2-_____________ __

' ~

QTP 300-S32

, . Revision 1

,. QUAD-CITIES REFUELING March 1978-INFORMATION REQUEST

1. Unit: 02 Reload: 9 Cycle: 10

2. Scheduled date for next refueling shutdown: 2-3-90 3 Scheduled date for restart following refueling: 5-7-90'

4. Will' refueling or resumption of operation thereafter require a technical specification change or other IIcense amendment:

NOT AS YET DETERMINED.

5 Scheduled date(s) for submitting proposed licensing action and supporting information:

NOVEMBER 2, 1990

6. Important licensing considerations associated with refueling, e.g., new or

' different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:

NONE AT PRESENT TIME.

7 The number of fuel assemblies,

a. Number of assemblies in core: 724

b. Number of assemblies in spent fuel pool: 1475

8. The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is planned in number of fuel assemblies:

a. Licensed storage capacity for spent fuel: 3897 l b. Planned increase in licensed storage: 0 9 The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 200Rr WPPROVED AJ)1R 2 01978 C). C:. C). S. R.

- =-__=_ _ _ _ _ _ _ - _ _

VIII. GLOSSARY The following abbreviations which may have been used in the Monthly Report, are defined below:

ACAD/ CAM - . Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring ANSI -

American National Standards Institute APRM - Average Power Range Monitor ATHS - Anticipated Transient Without Scram BWR --

Boiling Water Reactor.

CRD -

Control Rod Drive EHC -

Electro-Hydraulic Control System E0F -

Emergency Operations Facility GSEP - . Generating Stations Emergency Plan HEPA -

High-Efficiency Particulate Filter HPCI -

High Pressure Coolant Injection System HRSS -

High Radiation Sampling System IPCLRT -

Integrated' Primary Containment Leak Rate Test IRM -

Intermediate Range Monitor ISI - Inservice Inspection LER - Licensee Event Report LLRT -

Local Leak Rate Test LPCI - Low Pressure Coolant Injection Mode of RHRS LPRM - Local Power Range Monitor MAPLHGR - Maximum Average Planar Linear Heat Generation Rate MCPR -

Minimum Critical Power Ratio MFLCPR - Maximum Fraction Limiting Critical Power Ratio MPC -

Maximum Permissible Concentration MSIV -

Main Steam Isolation Valve NIOSH - National Institute for Occupational Safety and Health PCI -

Primary Containment Isolation PCIOMR - Preconditioning Interim Operating Management Recommendations RBCCH - Reactor Building Closed Cooling Hater System RBM -

Rod Block Monitor RCIC - Reactor Core Isolation Cooling System RHRS . - Residual Heat Removal System RPS - Reactor Protection System RWM -

Rod Horth Minimizer SBGTS -

Standby Gas Treatment System SBLC -

Standby Liquid Control '

SDC - Shutdown Cooling Mode of RHRS SDV - Scram Discharge Volume SRM -

Source Range Monitor TBCCW - Turbine Building Closed Cooling Hater System  ;

TIP - Traversing Incore Probe TSC - Technical Support Center I

0027H/0061Z l

j ,. ..

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,!T3OCommonzalth Edison

) ould Citts Nucleir Power Station

'y. .d - ~ * {; t.,3; ', Co22710 do 206 Avenue North

.\ r va, Illinois 61242-9740

'*" v -

. Telephone 309/654 2241

. RAR-89-04

. February 06, 1989 Director of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Mall Station Pl-137 Washington, D. C. 20555 Enclosed for your information is the Monthly Performance Report covering the operation of Quad-Cities Nuclear Power Station, Units One and Two, during the month of January, 1989.

Respectfully, COMMONWEALTH EDIS0N COMPANY QUAD-CITIES NUCLEAR POWER STATION R. A. Robey f Services Superintendent RAR/vmk/eb Enclosure l^

0027H/0061Z

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