ML20245F323

Monthly Operating Repts for May 1989 for Quad-Cities Nuclear Power Station

ML20245F323
Person / Time
Site:	Quad Cities
Issue date:	05/31/1989
From:	Deelsnyder L, Robey R COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RAR-89-37, NUDOCS 8906280098
Download: ML20245F323 (28)

Text

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Commonwealth Edison

~ ouad Cities Nuclear Power Station 22710 206 Avenue North Cordova, Illinois 61242

Telephone 309/654-2241

.RAR-89-37 June 1, 1989 Director of Nuclear Reactor Regulations U. S. Nuclear. Regulatory Commission Hall 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 May, 1989.

^ Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION

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QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT MAY, 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 h :.

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l TABLE OF CONTENTS 1.

Introduction II.

Summary of Operating Experience A.

Unit One B.

Unit Two l

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.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 i-C.

Tests and Experiments Requiring NRC Approval i

D.

Corrective Maintenance of Safety Related Equipment IV.

Licensee Event Reports V.

Data Tabulations A.

Operating Data Report l

B.

Average Daily Unit Power Level C. -Unit Shutdowns and Power Reductions VI.

Unique Reparting Requirements A.

Main Steam Relief Valve Operations B.

Control Rod Drive Scram Timing Data VII.

Refueling Information VIII.

Glossary 0027H/00612

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INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Boiling Water Reactors, each with a Maximum Dependable Capacity of 769 MHe Net, located in Cordova, Illinois.

The Station is 1ointly owned by Commonwealth Edison Company and Iowa-Illinois Gas & Electric Company. The Nuclear Steam Supply Systems are General Electric Company Boiling Hater 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.

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Q II.

SUKHARY OF OPERATING EXPERIENCE A.

Unit One Unit One began the month of May holding load at 454 MWe.

At 0455 hours0.00527 days <br />0.126 hours <br />7.523148e-4 weeks <br />1.731275e-4 months <br />, a load increase to' full power was taken and achieved at 0955 hours0.0111 days <br />0.265 hours <br />0.00158 weeks <br />3.633775e-4 months <br />. On May 2 at 1052 hours0.0122 days <br />0.292 hours <br />0.00174 weeks <br />4.00286e-4 months <br />, the unit was placed in Economic Generation Control-(EGC) at the request of the Chicago Load Dispatcher. The unit remained in EGC until May 3.

At 0446 hours0.00516 days <br />0.124 hours <br />7.374339e-4 weeks <br />1.69703e-4 months <br />, the unit was taken out of EGC due to 'B' recirculation pump speed oscillations. At 0515 hours0.00596 days <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br />, a load increase to full power was taken per the request of the Load Dispatcher. B10 BWe was achieved at 0575 hours0.00666 days <br />0.16 hours <br />9.507275e-4 weeks <br />2.187875e-4 months <br />.

Power levels were held constant until May 4.

At 2150 hours0.0249 days <br />0.597 hours <br />0.00355 weeks <br />8.18075e-4 months <br />, a power reduction to 500 MWe was taken with recirculation pumps to repair a leaking valve.

On May 5, repairs were completed, and at 0505 hours0.00584 days <br />0.14 hours <br />8.349868e-4 weeks <br />1.921525e-4 months <br />, the Load Dispatcher requested an increase in power levels. At 0720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />, 820 MWe was achieved. The unit was placed in EGC.

The unit remained in EGC until May 6.

At 2225 hours0.0258 days <br />0.618 hours <br />0.00368 weeks <br />8.466125e-4 months <br />, EGC was tripped and a power reduccion was taken to perform maintenance on the Electro-Hydraulic Control (EHC) system.

On May 7 at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, the generator was takan off-line to replace electrical boards in the EHC system. At 1120 hours0.013 days <br />0.311 hours <br />0.00185 weeks <br />4.2616e-4 months <br />, the turbine was reset, and the EHC pumps were turned on.

At 1457 hours0.0169 days <br />0.405 hours <br />0.00241 weeks <br />5.543885e-4 months <br />, the mode switch was.placed in RUN and at 1645 hours0.019 days <br />0.457 hours <br />0.00272 weeks <br />6.259225e-4 months <br />, the generator was synchronized to the grid. An increase in power was begun with control rods and recirculation pumps. At 1950 hours0.0226 days <br />0.542 hours <br />0.00322 weeks <br />7.41975e-4 months <br />, 450 MWe was achieved and power levels were held constant at the request of the Load Dispatcher until May 8.

At 0105 hours0.00122 days <br />0.0292 hours <br />1.736111e-4 weeks <br />3.99525e-5 months <br />, a load increase to full pwoer was taken and 815 MWe was achieved at 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />.

At 0955 hours0.0111 days <br />0.265 hours <br />0.00158 weeks <br />3.633775e-4 months <br />, a " Turbine Bypass Valve Open" alarm was received in the control room and the turbine control vavles were found to be spiking 70% - 100% open. A power reduction was taken to 450 MWe and held constant while this problem was resolved. On May 9 at 0410 hours0.00475 days <br />0.114 hours <br />6.779101e-4 weeks <br />1.56005e-4 months <br />, power levels were increased to 750 MWe.

On May 10 at 0010 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, the unit was placed in EGC.

The unit remained in EGC until May 16 with minor interruptions to perform routine surveillance.

At 2035 hours0.0236 days <br />0.565 hours <br />0.00336 weeks <br />7.743175e-4 months <br />, the control room received a report that the leak on the low flow feedwater regulating valve drain line had significantly increased.

This leak is unisolatable, thus a unit shutdown was begun as rapidly as systems would allow. The main generator was taken off-line at 2213 hours0.0256 days <br />0.615 hours <br />0.00366 weeks <br />8.420465e-4 months <br />, and a manual scram was inserted at 2215 hours0.0256 days <br />0.615 hours <br />0.00366 weeks <br />8.428075e-4 months <br />.

From the remainder of May 16 thru May 19, the feedwater system was isolated, and repairs were made on the low flow feedwater drain line.

On May 19 at 1257 hours0.0145 days <br />0.349 hours <br />0.00208 weeks <br />4.782885e-4 months <br />, maintenance was completed on the drain line, and rod maneuvers were begun. At 1526 hours0.0177 days <br />0.424 hours <br />0.00252 weeks <br />5.80643e-4 months <br />, the reactor was made i

critical.

On May 20 at 0910 hours0.0105 days <br />0.253 hours <br />0.0015 weeks <br />3.46255e-4 months <br />, the mode switch was placed to RUN and at 1330 hoars, the generator was synchronized to the grid.

A load increase was begun, using control rods and recirculation pumps.

450 MWe was achieved at 2022 hours0.0234 days <br />0.562 hours <br />0.00334 weeks <br />7.69371e-4 months <br /> and held constant until May 22.

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u On May122 at 0715 hours0.00828 days <br />0.199 hours <br />0.00118 weeks <br />2.720575e-4 months <br />, a load increase to full power was begun at the request of'the Chicago Load Dispatcher.

At 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />, 812 MWe was achieved.

Full load was. held until May 24 when power levels were adjusted, and the unit was placed in EGC.

The unit remained in EGC until May 25.

At 0733 hours0.00848 days <br />0.204 hours <br />0.00121 weeks <br />2.789065e-4 months <br /> EGC was tripped and an ascent to full power was taken per the Load Dispatcher.. At 1722 hours0.0199 days <br />0.478 hours <br />0.00285 weeks <br />6.55221e-4 months <br />, power levels were adjusted, and the unit was placed in EGC.

On May 26 at 0350 hours0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br />, the unit was taken off EGC and a power reduction to 580 MWe was taken. At 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, an ascent back to full power was begun per the Load Dispatcher. At 0602 hours0.00697 days <br />0.167 hours <br />9.953704e-4 weeks <br />2.29061e-4 months <br />, the unit was placed in EGC.

The unit remained in EGC until May 28.

At 0140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br />, the unit was taken off EGC and a load reduction to 568 MWe was taken. At 0440 hours0.00509 days <br />0.122 hours <br />7.275132e-4 weeks <br />1.6742e-4 months <br />, a further reduction in power was taken to 460 MWe.

Power levels were held constant until 0830 hours0.00961 days <br />0.231 hours <br />0.00137 weeks <br />3.15815e-4 months <br />. A load increase to 650 MWe was taken at the request of the Load Dispatcher.

On May 29 at 1140 hours0.0132 days <br />0.317 hours <br />0.00188 weeks <br />4.3377e-4 months <br />, an ascent to full load was taken. On May 30 at 0220 hours0.00255 days <br />0.0611 hours <br />3.637566e-4 weeks <br />8.371e-5 months <br />, a power reduction to 595 MWe was taken. At 2122 hours0.0246 days <br />0.589 hours <br />0.00351 weeks <br />8.07421e-4 months <br />, the unit was placed in EGC.

On May 31 at 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br />, control rods were inserted to reduce power to 450 MWe due to control valves spiking'and a " Turbine Bypass Valve Open" alarm recieved in the control room. Unit was held at 450 MWe for the remainder of the month.

B.

Unit Two Unit'Two' began the month of May operating in Economic Generation Control (EGC). Normal operational activities occurred for Unit Two until May 22.

The unit operated near full power or remained in EGC with minor interruptions to perform routine surveillance.

Power levels were adjusted accordingly.

On May 22 at 1753 hours0.0203 days <br />0.487 hours <br />0.0029 weeks <br />6.670165e-4 months <br />, a power reduction to 450 MWe was taken at the request of the Chicago Load Dispatcher. At 2120 hours0.0245 days <br />0.589 hours <br />0.00351 weeks <br />8.0666e-4 months <br />, a further reduction in power was taken to 200 MWe to de-inert the drywell in preparation for entry to repair leakage discovered.

At 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, repairs were completed and the drywell was inerted.

At 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />, an ascent to full power was begun.

818 MWe was achieved at 1241 hours0.0144 days <br />0.345 hours <br />0.00205 weeks <br />4.722005e-4 months <br />. At 2315 hours0.0268 days <br />0.643 hours <br />0.00383 weeks <br />8.808575e-4 months <br />, a power reduction to 450 MWe was taken per request of the Load Dispatcher.

On May 24 at 0415 hours0.0048 days <br />0.115 hours <br />6.861772e-4 weeks <br />1.579075e-4 months <br />, Unit Two drywell floor drain leakage rate increased above the Tech Spec limit of 5 gallons per minute. The Shift Engineer was notified of this abnormality. A GSEP unusual event was declared, and at 0640 hours0.00741 days <br />0.178 hours <br />0.00106 weeks <br />2.4352e-4 months <br /> a unit shutdown was initiated.

Investigation was begun into the cause of the excess leakage and was identified as coming from the Reactor Core Isolation Cooling (RCIC) Turbine Steam Supply. Inboard Isolation Valve. At 1540 hours0.0178 days <br />0.428 hours <br />0.00255 weeks <br />5.8597e-4 months <br />, the generator was taken off-line, and at 1559 hours0.018 days <br />0.433 hours <br />0.00258 weeks <br />5.931995e-4 months <br />, a manual scram was inserted. The RCIC isolation valve was repacked, and the valve was tested for proven operability. At 2017 hours0.0233 days <br />0.56 hours <br />0.00333 weeks <br />7.674685e-4 months <br />, the GSEP unusual event was terminated.

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On May 25 at.0537 hours0.00622 days <br />0.149 hours <br />8.878968e-4 weeks <br />2.043285e-4 months <br />, the mode switch was placed in STARTUP and at 0705 hours0.00816 days <br />0.196 hours <br />0.00117 weeks <br />2.682525e-4 months <br />, control rod maneuvers were begun. At 0945 hours0.0109 days <br />0.263 hours <br />0.00156 weeks <br />3.595725e-4 months <br />, the reactor was made critical. At 1505 hours0.0174 days <br />0.418 hours <br />0.00249 weeks <br />5.726525e-4 months <br />, the generator was synchronized to the grid. A gradual increase to full power was begun and on May 26 at 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br />, 808 MWe was achieved. Power levels were held constant until-May 28.

At 0039 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br />, power levels were adjusted, and the unit was placed in EGC. The unit remained in EGC until May 29.

At 2055 hours0.0238 days <br />0.571 hours <br />0.0034 weeks <br />7.819275e-4 months <br />, EGC was tripped due to the loss of 2B Recirculation Pump Speed indication.

On May 30 at 0012 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, it was discovered that neither of the Drywell Floor Drain Sump Pumps were operating properly.

Per Generic Letter 84-11 if the drywell floor drain sump monitoring system is inoperable and cannot be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the unit must initiate an orderly shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and be in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. At 1821 hours0.0211 days <br />0.506 hours <br />0.00301 weeks <br />6.928905e-4 months <br />, the generator was taken off-line and at 1829 hours0.0212 days <br />0.508 hours <br />0.00302 weeks <br />6.959345e-4 months <br />, a manual scram was inserted.

Unit Two remained shutdown thru the end of the month while maintenance was performed on the drywell floor drain sump pumps.

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.III.

PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY i

_R_ ELATED MAINTENANCE A.

Amendments to Facility License or Technical Specifications There were no Amendments to the Facility License or Technical Specifications for the 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.

Test and Experiments Requiring NRC Approval There were no Tests or Experiments requiring NRC approval for toe 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 l

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.:

Q66683 LER NUMBER'.

NA COMPONENT: System 6600 - The 1/2 DG would not close in to Bus 13-1.

CAUSE OF MALFUNCTION:. While performing monthly procedure QoS 6600-1 in preparation for the Diesel Generator maintenance, the 1/2 Diesel Generator would not close to Bue 13-1.

Upon investigation, the cause for the 1/2 Diesel Generator's initial failure to close in to Bus 13-1 could not be determined because operating personnel racked out the breaker before Electrical Maintenance could inspect it in the failed state.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plant and personnel were unaffected because the breaker could be closed successfully on subsequent attempts.

ACTION TAKEN TO PREVENT REPETITION: After the 1/2 Diesel Generator was declared inoperable, Electrical Maintenance inspected both the 13-1 and 23-1 breakers.

In addition, strip charts were placed in the 1/2 Diesel Generator logic to Buses 13-1 and 23-1.

The breaker was replaced under Work Request Q66682. The old breaker was rebuilt under Work Request Q66683.

WORK REQUEST No.: Q68258 LER NUMBER:

87-008 COMPONENT: System 1000 - Inspect and verify that the anchor bolts for all CEA plates with RHRSW support attachments are properly torqued. The bolts were inspected and checked by torquing the bolts to 45 ft. Ibs.

CAUSE OF MALFUNCTION: While performing a visual inspection of the "B" and "C" RHRSW pump vault, it was discovered that two of the four bolts on a Concrete Expansion Anchor (CEA) plate were sheared off.

Pipe hangar M-994D-107, which provides support for the "C" RHRSW Pump Discharge Line (1-1004B-12"), is attached to that plate. Work Request Q55793 was initiated to repair the problem. The operating engineer determined that the damage to the hangar did not make the 1C RHRSW pump inoperable. The cause of the anchor bolt failure was determined to be due to anchor bolt nuts loosening in service while being subjected to operational vibration near the location.

RESULTS 6 EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal because at no time was the RHRSW eystem, or any component thereof, inoperable.

ACTION TAKEN TO PREVENT REPETITION: The initial corrective action was to replace the existing bolts and CEA plate with heavy duty bolts and plate.

I Work Request Q55793 was written to do the work. The supports for other RHRSW pump discharge lines were visually examined by a Techanical Staff Eagineer et the time of the problem. Work Request Q68258 was written to inspect all CEA on Unit One.

A followup by Impell concluded that the bolt failure was not a result of poor bolt material.

Pump vibration caused the loosening of the nuts and resulted in the fatigue failure of the bolts.

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l WORK REQUEST No.: Q71741, 71742, 71743, 71744 LER'NO: NA COMPONENT:

System 5600 - Inspected scram limit switch on fast-acting solenoid for loose mounting screws on Turbine Control Valve (TCV) #1, #2, #3, and #4.

l EPN 1-5601-CV1, CV2, CV3, and CV4 l

CAUSE OF MALFUNCTION: While control room personnel were performing QOS 5600-1, Turbine Control Valve Fast Closure Scram Instrumentation Functional Test, the #4 TCV failed to give the expected Reactor Protection System (RPS)

"B" channel half-scram on two of four actuations of the fast-acting solenoid.

The cause of the failure of TGV #4 to give a half-scram on RPS channel "B" when the fast-acting solenoid was energized was determined to be a loose limit switch on the fast-acting solenoid.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consegrences of the event were considered to be minimal.

Immediately prior to the event, TCV #1, TCV #2, and TCV #3 were all successfully tested.

If a generator load rejection had occurred while the TCV #4 fast-acting solenoid was inoperable, the fast-acting solenoid on TGV #3 would have provided a scram signal to RPS channel "B".

ACTION TAKEN TO PREVENT REPETITION: Work Request Q70721 was written to reinstall the screw and tighten the limit switch in place for TCV #4.

The limit switches on the other Unit Iwo TCV's were inspected under Work Requests Q71745, Q71746, and Q71747. Unit 1 TCV's were also inspected under Work Request Q71741, Q)1742, Q71743, and Q71744.

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UNIT 2 MAINTENANCE

SUMMARY

E WORK REQUEST NO.: Q66106 LER NUMBER: 27-019-COMPONENT:

System 1000 - Piping support on Line 1-1024B-20" found to be g'

outside FSAR criteria.

CAUSE-OF MALFUNCTION: Two piping supports were found to be out of FSAR compliance. The affected line was 2-1024B-20" (Residual Heat Removal

.2C/2D Suction). The apparent cause of the event was design error involving A/E and contractor personnel.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plant and personnel were not affected during this event because the piping met operability requirements even though it didn't meet FSAR compliance.

ACTION TAKEN TO PREVENT REPETITION: The corrective action was to replace a rigid strut on line 1-1024B-20".

Work Request Q66106 was eritten to replace the strut.

WORK REQUEST NO.:

Q70920, Q70921 LER NUMBER: NA COMPONENT:

System 1000 - Two spring can piping supports were found to be out of adjustment. The line affected was line 2-1012A-16".

CAUSE OF MALFUNCTION:

Piping supports 1012A-M-204 and 1012B-M-207 were found to be out of adjustment. The cause of the event was inadequate work instructions for Work Requests Q69427 and Q69851. The piping supports had been improperly adjusted to the " Hot Load" setting instead of to the

" Cold Load" setting. The piping for the RHR system was not in use; therefore the system would be considered " Cold".

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal because engineering analysis showed that the spring type piping supports were operable.

ACTION TAKEN TO PREVENT REPETITION:

The corrective action was to recet the spring cans. The work was done under Work Requests Q70920 and Q70921.

To prevent future problems, training was to be given to Mechanical Maintenance 1

Analysts and ISI group for the purpose of defining proper pipe support adjustment procedures.

U.

W WORK REQUEST No.: Q71745, Q71746, Q71747 LER NUMBER:

NA.

COMPONENT:

System 5600 - Inspected scram limit switch on fast-acting solenoid for loose mounting screws on Turbine Control Valve (TCV) #1, #2, and #3.

EPN 2-5601-CV1, CV2, and CV3 CAUSE OF MALFUNCTION: While control room personnel were performing QOS 5600-1, Turbine Control Valve Fast Closure Scram Instrumentation Functional Test, the #4 TCV failed to give the expected Reactor Protection System (RPS)

"B" channel half-scram on two of four actuations of the fast-acting solenoid.

The cause of the failure of TGV #4 to give a half-scram on RPS channel "B" when the fast-acting solenoid was energized was determined to be a loose 1*mit switch on the fast-acting solenoid.

RESULTS 6' EFFECTS ON SAFE OPERATION: The safety consequences of the event were considered to be minimal.

Immediately prior to the event, TCV #1,-

TCV #2, and TCV #3 were all successfully tested.

If a generator load rejection had occurred while the TCV #4 fast-acting solenoid was inoperable, the fast-acting solenoid on TCV #3 would have provided a scram signal to RPS channel "B".

ACTION TAKEN TO PREVENT REPETITION:

Work Request Q70721 was written to reinstall the screw and tighten the limit switch in place for TCV #4.

The limit switches on the other Unit Two TCV's were inspected under Work Requests Q71745, Q71746, and Q71747. Unit 1 TCV's were also inspected under Work Request Q71741, Q71742, Q71743, and Q71744.

WORK REQUEST NO.: Q72095 LER NUMBER: NA COMPONENT:

System 1000 - 2A RHRSW pump didn't meet required flow.

CAUSE OF MALFUNCTION: While performing QOS 1000-S4, "FERSW Pump Flowrate Testing Data Sheet", the 2A RHRSW pump did not meet the required flowrate.

It was then declared inoperable. The cause of the event was material found in the main pump and the wear rings out of tolerance on the booster pump.

RESULTS & EFFECTS ON SAFE OPERATION: The safety significance of the event was minimal. All other components of the containment cooling system were operable as rcquired by Tech Specs.

ACTION TAKEN TO PREVENT REPETITION: Work Request Q72095 was written to investigate the low flowrate problem. Mechanical Maintenance brought the wear rings back into tolerance by knurling the cotor wear ring surface.

4 WORK REQUEST No.: Q72892 LER NUMBER: ;NA COMPONENT:

System 260 - Thermocouple 261-13G on Safety Valve 203-4G was reading erratically.

CAUSE OF MALFUNCTION: A failed thermocouple or loose electrical connection was suspected as the cause of the problem.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal. The temperature indication is a passive device and in no way affects the operation of the safety valve.

The acoustic mon 1 tor for this valve was functional; therefore, adequate valve position indication was available to satisiy technical specifications.

ACTION TAKEN TO PREVENT REPETITION: The thermocouple was replaced.

It was verified that a proper reading was received. No futher corrective action was required.

WORK REQUEST NO.

Q73543 LER NUMBER:

NA COMPONENT:

System 1700 - Fuel Pool Radiation Monitor tripped due to loose power supply connector.

The Radiation Monitor affected was 2-1705-16A.

CAUSE OF MALFUNCTION: While performing ST-35, Fuel Pool Radiation Monitor Functional, it was thought that a loose connector caused both the 2A Fuel l

Pool and 2A Reactor Building Ventilation Radiation Monitors to trip downscale.

The actual cause of the inoperable 2A Fuel Pool and 24 Reactor Building Ventilation Radiation Monitors was the failure of the high voltage power supply, which is the common supply to both of these radiation monitors.

RESULTS & EFFECTS ON SAFE OPERATION: The consequences were minimal since the Reactor Building Vents were already isolated.

in addition, the Standby 1

Gas Treatment System was already running when the power supply failed.

All trips and alarms functioned as designed during this event.

ACTION TAKEN TO PREVENT REPETITION: The two monitors were bypassed to allow work on the high voltage power supply.

Under Work Request Q74543, it was verified that there were no problems with the multipin connector and a new high voltage supply was installed. Work Request Q73582 was written to repair the faulty high voltage supply.

I l

l

1 1

WORK REQUEST No.: Q73576, Q73577, Q73579, Q73580, Q73581 LER NUMBER: NA l

COMPONENT:

System 2300 - Replaced existing splice for connecting the pigtails of the Rosemount conduit seal to its associated power leads with a Environmentally

-Qualified (EQ) splice.

EPN 2-2352, 2-2353, 2-2389B, 2-2389C, and 2-2389D CAUSE OF MALFUNCTION:

It was determined that the signal leads for the Rosemount conduit seal pigtails on Rosemount transmitters 2-2352, 2-2353, and 2-2389A, B, C, and D had questionable EQ splicing configurations. A managmeent deficiency error was determined to be the reason for the improper installations. This was caused by a failure of BWRED in its review process to recognize that the AE's design could not be implemented. An additional cause was that, on discovery of the inability to properly install the conduit seal splicers, the Substation Construction Department failed to inform BWRED of the discrepancy.

RESULTS & EFFECTS ON SAFE OPERATION: The safety significance was minimal because all transmitters were declared operable although they were not EQ qualified.

ACTION TAKEN TO PREVENT REPETITION:

Splices for 2-2389B, C, and D were replaced under Work Request Q73579, Q73580, and Q73581.

Splices 2-2352 and 2-2353 were replaced under Work Requests Q73576 and Q73577.

In order to be assured that all.EQ criteria have been met in past installations of Raychem sp.lices, Quad-Cities Station will perform a 100 percent walkdown of all identified EQ cable splices in harsh environments.

This will be done by the end of the next refuel outage.

l

r__________-__-

4 IV.

LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for 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.I. and 6.6.B.2. of the Technical Specifications.

UNIT 1 Licensee Event Report Number DATE Title of Occurrence 89-005 5-22-89 RCIC Inoperable UNIT 2 89-002 5-24-89 DW Floor Drain Leakage

> 5 gpm 89-003 5-29-89 Loss of Secondary Containment - Interlock Doors Open 0027H/0061Z

r l:

l IV.

LICENSEE EVENT REPORTS The-following is a tabular summary of all licensee event reports for-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.0ccurrence 89-003 4-12-89 Manual Scram.

EHC problems89-004 4-17-89 Manual Scram, Stuck open on Relief valve 89-005

• cancelled 4-17-89 Control Room Emergency Air Filtration Unit inoperable UNIT 2

'89-001 4-06-89 Turbine Trip - Reactor Scram while testing Turbine Master trip solenoid

• This has been cancelled since the report in April due to the Control Room Emergency Air Filtration Unit was not inoperable.

l I

0027H/00612

I V.

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

A.

Operating Data Report 8.

Average Daily Unit Power Level C.

Unit Shutdowns and Power Reductions l

l 1

l l

0027H/0061Z

.3 APPENOlX C OPERATING DATA REPORT

' 00CKET NO.

50-254 UNIT One DATE June 9. 1989 COMPLETED gy Lynne Deelsnyder l

TELEPHONg 309-654-2241 OMHATING STATUS 0000_

050185 053189

s. REPORTING PERICO. 2400 ORCES MOURS IN REPORTINO PERIOG:

2511 MAX. 0EPENO. CAPACITY tasWo.Ned 769

' 3. CURRENTLY AUTMORISEO POWER LEVEL tesutt

/UV DESIGN ELSCTRICAL RATING tesW>Ned:.

N/A

3. POWER LEVEL TO WMN'.N RWTRICTED (IP ANY) lehNed:

4. REAS0ss POR RESTRICTION (IP ANYh THIS MONTH YRTO OATE CutsWLAftVE 678.6 3432.8 120975 S. Nuts 98R OF MOURS REACTOR WA8 CRITICAL...............

O.0 0.0 3421.9

8. REACTOR RESERVE SMUT 00Wes MOURE...................

645.0 3363.7 117022.9

7. MOURE GENERATOR ON UNS.......................

0.0 0.0 909.2 E. UNIT RESERVE SMUT 00Wd '40URS.....................

1414370 7768701 249458780

s. ORCES THERMAL ENERGY GENERATED 14NUMI 455303 2528938 8088655 to. ORCEE ELECTRICAL ENERGY GENERATED tesWHI.............

433876 2419860

75985134,

11. NET ELSCTRICAL ENERGY GENERATED (MWHI

.................... _ 91.2

, 94.8 80.9

12. e TACTOR SERVICE P ACr0R......

91.2 94.8 83.2

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

86.7 92.8 78.3

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

86.7 92.8 78.9

15. UNIT AV AILAtluTY P ACTOR.........................

75.8 86.9 66.1

18. UNIT CAPACITY P ACTOR (Using MOCl.....................

73.9 84.7 64.4

17. uMIT CAPACITY PACTOR tumas Osman MWel.................

13.3 7.2 5.4_

18. UNIT PORCEO QUTAGE RATE...................

l

15. SMUT 00gWs3 SCMEOUL50 OVER NEXT E MONTHS ITYPE. DATE. AND DURATION OF EACMl:

20. IP SMUT DOWN AT END CP REPORT PERIOD. ESTIMATED DATE OF STARTUP:

21. UNITS IN TEST ITATUS (PRIOR TO COMMEMCI AL OPERATIONI:

PORECAST ACHIEVED INITI AL CRITICALITY INITIAL ELECTRICITY COMestRCIAL OPERAT100s I.IM 1

1 i

b I

~

a APPENDIX C OPERATlivG DATA REPO(17 DOCKET NO.

50-265 UNIT _ Iwn DATE June 9, 1989 COMPLETED BY Lynne Deelsnyder l

TELEPHONg 309-654-2241 OPERATING STATUS 0000 050189 2400 053189 740 ORoss MOURs IN R$ PORTING PSR100*

1. REPORTING PERICO:

2511 max.otreNo. CAPACITY temerated. 769

2. CURPSNTLY AUTHOR 12EO POWER LEVEL gyoY 08880N GLSCTRICAL Rd. TING iMWo.med:

N/A

. 3. POWER LEV 9L TO WHICM RGBTRICTED (17 ANY1 (ARM >Ned:

4. REASONS POR RESTRICTION (17 ANYl:

TMitPONTN YR"9 OATE CutsutATIVE 696.7 3W 7.4 114457.3

5. NUMSER OF MOURS REACTOR IEJ CRITICAL...............

O. 0,

D.0 2985.8

8. REACTOR RESERVE SveUTOCWN MOURS...................

690.9 3482.7 111214.4

7. MOURS OSNERATOR 000 LINE.......................

0.0 0.0 702.9

8. UN47 RestRVE SMUTOOWN MOURS......................

S. ORoss THERMAL ENtROY GENERAT80 (MWHI............., 1505911 7929015 23883G?RR 486806 2589827 76523298

10. GROSB ELSCTRICAL ENERGY GENER ATED (WWMI....

,5183 2480999 72217576

11. NET WLECTRICAL ENERGY CENGMATED (MWHI 93.6 96.8 77.0

12. ' E ACTOR SE RVICE P ACTOR...........................

93.6 96.8 79.0

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

92.9 96.1 74.8

14. UNIT SSRvlCE P ACTOR..............................

92.9 96.1 75.3

19. UNIT AV AILASILITY P ACTOR........................

81.3 89.0 63.2

18. UNIT CAPACITY P ACTOR tueWig MOCl.....................

79.2 86.8 61.6

~f

17. UNIT CAPACITY PACTOR tumac 0ssipi MWel.................

7.1 3.9 8.3

18. UNIT PORC80 0UTAGE RATE..........................

l l

19. SMUT 00Wh8 SCHEDULEO OVER NExT s MONTHS ITYPE. DATE. AND OURATION OF EACM):

20. IP BMUT DOWN AT ENO 07 REPORT PERIOD. ESTIMATED DATE OP STARTUP:

11. UNITS 1:4 TEST STATUS (PP.IOR TO COMMk ACI AL OPERATIONI:

PORECAST ACHIEVfD INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERAT1000 1

$.IN

S.

f..

7 APPENDIX B

-I AVERAGE DAILY UNIT POWER LEVEL l

1 i

i DOCKET NO.

50-254

)

UNIT one DATE June 5, 1989 COMPLETED BY tvnna Deelsnyder'

l TELEPHONE 309-654-2241-l l

MONTH May, 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL i

(MWe Net)

(Mwe. Net) 689

-12 1

37 2

755 13

-13 746

-13 3

gg 744 132 4

y 743-403 p

5 21 i

731 672 a

22 149 773 7

23 642 765 8

660 744 g

655 686 10 y

701 686 11 27 756 564 12 3

13 691 29 625_

583 706 14 g

15 731 742 33 13 624 INSTRUCTIONS On this form. list the average daily unit power level in MWe. Net for each day in the reporting month. Compute to the nemiest whole megawatt.

These l'q' ires will be used to plot a graph for cach reporting month. Note that when rnaximum dependable capacityis uwd for the net electrical rating of the unit, there may be occasions when the daily average power level exceeds the 100'8 line (or the restricted power icvel line). In such cases, the average daily unit power output sheet should be footnoted to explans the apparent anomaly.

1.16 8 44

• * ^

9 e

c.

s w-

=e

~

z 33

, =,j

....,a

' APPENDlX B l,

AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-265 I

UNIT Two

'?

-DATE-. June 5, 1989.'

COMPLETED BY _Lynne Deelsnyder--

TELEPHONE 309-654-2241 MONTH _

May, 1"989 o

L DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe Net)

.1,

-720 17 710 727 708 2.

g 3-13 680' 713 4L 20 5'

713~

674

-[

,1 g:

721 707 22 7-683 461 23 8

704 24 198 g

769' 3

65 751 515 10.

.g 711 715 11 27

~ 12. -

734 715 g

., 13 677 688 3

--14 ;

703 3

412

' l 15-690 31

-11 16 71I

}

INSTRUCTIONS On this form, list the average daily unit power lent in MWe-Net for each day in the reporting month. Compute to the neassst whole megawatt.

These figures will be used to plot a graph for each reporting month. Note that when maximum dependable capacityis g.

used for the net electrical rating of the unit, there may be occasions when the daily average power level exceeds the 100'A line (or the restneted power icwl line), in such cases, the average daily unit power output sheet should be footnoted to explaua the apparent anomaly.'

l 1.16 8 j

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{

.{

i VI. UNIQUE REPORTING REQUIREMENTS The following items are included in this report based on 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.

1 1

0027H/00612 b

m__-_---_-_.

g ;, e '

ie l-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.

0027H/00612 l

~

QTP 300-S32 Revision 1 QUAD-CITIES REFUELING March 1978 INFORMATION REQUEST

~

1.. Unit:

01 Reload:

9 Cycle:

10 j

i 2.

Scheduled date for next refueling shutdown:

9-9-89 i

l 3.

Scheduled date for restart following refueling:

12-11-89 q

4.

Will refueling or resumption of operation theretf ter require a technical specification change or other IIconse 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.

i a.

Number of assemblies in core:

724 b.

Number of assemblies in spent fuel pool:

1773

)

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 assembliest a.

Licensed storage capacity for spent fuel:

us7 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 IIconsed capacity: 2008 4 P P R O. V E D APR 2 01978 Q.c.o.S.R.

i

• b. '

~

QTP 300-532 Revision 1 QUAD-CITIES REFUELING ~

March 1978 INFORMATION REQUEST i

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 license ammidment:

NOT AS'YET DETERMINED.

5.

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

1 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 cora:

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 assemblie :

a.

Licensed storage capacity for spent fuel:

3897 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 IIconsed capacity: 7008 WPPROVED APR 2 01978 Q. C. O. S. R.

O..

's f

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

i ACAD/ CAM -

Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring ANSI American National Standards Institute APRM Average Poeer Range Monitor ATHS Anticipated Transient Without Scram BHR Boiling Water Reactor

-CRD Control Rod Drive Electro-Hydraulic Control System EHC EOF 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 RBCCW Reactor Building Closed Cooling Water System RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling System RHRS Residual Heat Removal System RPS Reactor Protection System RWM Rod North Minimizer i

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 Water System TIP Traversing Incore Probe TSC Technical Support Center i

0027H/006:2