ML20246L848

Monthly Operating Repts for June 1989 for Quad-Cities Units 1 & 2.Refueling Info Also Encl

ML20246L848
Person / Time
Site:	Quad Cities
Issue date:	06/30/1989
From:	Deelsnyder L, Robey R COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
RAR-89-44, NUDOCS 8907190012
Download: ML20246L848 (26)

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Commonwealth Edison N,.

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-July 3, 1989 Director.of Nuclear' Reactor Regulations

U.'S. Nuclear Regulatory Commission Mail ~ Station Pl-137-

-Washington, D. C.

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. Enclosed for your information is the Monthly Performance Repor't covering the operation of Quad-Cities Nuclear Power Station, Units One?and Two, during the month of June. 1989.

Respectfully.

COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION hh.

R. A. Robey.

Technical Superintendent RAR/LFD/vmk Enclosure ~

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I QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT JUNE, 1989 COMMONWEALTH EDISON COMPANY AND

-IONA-ILLIN0IS GAS & ELECTRIC COMPANY NRC DOCKET N05. 50-254 AND 50-265 LICENSE NOS. DPR-29 AND DPR-30 i

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TABLE OF CONTENTS LL ~

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" Introduction p

II;

-Summary.of: Operating Experience

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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-I V.'

. 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 L..

B.

. Control' Rod Drive Scram Timing Data L

VII.

Refueling Information i

sVIII.

Glossary i

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INTRODUCTION y

R 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 s

Company and Iowa-Illinois' Gas & Electric Company.

The Nuclear Steam Supply Systems are General Electric Company Boiling 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 21.85 and 2240.

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.II.

SUMMARY

OF OPERATING EXPERIENCE A.

Unit One

-Unit One began the month of June operating at 635 MWe.

At 0110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br />, a power reduction to 450 MWe was taken. At 540 hours0.00625 days <br />0.15 hours <br />8.928571e-4 weeks <br />2.0547e-4 months <br />, a load increase was taken to full power. At 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, a turbine bypass valve open alarm was received in the control room and as a result, a load reduction was taken to 650 MWe.

On June 2, at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, another power reduction to 430 MWe was taken due to continuous control valve spiking problems.

At 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />, the Chicago Load Dispatcher requested an increase in power levels. At 1030 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91915e-4 months <br />,'775 MWe was achieved. Power levels were held constant until June 4.

At 0115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br />, a load reduction to 450 MWe was taken. At 1021 hours0.0118 days <br />0.284 hours <br />0.00169 weeks <br />3.884905e-4 months <br />, the unit was taken to 610 MWe per the Load

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Dispatcher. At 2340 hours0.0271 days <br />0.65 hours <br />0.00387 weeks <br />8.9037e-4 months <br />, a power reduction to 450 MWe was taken.

At 0241 hours0.00279 days <br />0.0669 hours <br />3.984788e-4 weeks <br />9.17005e-5 months <br />, on June 5, another power reduction to 370 MWe was taken due to alarms received in the control room indicating 1A2 Motor Generator (MG) set oil pump off and emergency on.

The 1A MG set was tripped until 0307 hours0.00355 days <br />0.0853 hours <br />5.076058e-4 weeks <br />1.168135e-4 months <br /> when the 1A MG Set was reset and a load increase was taken with control rods and recirculation pumps. At 0945 hours0.0109 days <br />0.263 hours <br />0.00156 weeks <br />3.595725e-4 months <br />, full power levels were achieved.

From June 6 through June 20, normal operational activities and sur-veillances were performed.

Power levels were adjusted according to demands as requested per the Chicago Load Dispatcher.

On June 20, at 0710 hours0.00822 days <br />0.197 hours <br />0.00117 weeks <br />2.70155e-4 months <br />, Economic Generator Control (EGC) limits were received. Power levels were adjusted and at 1003 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.816415e-4 months <br />, the unit was placed in EGC.

The unit remained in EGC until June 21.

At 0310 hours0.00359 days <br />0.0861 hours <br />5.125661e-4 weeks <br />1.17955e-4 months <br />, EGC was tripped and a power reduction to 450 MWe was taken to perform weekly and monthly turbine surveillance. At 0630 hours0.00729 days <br />0.175 hours <br />0.00104 weeks <br />2.39715e-4 months <br />, testing was completed and an ascent to full power was taken. At 0905 hours0.0105 days <br />0.251 hours <br />0.0015 weeks <br />3.443525e-4 months <br />, 790 MWe was achieved. On June 22, at 1645 hours0.019 days <br />0.457 hours <br />0.00272 weeks <br />6.259225e-4 months <br />, power levels were adjusted and the unit was placed in EGC.

At 2350 hours0.0272 days <br />0.653 hours <br />0.00389 weeks <br />8.94175e-4 months <br />, EGC was tripped and a power increase was taken to 790 MWe.

Power levels were held constant for various operational activities.

On June 23, at 2221 hours0.0257 days <br />0.617 hours <br />0.00367 weeks <br />8.450905e-4 months <br />, all testing was completed and the unit was placed in EGC.

On June 23, at 2305 hours0.0267 days <br />0.64 hours <br />0.00381 weeks <br />8.770525e-4 months <br />, the control room received a Channel B 1/2 Scram alarm.

Continuous problems occurred when attempting to reset this alarm. Upon investigation, it was determined that this reset problem was due to the failure of relay 590-100B for the Scram Discharge Volume (SDV) Hi-Hi level alarm. At 0047 hours5.439815e-4 days <br />0.0131 hours <br />7.771164e-5 weeks <br />1.78835e-5 months <br />, on June 24, the relay fuse was pulled after the relay began to chatter rapidly.

The SDV level switch card was replaced and Channel B 1/2 Scram was successfully reset.

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4 Until June 29, the unit operated in EGC or operated near full power while

. normal-operational activities occurred and routine surveillance were performed. Power levels were adjusted accordingly at the request of the Load Dispatcher and Shift Engineer.

On June 29, at 2239 hours0.0259 days <br />0.622 hours <br />0.0037 weeks <br />8.519395e-4 months <br />, several alarms were received in the control room and-the' reactor scrammed due to a turbine trip.

Investigations-began into the cause of this scram and the unit remained shutdown for the remainder of the month.

B.

Unit Two Unit.Two began the month of June in the SHUTDOWN mode (due to drywell floor drain sump pumps-failing). Maintenance was completed on the pumps and startup procedures were commenced. At 0049 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br />, on June 1, the reactor was made critical. At 1137 hours0.0132 days <br />0.316 hours <br />0.00188 weeks <br />4.326285e-4 months <br />, the mode switch was riaced to RUN. At 1809 hours0.0209 days <br />0.503 hours <br />0.00299 weeks <br />6.883245e-4 months <br />, the generator was synchronized to the reid. A load increase was begun with control rods. On June 2, at 1029 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.915345e-4 months <br />, 740 MWe was achieved. At 1032 hours0.0119 days <br />0.287 hours <br />0.00171 weeks <br />3.92676e-4 months <br />, it was noticed that corJenser vacuum was deteriorating. A load decrease was taken to help recover vacuum. Power levels were held constant at 480 MWe.

Rods were inserted, but condenser vacuum continued to deteriorate. A further load reduction was taken, vacuum problem continued. At 1208 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.59644e-4 months <br />, power levels were reduced to 80 MWe.

Torus cooling was on in anticipation of loss of condenser. At 1208 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.59644e-4 months <br />, condenser vacuum began to improve and control rods were withdrawn to increase power levels to 250 MWe.

Power levels were held constant until June 3 when a further load increase I

was taken to full power.

818 MWe was achieved at 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br />.

Normal operational plant activities and surveillance were performed through June 8.

At 2312 hours0.0268 days <br />0.642 hours <br />0.00382 weeks <br />8.79716e-4 months <br />, the unit was placed in EGC.

The unit remained in EGC until June 9.

At 2350 hours0.0272 days <br />0.653 hours <br />0.00389 weeks <br />8.94175e-4 months <br />, EGC was tripped and a load reduction was taken per the Load Dispatcher. On June 10, at 1005 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.824025e-4 months <br />, a power increase was taken to 815 MWe.

At 1350 hours0.0156 days <br />0.375 hours <br />0.00223 weeks <br />5.13675e-4 months <br />, load was decreased to 450 MWe at the request of the Load Dispatcher.

On June 11, at 0315 hours0.00365 days <br />0.0875 hours <br />5.208333e-4 weeks <br />1.198575e-4 months <br />, a further power reduction was taken to 250 MWe per the Load Dispatcher.

Power levels were held constant until June 12.

At 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />, the unit I

was taken to full power.

From June 14 through June 30, normal plant operational activities and routine surveillance were performed.

Power levels were adjusted accordingly.

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T III.

PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A.

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

Q71631 LER NUMBER: N/A COMPONENT:

System 201 - While performing QOS 201-1, Manual.0peration of Electromatic Relief Valves, acoustic monitor 1-261-60A for the 1-203-3A relief valve did not respond when the valve was opened. Turbine bypass valve closure and relief valve exhaust line temperature indicated that the relief valve did actually open.

CAUSE OF MALFUNCTION: A preliminary investigation by the Instrument Maintenance Department showed that the problem was inside the drywell and not with the electronic components between the drywell and the control room.

The monitor had failed twice in the past year due to faulty cable connections. Cable connections are believed to have played a part in this failure. The problem was investigated and repaired under Work Request Q71631.

RESULTS & EFFECTS ON SAFE OPERATION:

Since the thermocouple (TE 1-261-14A) in the tailpipe was still operable and capable of detecting valve leakage, the safety consequences of the event were considered to be minimal.

ACTION TAKEN TO PREVENT R,EPETITION: Work Request Q71631 was written to repair the acoustic monitor.

WORK REQUEST No.:

Q73574, Q73575 LER NUMBER: N/A

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COMPONENT:

System 2300 - Replaced existing splices for connecting the pigtails of the Rosemount conduit seal to its associated power leads with a qualified Environmentally Qualified (EQ) splice.

EPN 1-2352 and 1-2353.

CAUSE OF MALFUNCTION: It was determined that the signal leads for the Rosemount conduit seal pigtails on Rosemount transmitters 1-2352, 2353, and 2389A, B, C and D had questionable EQ splicing configurations. A manage-ment deficiency error was determined to be the reason for the improper installations.

There was a lack of instruction and detail in the work package to insure that the splices were installed in a tested EQ configura-tion that was qualified per 10CFR50.49.

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

ACTION TAKEN TO PREVENT REPETITION:

Splices for 1-2352 and 1-2353 were replaced and Work Requests Q73574 and Q73575 were generated.

In order to be assured that all EQ criteria have been met, Quad Cities will perform a 100% walkdown of all identified EQ cable splices in harsh environments.

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

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IWORK REQUEST No.: Q73994 LER NUMBER: N/A COMPONENT: System 1000 - While performing QOS 1000-9 Residual Heat Removal (RHR) System Power Operated Valve-Testing, the breaker for the motor-operated (MO) 1-1001-20 valve, inboard isolation for RHR reject to radwaste, tripped while the Nuclear Station Operator (NS0) was attempting to open the valve.

The breaker was reset and tripped again when' attempting to open the valve.

The problem was investigated on Work Request Q68944.

CAUSE OF MALFUNCTION: The cause of the event was inadvertent pinching of the motor leads during the replacement of the motor cover after an inspection.

A contributing cause was the small size of the operator, making it necessary to take extra care when replacing the cover.

RESULTS 6 EFFECTS ON SAFE OPERATION: The safety significance of the event was minimal. The MO 1-1001-20 valve was in the closed position at all times.

This is the isolation position for the valve.

ACTION TAKEN TO PREVENT REPETITION: The MO 1-1001-20 valve was repaired under Work Request Q73994. Work Requests Q72964 and Q72966 were written to inspect other valves. In addition, training will be given to electricians to insure awareness of the care that should be taken when replacing motor operator covers.

To insure that such events are promptly discovered, valve stroking will be performed as identified in the post-maintenance testing program.

WORK REQUEST No.: Q75124 LER NUMBER:

89-004 COMPONENT: System 203 - While the Operating Department was performing QOS 201-1, Auto Pressure Relief System Manual Operation of Relief Valves, the 1-203-3D relief valve would not reseat after it was opened. The position indication on the pilot solenoid showed that the pilot valve was closed, but the acoustic monitor, exhaust temperature and bypass valve positions all indicated that the valve was open. The reactor was manually scrammed after several failed j

attempts to close the valve.

CAUSE OF MALFUNCTION: The cause of the event was determined to be component f ailure. The relief valve stayed in the open position due to a combination of failures. The pilot valve showed signs of steam leakage past the seat, and the 1/16" drain orifice in the disc retainer of the main valve was plugged with a small piece of metal. The steam leakage past the pilot valve seat was normal wear.

I RESULTS & EFFECTS ON SAFE OPERATION: The safety significance for the event was minimal. All ESF actuations occurred as expected to bring the reactor to a safe shutdown condition. The relief valve closed when the spring pressure overcame reactor pressure prior to 20 psig, and the Unusual Event was terminated when the reactor was in a cold SHUTDOWN condition. In addition, High Pressure Coolant Injection Residual Heat Removal and Core Spray systems were available at all times during the occurrence to supply water to the reactor vessel.

However, normal reactor feedwater was adequate to control level.

i ACTION TAKEN TO PREVENT REPETITION:

Immediate corrective action was to replace

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the 1-203-3D electromatic relief valve, including its pilot valve. This was done i

under_ Work Request Q75124. The pilot valves for the 1-203-3B and 1-203-3E were replaced as a preventative measure. A representative of Dresser Industries, manufacturer of the Electromatic Relief Valves (ERV), was present at the valve disassembly, and will provide a report detailing activities that can be implemented to enhance the reliability of the ERV's and pilot valves.

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4 WORK REQUEST NO.:

Q76283 LER NUMBER:

89-008 COMPONENT:

System 1600 - While Operating personnel were performing QOS 1600-15, Pressure Suppression Systems Power-Operated Valve Fail-Safe Testing, it was determined that Air-Operated (AO) valve 1-1601-23 would not fail in the closed (fail-safe) position. Valve 1-1601-23 is the inboard veat isolation valve to the drywell and is required to fail in the closed position for Primary

. Containment Isolation (PCI). The valve.was declared inoperable, and QOS 1600-01, Containment / System Isolation Valve Inoperable Outage Report, was initiated.

An investigation of the manifold on the air operator for valve 1-1601-23 showed that.the check valves installed in the air supply lines to valves 1-1601-24 and 1-1601-63 were not all the same.

Several valves were check valves and several were flow / check valves.

CAUSE OF MALFUNCTION: The apparent cause of the failure of valve 1-1601-23 air operator was the use of a flow / check valve in place of a check valve.

This flow / check valve was installed initially in the air operator. When installed backwards with the needle valve fully closed, the flow / check valve acts like a check valve.

- RESULTS & EFFECTS ON SAFE OPERATION: This event was potentially significant because in the event of a loss of instrument air, the flow / check control valves.would not have acted as check valves.

This would have potentially prevented the valve from failing in the safe position.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to take the affected valves out of service in their respective fail-safe positions. The flow / check control valves were replaced with check valves in the 1-1601-23 air operator under Work Request Q76283. Dresden and LaSalle Stat #.cn personnel were contacted and informed of the problem.

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6 UNIT 2 MAINTENANCE

SUMMARY

l WORK REQUEST NO.:

Q65775, Q65776, Q65777, Q65778, Q65779, Q65780, Q65781, Q65782, Q65784 LER NUMBER: 88-006 COMPONENT:

System 2300, 1000, 1400, 3700, 300, 0020 - With Unit 2 operating at 93% of rated thermal power, tne station was notified by the Boiling Water Reactor Engineering Department (BWRED) that eleven flued head anchors did not meet the design requirements specified in the Quad-Cities Final Safety Analysis Report (FSAR).

CAUSE OF MALFUNCTION:

It was discovered that the flued head anchor structures at Dresden and Quad-Cities were not included under the I.E. Bulletins No. 79-14 and 79-02 scope of work. The exclusion of 'the structures in the 79-02 and 79-14 programs was due to misinterpretation of the scope requirements.

There-fore, analysis for these structures was not reassessed for design base requirements.

RESULTS & EFFECTS ON SAFE OPERATION: The health and safety of Gie public and of plant personnel was not adversely affected by this event.

Since the anchor assemblies were analyzed and considered operable, the associated safety significance was minimal.

ACTION TAKEN TO PREVENT REPETITION: Modification M-4-2-88-017 was initiated to revise the structures to a condition that complies with FSAR design require-ments. Work Request Q65775 was written to repair HPCI support X-11.

Work Requests Q65776 a..d Q65777 were written to repair RHR supports X-13A and X-13B.

Work Requests Q65778 and Q65779 were written to repair core spray supports X-16A and X-16B.

Work Requests Q65780 and Q65781 were written to repair cooling water supports X-23 and X-24.

Work Request Q65782 was written to repair CRD Return Line Support X-36.

Work Request Q65784 was written to repair MSIV supports X-7A, X-7B, X-70, X-7D, X-8, X-9A, X-9B, X-10, X-12 and X-17, i

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WORK REQUEST No.: Q69723 j

LER NUMBER: N/A COMPONENT:

System 1000 - While the Instrument Maintenance Department was performing QIS 6-1, High Drywell Pressure SCRAM Calibration,- it was discovered that pressure switch (PS) 2-1001-88A tripped at 2.56 psig, instead of less than or equal to 2.5 psig as required by Technical Specifications. The switch was recalibrates to 2.41 psig.

l' CAUSE OF MALFUNCTION: The cause of the event was instrument setpoint drift.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plant and personnel was not affected during this event. High drywell pressure switches 2-1001-88A, P, C and D provide SCRAM and Group 2 isolation functions. These functions would not have been inhibited by the setpoint drift of this single pressure switch because the switches are arranged in one-out-of-two taken twice logic and the other three switches were within Technical Specification limits.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to calibrate the pressure switch to 2.41 psig. This switch was previously replaced per D/R 4-2-88-004.

The pressure switch was replaced again under

' Work Request Q69723. The removed pressure switch was sent to Static-0-Ring (the manufacturer) for examination.

WORK REQUEST NO.:

Q72189 LER NUMBER: N/A COMPONENT:

System 700 - While performing QOS 700-9, Traversing Incore Probe System Power Operated Valve Testing, the ball valve, 1-737-1B, on machine two would not close. This valve is required to close upon receipt of a Group 2 isolation signal. The shift engineer issued the Nuclear Station Operator (NS0) the key for the Traversing Incore Probe (TIP) shear valve, 1-737-2B.

The shear valve is capable of isolating the tip tubing and probe should a Group 2 isolation be required. After further attempts to close the valve from the control room, an equipment operator was dispatched to the TIP room to close the manual isolation valve, 1-4799-490B. The manual isolation valve was then taken out of service to insure primary containment.

CAUSE OF MALFUNCTION: The ball valve was removed under Work Request Q72189 and was to be rebuilt under Work Request Q72280. The cause of the failure will be determined under Work Request Q72280.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of this event were minimal due to the issuance of the shear valve key and the short period of time between the ball valve failure and manual valve closure.

In the event of a Group 2, the NSO would have been able to isolate 'he line with the shear valve, if required.

ACTION TAKEN TO FREVENT REPETITION:

The immediate corrective action was to issue the NSO the key to the shear valve and dispatch an equipment operator to close the manual isolation valve. The ball valve and manual isolation valves were taken out of service to insure primary containment integrity.

The ball valve was replaced like-for-like.

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" WORK REQUEST NO.: Q74748 i

LER NUMBER: N/A l

COMPONENT:

System 700 - While operating at 100% power, Quad Cities Unit 2

. received a half scran on Channel A of the Reactor Protection System (RPS).

This was due to a Hi-Hi signal from Average Power Range Monitor (APRM) Channel 1.

A check of APRM Channel I revealed that the DC (+) power supply for the APRM was,found to be indicating downscale.

CAUSE OF MALFUNCTION: The cause of the event was thought to be a problem with the APRM's power supply, but the power supply was found to be functioning properly during the Instrument Maintenance Department investigation of the component. Although the cause was unknown, the decision was still made to replace the power supply.

RESULTS & EFFECTS ON SAFE OPERATION:

The safety significance of RPS Channel A tripping due to an APRM Hi-Hi signal from APRM Channel 1 was minimal. The RPS system requires two channels (A and B) to trip before the reactor is scrammed. A spurious signal causing one channel to trip is not significant due to this two channel design of RPS.

If the reactor had scrammed, the reactor would have gone to a safe shutdown condition.

ACTION TAKEN TO PREVENT REPETITION: The APRM power supply was replaced under Work Request Q74748. As a precaution, the circuit breaker for the power supply was replaced under Work Request Q74866.

WORK REQUEST No.:

Q75462, Q75463 LER NUMBER:

N/A COMPONENT:

System 1300 - With Unit 2 at 100% rated core thermal power, a walkdown was being dconducted of butt splices within junction boxes and pull boxes. The walkdown was prompted by concerns involving butt splices related to Environmentally Qualified (EQ) equipment at Dresden Station. As a result of the walkdown, four butt splices were discovered without a qualified jacket.

The unqualified splices were found on cables 20107 and 20108, which feed temperature switches (TS) 2-1360-16B and 2-1360-17B.

CAUSE OF MALFUNCTION: The instrument cables for the 2-1360-16B and 2-1360-17B temperature switches are original plant equipment. At the time of their installation, EQ requirements were not yet in place.

The failure to recognize the possibility of unqualified splices in pull and junction boxes associated with EQ equipment led to them being overlooked after Environmental Qualifications was required.

RESULTS & EFFECTS ON SAFE OPERATION: Due to the mild environment, the thermal aging experienced by the splices was insignificant.

BWRED concluded that the subject splices were qualified and operable for the required operating time.

Thus, the safety consequences of the event were minimal.

ACTION TAKEN TO PREVENT REPETITION: Work Requests Q75462 and Q75463 were generated to replace the unqualified splices.

In addition, a walkdown of environmentally qualified equipment splices and terminal blocks was conducted to identify any plact equipment that may have been undetected during original qualifications of equipment.

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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 Occurrence 89-006 6-5-89 1A Recirc Pump Trip-While Looking for Grounds89-007 6-9-89 1A Refuel Floor Rad Monitor Spiked High 89-008 6-7-89 1-1601-23 Valve Failed Fail Safe Test 89-009 6-8-89 Unmonitored Release of Laundry Drain Water to Discharge Bay UNIT 2 There were no Licensee Event Reports for Unit 2 for this reporting period.

0027H/00612 1

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V.

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

l A.

Operating Data Report l

B.

Average Daily Unit Power Level C.

Unit Shutdowns and Power Reductions 0027H/0061Z

--__-_____--___-___--___A

.R APPENDIX C-L 1

OPERATING DATA REPORT

\\

L 1

I DOCKETNO. 50-254 UNIT ' One DATE July 11. 1989

)

I COMPLETED BY LYnne Deelsnyder 1

[

TELEPHONE 309-654-2241 OPERATING STATUS 0000 060189

-l 720 t REPORTING PERs00, 2400 063089 GROSE HOURE IN REPORTING PER800:

i 2511 Max. DEPENO. CAPACITY m 769

& CURRENTLY AUTHORIZED POWER LEVEL (g9 DSSION ELuiCTRICAL MATING (MWo.Neth N/A

3. POWGR LEVEL TO WHICM RGETRfCTED (17 ANYI (MWo. Net):

A, REASONS POR RESTRICTION (IP ANYl:

THIS MONTH YA TO DATE CUMULATIVE

..... 694.6 4127.4 121669.6 E. NUMOOR OF MOURS REACTOR WAS CRITICAL.........

0.0 0.0 3471.9_

8. REACTOR RESERVE SMUTOOWN MOURE................... 694.6 4058.3 117717.5

7. MouRs GENERATOR oN LINE..........................

0.0 0.0 909,2 E. UNIT RESERVE SMUTDOWN MOURS......................

s. Gross THERMAL ENERGY GENERATED (MWM)........... 1438058 9206759 250896838 456672 2985610 81343223

10. GROSS RLECTRICAL ENERGY GENER ATED (MWM).............

11. NET ELECTRICAL ENERGY GENERATED (MWM).............. 434752 2354612 76419886 96.5 95.0 81.0

12. REACTOR SERvlCE P ACTOR......

96.5 95.0 83.3_

13. REACTOR AV AILAstLITY P ACTOR......................

96.5-93.4 78.4

14. U NIT SERVICE P ACTOR.............................. 96.5 93.4 79.0

15. UNIT AV AILAel LITY P ACTOR..........................

78.4 72.5 65.8

18. UNIT CAPACITY P ACTOR (Using MOCl 76.4 70.6 64.1

17. UNIT CAPACITY PACTOR (Usans Demon MWel.,..............

3.5 6.6 5.4

18. UNIT PORCEO OUTAGE MATE.......

19. SMUTDOWNS SCH800 LEO OVER NEXT E MONTHS (TYPE. DATE. ANO QURATION OF EACMh

20. IP SMUT DOWN AT END OF REPORT PERIOD. ESTIMATED DATE OP ffARTUP:

21. I a.'*TS IN TEST STATUS (PRIOR TO COMMERCIAL QPERATIONi:

PORECAST ACMIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 1.lM

APPENOfX C I

OPERATING DATA REPORT DOCKET NO.

50-265 UNIT h

DATE July 11, 1989 COMPLETE 0 8Y Lynne Deelsnyder l

TELEPNONg 309-654-2241 OPERATING STATUS 0000 060189 720 2400 063089 GROSB MOURS IN REPORTING PERICO:

1. REPORTues PER800:

2. CURRENTLY AUTMont2EO powen LEVEL (MW3g ' 2511 max.ogetNo.CAPActTY ttRuedesel. 769 0084088 ELECTRICAL RATING (Mus 8esch

/09 N/A 1 POWGR LEVELTO WUMtCN RESTRICTED (IP ANY) (MWs4esch

4. REASONS POR RESTRICT 10N llP ANY):

TMt3 MOfffM YM70 3A73 CyMg(Argyg 4226.6 115176.5

5. NUMSSR OF MOURS REACTOR WAS CRITICAL............. 719. 2 0.0 0.0 2985.8_

8. REACTOR ReBERVE SMUTDoupe MOURs...................

4184.6 111916.3

f. MOURS GENERATOR 001 LING......................... 7 01. 9 0.0 0.0 702.9

8. UNIT RESERVE SMUTDOWN MOURS,......................

............. 1373503.2 9302518.2 240212791.2

9. GRoeS THERMAL ENERGY GENERATED (MyWMi 438782 3028609 76962080

10. Gross ELECTRICAL ENERGY GEN 8 MATED (MWMi.............

417996 2898995 72635572

11. NET ELECTRICAL ENERGY GENERATED (MWHI.............

99.8 97.3 77.1

12. ' EACTOR SERVICE P ACTOR..................

99.8 97.3 79.1

13. REACTOR AV AILAtiLITY P ACTOR......................

97.5 96.4 74.9

14. UNIT SERvlCE P ACTOR........................

97.5 96.4 75.4

15. UNIT AV AILASILITY P ACTOR..........................

84.0 74.3 62.9 '

18. UNIT CAPACITY PACTOR susing MOCl 81.9 72.4 61.3

17. UNtf CAPACITY PACTOR (Uesag Desip Munst.................

2.5 3.6 8.3

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

l

19. SMUTDOWNS SCHEDULE 0 OVER NEXT E MONTHS (TYPE. DATE. AND OURATION OF EACMh i

20. IF SMUT DOWN AT END OF REPORT PERICO. ESTIMATED DATE OF STARTUP:

21. UNITS IN TEST STATUS (PRIOR TO COMMERCI AL QPERATIONh PORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION 1.IH

I-'

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1 i

APPENDIX 8 i

AVERAGE DAILY UNIT POWER LEVEL j

DOCKET NO.

50-21.,

l UNIT one i

DATE July 11, 1989 COMPLETED BY Lvnne Deelsnyder, TELEPHONE 309-654-2241 MONTH June, 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe Net) i 599 624 37 641 624 2

gg 697 693 3

3, 510 694 4

3

[

5 576 21 657 661 758 8

y y

655 747 23 8

619 690 24 g

490 621-g 10

. 543 701 g

362 704 39 y

12 -

486 682 g

13 668 3

595 14 622 3

-14 15 524 39 16 651 INSTRUCTIONS On this form, list the average daily unit power level in MWe. Net for each day in the reporting month. Compute to the neasest whole megawatt.

These figures will be used to plot a graph for cach reporting month. Note that when rnaximum dependable capacity is u ed for the net electrical rating of the unit, there may be occasions when the daily average power level exceeds the 100'# line (or the restrwicd power level line). In such cases, the average daily unit power output sheet should be footnoted to explam the apparent anomaly.

1.16 8 e

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. ' APPENDIX B AVER AGE DAILY UNIT POWER LEVEL i

DOCKET NO.

50-265 UNIT Two DATE Julv 11, 1989 COMPLETED BY 1,ynne Deelsnyder TELEPHONE 309-654-2241 MONTH June, 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe. Net) 1 18 17 575-2 268 633 g,

3' 457 700 g,

4 431 659 20 f'

627.

5 586 21 6

639 637 22 751 7

724 23 683 8

758 24 g

676 559 25 686 10*

565 g

691 11 366 27 12 455 650 28 13 624 527

.y 607 14 669 30 550 15 3,

16 658 INSTRUCTIONS On this form, list the average daily unit power level in MWe-Net for each day in the reporting month. Compute to the neasest whole megawatt.

These figures will be used to plot a graph for cach reporting month. Note that when rnaximum dependable capacity is used for the net electrical rating of the unit, there may be occasions when the daily average power level exceeds the

.100't line (or the restneted power level line). In such cases. the average daily unit power output sheet should be footnoted to expl.sua the apparent anomaly.

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VI. UNIQUE REPORTING REQUIREMENTS r

The following items are included in this report based on prior commitments to the commission:

1 A.

MAIN STEAM RELIEF VALVE OPERATIONS Relief valve operations during the reporting period are summarized in the following table. The table includes information as to which reli:.

j valve was actuated, how it was actuated, and the circumstances resulti:

in its actuation.

Unit: One Date: June 29, 1989 Valves Actuated.

No. & Type of Actuation 1-203-3C 1 Automatic Plant Conditions: Reactor Pressure - 1105 Description of Events: Automatic actuation following reactor scram from a turbine trip.

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/0061Z

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^

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.

l l

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0027H/00612 w-

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QTP 300-532 Revision 1-QUAD-CITIES REFUELING March 1978 l

INFORMATION REQUEST

~

1.

Unit:

01 Reload:

9 Cycle:

10 2.

Scheduled date for next refueIIng shutdown:

9-9-89 1

3 Scheduled date for restart following refueling:

12-11-89 i

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

j NOT AS YET DETERMINED.

5.

Scheduled date(s) for submitting proposed IIconsing action and supporting i

information:

JUNE 10, 1989 6.

Important IIconsing 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 assembites, a.

Number of assemblies in core:

724 b.

Number of assemblies in spent fuel pool:

1773 8.

The present Ilconsed 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 fusi assemblies:

a.

Licensed storage capactty for spent fuel:

us7 b.

Planned increase in licensed storage:

0 9.

The projected date of the last refueling that can be dische to the spent fuel pool assuming the present licensed capacity: 2 X P P R O. V E D ' APR 2 01978 Q. C. o. S. R.

1

r-

)

.C QTP 300-532 Revision 1 QUAD-CITIES REFUELING March 1978 INFORMATION REQUEST

~

1.

Unit:

'02 Reload:

9 Cycle:

10 J

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 i

specification change or other license amendment:

)

i NOT AS YET DETEltMINED.

5.

Scheduled date(s) for submitting proposed licar.r*r.s action and supporting Information:

NOVEMBER 2, 1990 i

6.

Important Ilconsing considerations associated with refueling, e.g., new or I

'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 b.

Planned increase in licensed storage:

0 9

The projected date of the last refueling that can be discharged to the spent fuel pool asstaning the present licensed capacity: 2008 XPPROVED APR 2 01978 Q. C. O. S. R.

1

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l 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 Hithout Scram BWR Boiling Hater Reactor CRD Control Rod Drive EHC Electro-Hydraulic Control System EOF Emergency Operations Facility GSEP Generating Stations Emcrgency 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 i

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 North Minimizer SBGTS standby Gas Treatment System SBLC Standby Liquid Control SDC Shutdown Cooling Mode of RHRS SDV Scram Discharge Volume SRM Source Range Monitor TBCCH Turbine Building Closed Cooling Hater System TIP Traversing Incore Probe Technical Support Center TSC 0027H/0061Z

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