ML20011D159

Monthly Operating Repts for Nov 1989 for Quad-Cities Units 1 & 2

ML20011D159
Person / Time
Site:	Quad Cities
Issue date:	11/30/1989
From:	Deelsnyder L, Robey R COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
RAR-89-81, NUDOCS 8912210119
Download: ML20011D159 (26)

Text

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) ound Cmes Nuclear Power S: tion i

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Cordova. Illinos 61242 9740 i

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December 1, 1989 J

Director of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Mail Station Pl-137 Washington, D. C.

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1 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 November, 1989.

Respectfully, COMMONNEALTH EDISON COMPANY

' QUAD-CITIES NUCLEAR POWER STATION i

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.g-ng R.A.hobe,y Technical Superintendent RAR/LFD/djb Enclosure ll 0027H/0061Z 8912210119 ep33go l{DR1ADOCK05000254 PDC.

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QUAD-CITIES NUCLEAR POWER STATION

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UNITS 1 AND 2 i

MONTHLY PERf0RMANCE REPORT NOVEMBER, 1989 i

COMMONWEALTH EDISON COMPANY AND IOWA-ILLINOIS GAS & ELECTRIC COMPANY NRC DOCKET NOS 50-254 AND 50-265 LICENSE NOS. OPR-29 AND DPR-30 t.

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I TABLE OF CONTENTS I.

Introdvetion 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 i

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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 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 On"ket 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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SUMMARY

OF OPERATING EXPERIENCE f

A. ' Unit One

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L The End'of Cycle Ten Refueling Outage activities continued normally for Unit One during the month of November. On November 5 thru November 7 reactor

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vessel hydro testing was successfully performed.

Beginning~ November 8, I

f preparations were made for the performance of the Integrated Leak Rate Test which was successfully completed on November 16.

From November 16 to L

~ November 22. normal outage activities continued.

On November 23. the mode i

switch was placed in STARTUP, the master startup checklist was completed,

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and the. reactor became critical at 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br />.

On November 24, at 1050 hours0.0122 days <br />0.292 hours <br />0.00174 weeks <br />3.99525e-4 months <br />, l

l an alarm was received indicating 'A' recirculation pump inboard seal high-L temperature. The pump was tripped and isolated, and a seal leak from the drywell was reported. Upon investigation, it was discovered that the coupling

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on the recirculation pump seal pressure sensing line was the source of leakage.

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At 1454 hours0.0168 days <br />0.404 hours <br />0.0024 weeks <br />5.53247e-4 months <br />, a normal shutdown was commenced.

Repairs were made to the

. sensing line.

On November 25, reactor startup was commenced, and the reactor I

became critical at 1105 hours0.0128 days <br />0.307 hours <br />0.00183 weeks <br />4.204525e-4 months <br />. However, due to problems with

'A' recirculation pump seal, a normal shutdown was commenced on November 26, at 0806 hours0.00933 days <br />0.224 hours <br />0.00133 weeks <br />3.06683e-4 months <br />.

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The seal was replaced.

On November 27, reactor startup was commenced. At 2242 hours0.0259 days <br />0.623 hours <br />0.00371 weeks <br />8.53081e-4 months <br />, the reactor became critical.- On November 28, at 1245 hours0.0144 days <br />0.346 hours <br />0.00206 weeks <br />4.737225e-4 months <br />, the 1

main generator was synchronized to the grid. Power levels were held at 125 INe.

At 1705 hours0.0197 days <br />0.474 hours <br />0.00282 weeks <br />6.487525e-4 months <br />, hot scram timing for the control rod drive system was begun.

e On November 29, at 1215 hours0.0141 days <br />0.338 hours <br />0.00201 weeks <br />4.623075e-4 months <br />, scram timing was successfully completed.

For the remainder of the month, power levels were held constant while additional testing was performed before an ascent to full power could be taken.

'B.-

Unit Two d

Unit Two began the month of November operating at full power. On November 1,

-at-1827: hours, the unit was placed in Economic Generation Control (EGC).

The unit remained in EGC with minor interruptions until Novembe-4 when a

power reduction to 200 MWe was taken in preparation for drywell entry for

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' determination of leakage sources.

On Fr.rember 5, load was increased after the drywell entry was completed. At 1405 hours0.0163 days <br />0.39 hours <br />0.00232 weeks <br />5.346025e-4 months <br />, full power was achiaved.

On November 6, at 1809 hours0.0209 days <br />0.503 hours <br />0.00299 weeks <br />6.883245e-4 months <br />, power levels were adjusted and the unit was placed in ECC.

For the remainder of the month, normal operational activities and routine surveillances were performed. The unit remained near full power or operated in'EGC per the demands of the Chicago Load Dispatcher.

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Ill. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS. AND SAFETY RELATED MAINTE!ANCE A.,

Amendments to Facility License or Technical Specifications On October. 20, 1989.. the NRC issued Amentment 120 to License DPR-29.

This' amendment eliminated cycle specific and fuel specific thermal

. limits. These. limits have been incorporated into a Core Operating Limits Report.

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On November 15,'1989, Amendment 121 to License DpR-29 was issued.

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This amendment reduced the number of instrument channels required r

for HPCI-and RCIC area high temperature isolations.

The setpoints for these area high temperature suitches was also reduced.

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 Reepiring NRC Approval There were no Tests or Experiments requiring NRC approval for the reporting period.

D.

Corrective Mainteuance of Safety Related Equipment The following represents a tabular summary of the major safety related maintenance performed on Units One and Two during the rerorting period.

This summary includes the following: - Work Request Numbers, Licensee Event. Report Nuubers, Components Cause of Malfunctions Results and Effects on Safe Operation, and Action Taken to prevent Repetition.

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SUMMARY

WORK REQUEST NO.

Q72279 L

' LER NUMBER: N/A COMPONENT:- System 1400 - While Instrument Maintenance (IM) personnel were performing QIS 23-2, Core Spray Pump Discharge Pressure Functional, pressure

. switch PS-1-1462C would not reset. The switch was manually reset by pushing

{;t on the bourdon tube. The alarm and switch reset. Q1S 23-1, Core Spray Pump Discharge Pressure Calibration, was done to verify that the switch still tripped L

- at the proper setpoint. The functional test, QIS 23-2, was repeated several times with no' repeat of the original failure.

CAUSE OF MALFUNCTION: The cause of the event was undetermined. The pressure switch PS-1-1462C was replaced under WR Q72279.

RESULTS 6 EFFECTF ON SAFE OPERATION: The safety significance of the event was minimal. At no time was;the logic for Auto Blowdown inoperative.

There are two pressure switches per Core Spray Pump that provide input to the auto blow-l_

down logic.

Both switches need to pick up to allow the relief valves to open.

Since the switch was in the tripped position, it would not have prevented actuation.

t ACTION TAKEN TO PREVENT REPETITION: The pressure switch was replaced under WR Q72279.-

WORK REQUEST NO : Q75133 L,ER NUMBER: _89-004 COMPONENT: ' System 203 - While performing QOS 0201-S1, " Auto Pressure Relief System Manual Operation of Relief Valves", the 1-203-3D valve. stuck oren.

After numerous unsuccessful attempts to close the valve, the reactor was manually scrammed per procedure. All other systems operated as expected.

An Unusual Event was declared.

CAUSE OF MALFUNCTION: The cause of the event was component failure because of. steam leakage past the pilot valve seat. Adequate steam pressure could not be attained to close the main valve disc.

RESULTS & EFFECTS ON SAFE OPERATION:

The safety significance of the event'was' minimal. All Engineered Safeguard Feature (ESP) actuations occurred as expected to' bring the reaetor to a safe shutdown condition. The relief valve closed

- when. spring tension overcame reactor pressure, and the Unusual Event was terminated when'the reactor was in a COLD SHUTDOWN condition.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to l-replace the 1-203-3D electromatic relief valve, including its pilot valve, under WR Q75124. The relief valve was rebuilt under WR Q75133. The pilot valves for the other relief valves were replaced and rebuilt as a precautionary measure.

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c WORK REQUEST NO : Q78238, Q78239

LER NUMBER: N/A

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COMPONENT:

System 1600 - While in RUN mode at 77% of rated core thermal power.

l abnormal indications for containment post-accident monitoring suppression chamber 7

level indication 1-1640-10A and containment pressure indication 1-1640-IIA were i

discovered during normal control room panel checks. The two indicators and a

the chart recorder (1-1640-13A) were found downscale. Other Control Room sup-

.pression chamber level and containment pressure indicators revealed normal i

readings.. The two indicators were declared inoperable and Outage Report QOS 1600-05, was initiated. The Shift Engineer initiated Work Requests Q78238 and Q78239 for

the Instrument' Maintenance Department to investigate. A loose connection was found on a common negative voltage point between the two indicators. The loose-j connection was tightened and indication returned to normal.

CAUSE OF MALFUNCTION: The apparent cause of the event was attributed to a loose v

terminal connection on the negative side of the power supply. This caused the t

spiking downscale of the indicators.

r RESULTS & EFFECTS ON SAFE OPERATION: The safety significance of this event was minimal. A thirty-day Limiting Ccndition of Operation was established due to the inoperability of torus level indicator 1-1640-10A, as per Technicc1 Speci-l fications. Redundant torus level indication was available from level indicators 1-1640-10B, 1-1607-7, and the local torus level sightglass.

Therefore, sufficient means were available to monitor the level of the suppression pool and containment pressure throughout the event.

ACTION TAKEN TO PREVENT REPETITION: Work Requests Q78238 and Q78239 wers written to investigate the problem.

The loose connection was tightened and the readings-returned to normal.

To prevent future similar equipment problems, wiring in the control room panels will be checked for loose connections during refuel outages.

i WORK REQUEST No.:

Q78333, Q78334, Q78680 LER NUMBER:

89-014; 7

COMPONENT:- Systems 203, 1600 - While performing local 12ak rate testing (LLRT) on the Drywell/ Torus purge. volume, consisting of valves A0-1-1601-23, 24, 60, 61, 62, 63, the volume exceeded the maximum allevable leakage as designated by Technical Specifications.

In addition, a LLRT of two Main Steam Isolation Valves (MSIV). A0-1-203-2A and A0-1-20s-2D, showed a leakage in excess of Technical Specification limits.

CAUSE OF MALFUNCTION: The cause of.the leakage could not be deterrined at the time of this report. A supplemental report is to be filed documenting repairs.

RESULTS & EFFECTS ON SAFE OPERATION: The satety consequences of this event were minimal since LLRT is a conservative method for quantifying containment leakage.

l The actual leakage under accident conditions would be less than chat determined by

,LLRT because some lines would be filled with water rather than air, and some lines would be isolated by non-primary containment isolation valves. Also, where more than one valve is present in a line, as in the case of the MSIV's, it is realistic t'o expect the leakage to

.,e equal to the lesser leakage of the two valves. However, the maximum pathway leakage is used for comparison with Technical Specification requirements.

ACTION TAKEN TO PREVENT REPETT. TION: Valve 1-1601-60 was suspected to be one of the leaking valves in the Drywell/ Torus purge volume. Work Request Q78680 was written to disassemble and repair valve. Valves A0-1-203-2A and A0-1-203-2D were l

repaired under Work Requests Q78333 and Q78334, respecti y.

Further corrective action will be submitted in the supplemental report.

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h WORK REQUEST No.: 'Q78918

'LER NUMBER: N/A COMPONENT:

System 6600 - With the shared 1/2 Diesel Generator (DG) out-of-service 7

for planned maintenance and Unit One in the REFUEL mode. U-2 was in a Limiting Condition of Operation (LCO). The Unit Two DG was started and loaded using QOS i

6600-1, Diesel Generator Monthly Load Test, as required by QOS 6600-3, Shared Unit n

(1/2) Diesel Generator Outage Report Seven-Day Limitation. While the U-2 DG

.was still running, the operator in attendance noticed lube oil leaking from the local engine control panel.

Th3 Shift Foreman dispatched to investigate requested that the DG be shut down.' The rubber lube oil hose connecting the.DG to the MB2 oil pressure switch was found to have split. The U-2 DG was then declared inoperable E

and Unit Two was placed in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO. Work Request Q78918 was generated to replace h

the split oil hose and other rubber hoses in the control panel.

The U-2 DG was then restarted and declared operabic.

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CAUSE OF MALFUNCTION:

The cause of the event was the failure of the lube oil O

hose. A contributing cause was the procedure lack of specific inspection require-ments. QMMS 6600-1-S4, Diesel Inspection - Refueling Outage Checksheet, merely states that all rubber lines are to be inspected for damage, cracking and wear.

No specific accept / reject criteria is given.

RESULTS & EFFECTS ON SAFE OPERATION:

Technical Specifications state that if an Emergency Diesel Generator (either that of the unit or the shared diesel), then an orderly shutdown of the unit must be done within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The U-2 DG was in STOP for maintenance for a period of only 58 minutes. Two offsite lines were available throughout this period.- Had there been an energency situation, the leak could have been isolated to-allow operation of-the DG, ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to-Tdetermine the-location of the leak.

When the leak was located, the DG was shut-i down and all the rubber hoses in the panel were replaced as a preventative measure.

As a further corrective action, QMMS 6600-1-S4 will be revised to require replacement p

of all the rubber hoses on the fuel and lube oil lines every refueling outage.

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' WORK REQUEST NO.:

Q78934 LER NUMBER:

89-005 COMPONENT: System 5600 - The Electrical Maintenance Department (EMD) was in E'

the process of removing the limit switch on the #2 Main Stop Valve (MSV) under WR Q78934. 'The limit switch was being replaced as a result of problems identified.

during a surveillance test.

EMD personnel removed four electrical. leads on the limit switch that were identified in the work package.

Two leads were still g

attached that vare not mentioned in the work package.

When the limit switch was removed, the number 1, 3 and 4 MSV's began to close.

When the three MSV's reached 10% closure, a reactor scram occurred due to Turbine Stop Valve Closure.

A normal scram recovery then began. Since the $2 MSV is the master valve and the 1, 3 and 4 MSV's are the slave valves, when the limit switch was removed, a "CLOSE" signal was given to the other three MSV's and they began to close, causing the scram.

CAUSE OF MALFUNCTION: The cause of the event was attributed to personnel error.

The EMD Work Analyst inadvertantlj overlooked the JF5 and JF6 connections on the limit switch an the-preparation of the NWR instructions.

The Work Analysts' over-sight was a result of inattention to detail in reviewing the wiring diagram for the limit switch.

Contributing to this event was the inadequacy of the field verification of the wiring diagram performed by the EMD workers.

Poor lighting..

high temperature and radiation concerns contributed to the workers inadequate field verification.

RESULTS & EFFECTS ON SAFE OPERATION: The safety significance of this event is minimal. All expected ESF actuations occurred to bring the reactor to a safe shutdown condition.

If the Turbine Stop Valve Scram had failed, the reactor scram would still have occurred from an Average Power Range Monitor (APRM) High-Neutron Flux Scram.

ACTION TAKEN TO PREVENT REPETITION:

Immediate corrective actions included counseling the EMD Work Analyst.

Subsequent corrective actions will include training on this event, revisions to QAP 1500-2, guidance provided for contacting Instrument Maintenance Department (IMD) when EHC systems are involved, revisions-to the training lesson plans, perforning a wiring field verification on main turbine equipuent, and an Operating Memo requiring power reduction when working on. components af fecting EHC circuitry, e

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l' WORK REQUEST NO. - Q78966, Q79039, Q79066, Q79071 i

LER NUMBER: N/A

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L COMPONENT Eystem 7500 - Unit Two was in RUN mode at 96% of rated core thermal power. Unit One was in REFUEL mode at zero percent power. While the Operating L

Department was performing QOS 7500-5.. Standby Gas. Treatment System (SBGT) Monthly j-Operability Test, the B SBGT filter train was started. The flow indicator located on control room panel 912-5 was found to be reading greater than 5,000 cubic feet per minute (CFM). This exceeded the Technical Specification 3.7.B.l requirement l

of 4,000 CFM + 10%. The B SBGT train was declared inoperable and the A SBGT train L

was started in accordance with Technical Specification. Mechanical Maintenance Department (MMD) investigated the-problem with valve 1/2-7510B ander WR Q78966

- and had apparently fixed the problem.

During the operability test, the B SBGT F

train still did not perform as expected. More-maintenance was done under WR Q79039 L

. Department.(IM) to investigate and repair. A new actuator was installed and the and again the test failed. WR Q79066 was thus written for the Instrument _ Maintenance operability test was successfully completed. The'B SBCT train was then returned

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to service.

I' CAUSE OF MALFUNCTION: The high flow was due to_the improper functioning of the flow control valve,l/2-7510B. The removed positioner was disassembled and-inspected-(

. under'WR Q79071 and no obvious problems were found. Further investigation showed that the valve actuator and stem may have been sticking, RESULTS & EFFECTS ON SAFE OPERATION: The safety significance of the-event was p

minimal because the A SBGT was ;toven operable in accordance with Technical Specifications until the B train was successfully tested and returned to service.

g In addition, the'B train would have been able to provide some filtering in'needed.

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and any resultant release would have been fully monitored.

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ACTION TAKEN TO PREVENT REPETITION:

The valve actuator was replaced after several attempts at lubrication failed to correct the problem. Although previous sur-ve111ance-tests indicated acceptable performance, there were indications that the valve actuator and stem may have been sticking.

Because of this, both flow control valves on both SBGT trains will be replaced when the parts arrive.

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UNIT 2 MAINTENANCE

SUMMARY

WORK REQUEST NO.: Q74656 i

L LER NUMBER: N/A COMPONENT:

System 590 - With Unit 2 at 91% of rated core. thermal power,'a half f

scram was received due to a Local Power. Range Monitor (LPRM)' spiking high. LPIOi - 49D was placed in BYPASS and Work Request Q74646 was initiated.

i CAUSE OF MALFUNCTION:

The cause of the event was most likely a problem with l

the LPRM's connector. The LPRM had an Amphenol connector, which has been found.

to be susceptible to moisture and mechanical problems.

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RESULTS & EFFECTS ON SAFE OPERATION: The safety significance of RPS Channel A tripping due to a LPRM Hi-Hi signal to Average Power Range Monitor (APRM) channel 3 was in the safe condition. A spurious signal causing one channel of RPS to i

trip is not significant due to the two channel design of RPS.

If Channel A J

of RPS would have tripped while Channel B was tripped, the reactor would have scrammed and gone_to a safe chutdown condition.

ACTION TAKEN TO PREVENT REPETITION: The corrective action was to continue the f'

. process of replacing all the Amphenol connectors with Lemo connectors.

Lemo connectors are now required on all LPRM's.

All connectors for the station will be of the Lemo type by 1993.

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LICENSEE EVENT REPORTS s

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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 1 occurrence reporting requirements as set forth.in sections 6.6.B.1.:and 6.6.B.2.

1 ofLthe Technical' Specifications, s

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c Licensee Event' F

Report Number Date Title of Occurrence 89-019:

11-5-89 Reactor Level Inst.' Failed-

"j to Reset Properly causing a Group II and III and; Cont. Room HVAC Isolation p.

2 89-020 11-17-89 Reactor Full Scram While-Transferring the 24/48VE g

Batteries to Normal Line'Up 89-021L 11-24-89 Missed Semi-Ahnual Tech.:

Spec. Surveillance on HPCI-Fire Protection Deluge System:

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89-072 11-28-89 HPCI Inoperable'Due to Deluge System Actuation t

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UNIT 2 i;.

There were no Licensee Event Reports for Unit 2 for this reporting period.

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

A.

Opetating Data Report B.

Average Daily Unit Power Level C.

Unit Shutdowns and Power Reductions t

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APPENOlX C 3

OPERATING DATA REPORT DOCKET NO.

50-254 UNIT One i

DAfg December 4, 1989 COldPLETt0 OY Lynne Deelsnyder l

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TELEPHONg 309-654-2241 l

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EAM 0000 110189 720

?400 113089 OR0es nounsIN RgPORTesee PgRe00:

1. REPOImeseFOR800:

769 2511 Man. DePeses. CAPACITY M:

3. SWARONTLY Aufte0Regge P0 user Levat Assuer 0W1081 SLSCTW4 CAL RAflese Regu>sese 789 770

3. POUIOR LSWSL 70 EUM80N MNTR4CTSO tip ANY1 Lesuursesel:

& REASONBPORResTheCTeges ter A88Yh "C" heater out-of-service j

T Het as00 m l YW70 OATO CuesukAfive

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110.2 5877.6 123419,8_-

S. M OP MOURS ASACTOR WIAS CRmCAL e..............

0.0 0.0 3421.9-

8. REACTOR RWORvt SMUT 90gne M0ufW...................

59.3 5714.5 114373.7

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f.tetM100SN8RATORetfulet..........................

0.0 0.0 909.2

8. WMT A000RVE SMUT 90 fun MOURS...................... 40718

12350296, 254040375 S. OR000 THERMAL SNERGY 08MERAT80 lefuuMI 3964007 82321620

10. GROW SLSCTR6 CAL SNOROY OSNER ATSO lessuMI............. I1185 5631 376873E 77334012

11. ht? ELECTR6 CAL ENGA0Y GENERATSO IMufNi 15.3 73.3 70.R

13. r 3 ACTOR $8RVICS P ACTON...........................

15.3 73.3 82.0_

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

8.2 71.3 77.2

14. UMT 68RVICE P ACT081...........................

b*2 7I 3 77 8

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

1.0 61.1 65.0 '

16. UNIT CAPACITY P ACTOR iusing tAOCl.....................

1.0 59.6 63.4

17. W88tT CAPACITY P ACTOR iveme Demen tsvust.................

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0.0 6.4 s.4

18. UNIT PORC80 0WTASS Raft..........................

' 19. SMUT 90 West SCM000L80 CVER NExT 4 MONTMS (TYPE OATE. ANO OWRATION OF EACMh

20. IP BMWT DOWN AT tre0 0F REPORT PERIOO. ESTIMATED DATE OP STARTUP:

31. UNITS IN TEST ST ATUS IPRIOR TO COMMERCI AL OPERATION):

PORECAST ACMitV80 INITIAL CRITICAUTY INITIAL ELSCT9tICITY COMMERCIAL OPERA 79010 i.m

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APPEN0lX C OPERATING DATA REPORT DOCKET NO.

50-265

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UNIT Tm DATg December 4, 1989 COMPL TtoSY Lynne Deelsnyder l

ygtgpwopgg 309-654-2241 WMM STM 0000 110189 720 2400 113084 GRom MOUR$ lN Rp0RTies0 Pene00:

t. Ap0RTINSPORe00:

I 769

3. OURIh6NTLY AUTMemet00 Possem LSV8L temam, 2S 11 MAN. Otfeme.CAPActTV emerseeg; owe 0N SLSCTRsCAL MAftest issut>seest:

789

3. POUUOR LSV8L TO UIMcCM RSTRsCT80 tif AsuYi tessus.semes:

N/A

& RSAMIIS POR RSBThecT9000 86F AIIYl 1

THet as08fTM VR70 SATE Cut 0ULAftve 720,0 7690.7 118640.6 S. IsubseWI 0F MOURS RSACTOR tuAS CRITICAL,..............

0.0 2385.8__

8. REACT 0ft R000RVE SMUTOCURu MOURB................... 720.0 7622.8 115354.5 T. MOURO GSNERATOR ON Ulst..........................

0.0 0.0 702.9 S. UMT RE00Rv8 SMUT 00 gun MOURS......................

9. Omegg TNGRAAAL ENERGY OSN8AAft0 (MusMI............. 1675337 16858486 247768759 550535 5473532 79407053

10. OR000 SLOCTRICAL ENSROY OeNERAT80 tessuwt.............

526926 5234329 74970906

11. NET WLOCTRICAL ENOROY GENERATED (IssuMI..............

100.0 95.9 77.5 L

11. e TACTOR SERVICS P ACT0ft...........................

100.0 95.9 79.5

13. REACTOR AV A3LAtiLITY P ACTOR.......................

100.0 95.1 75.4

14. UM T SSRVICE P ACTOR..............................

100.0 95.1 75.8

15. UNIT AVAILAttuTY P ACTOR..........................

95.2 84.9 63.7-to. UNIT CAPACITY P ACTOR iussis te0Cl.....................

j 92.8 82.8 62.1 l

17. UNIT CAPACITY P ACT041Uisie Demen WWeel.............

0.0 4.2 8.2

18. UMT PORC80 0UTAOF. RATE..........................

t

19. SMUT 00 gums $CM00UL80 0vtR NtxT 4 MONTMS (TYPE. DATE. ANO DURATION OP EACND:

i

(-

20. IP SMUT 00 WIN AT END OF REPORT PERICO. ESTIMATED DATE OF STARTUP:

j

21. UMTS IN TEST STATUS (PRIOR TO COMMERCIAL OPERATIONl:

POR$ CAST ACMifV80 INITIAL CRITICALITY IMTIAL ELOCTRICITY C00mstRCIAL OPERAT1000 1.1M l.

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4 RPPD S!! B MARGE MILY LMIT POER LINEL Dochet No.56-254 Unit One Date December 4,1989 r

Completed By Lynne Deelsnyder Telephone 309-654-2241 MONTH ISOER MY MRAGE MILY PUKR LEVEL MY AVERAGE MILY POER LEVEL Otir4rt)

(fedeWet) l 1

-7 17

-7 2

-7 18

-7 3

-7 19

-7 4

-7 20 4

5 4

21 4

6

-7 22

-8 7

-8 23 4

8

-7 24 4

9

-7 25 4

it

-7 26 4

11

-8 27

-8 12

-7 28 53 13

-7 29 127 14

-8 30 255 15

-7 16

-7 9

INSTRlCTIONS 1

. On this form, list the average daily unit power level in Mie-Net for each day in the reporting month.

Compute to the nearest whole megawatt.

1 These figures will be used to plot a graph for each tiporting month. Note that when maximus dependable capacity is used for the vet 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 sich cases, the average daily unit power output sheet should be footno6ed to explain the apparent anomaly.

1.16-8

'l-j i

l t

i APPD GIX B MMBE MILY (MIT POER LEVEL i

Docket No.56-265 i

Unit Two j

Date December 4, 1909 Completed By Lynne Deelsnyder Telephone 309-654-2241 i

NINTH ICVDGER MY AVERAGE MILY POER LEVEL MY MilRGE MILY POER LEVEL (IthHiet1 Ulle-Net) 1 783 17 768 2

736 18 752 3

747 19 740 4

719 20 728 5

519 21 751 6

741 22 741 7

744 23 715 8

753 24 699 9

771 25 794 le 747 26 714 723 27 719

. 12 717 28 746 13 738 29 743 14 741 30 742 15 774 16 750 '

i 1

l INSTRUCTIONS l,

On this form, list the average daily unit power level in leir-Net for each day in the reporting month.

Compute to the nearest whole negawatt.

These figures will be used to plot a graph for each reporting month. Note that uhen marious dependable capacity is used for the not electrical rating of the unit, there may be occasions when the daily average power level exceeds the 100$ line (or the restricted power level line). In such cases, the average daily i.

' unit power output sheet should be footnoted to explain the apparent anomaly.

l:

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V1.

UNIQUE REPORTING KEQUIREMENTS I

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

A.-

Main Steam Relief Velve Operations V

Relief valve operations during the reporting period are summarized in the following table. The table includes'information as to which relief valve van actuated, how it was actuated and the circumstances resulting in its actuation.

. Unit: One

_Da t e : - November 26, 1980 Valves Actuated No. & Type of Actuation 1-203-3A 1 Manual 4

1-203-3B 1 Manual 1-203-3C 1 Manual 1-203-3D 1 Manual i

1-203-3E 1 Manual Plant Conditions:

Reactor Pressure - 920 Description of Events:

Semi-Annual, Manual Operation of Electromatic Relief Valves (QOS 201-S1): Tech Spec: Ref. 4.5.D.1.a j

Unit: One Date: November 28, 1989 Valves Actuated No. & Type of Actuation

.1-203-3A 1 Manual 1-203-3B 1 Manual 1-203-3C 2 Manual 1-203-3D 1 Manual 1-203-3E 1 Manual Plant Conditions:

Reactor Pressure - 920 Description of Events: HPCI Subsystem Outage Report (QOS 2300-01); Manual Operation of Electromatic Relief Valves (QOS 201-S1);

Tech Spec:

Ref. 3.5.C.2/4.5.C.2 1-203-3C was actuated twice because its tailpipe temperature did not decrease following valve actuation.

After second operation, tailpipe temperature returned to normal.

B.

Control Rod Drive Scram Timing Data for Units One and Two The basis for reporting this data to the Nuclear Regulatory Commission are specified in the surveillance requirements of Technical Specifications 4.3.C.1 and 4.3.C.2.

The following table is a complete summary of Units One and Two Control Rod Drive Scram Timing for the reporting period. All scram timing was performed with reactor pressure greater than 800 PSIC.

= - - -.

S*

RESULTS OF SCRAM TI: LNG MEASUREMENTS

..m PEV0ftED ON UNIT 1&2 CONTROL R0D DRIVES, FROM l-1-89 TO 12-31-89 AVERAGE TIME IN SEC0les AT %

MAX. TIIE INSERTED FROM FULLY WIT}ORAWN FOR 90%-

INSERTION DESCRIPTION NUIEER 5

20 50 90 Technical Specification 3.3.C.1 &

DATE OF RODS 0.375 0.900 2.00 3.5 7 sec.

3.3.C.2 (Average Scram Insertion Time) 2-4-89 88 0.30 0.67 1.43 2.49 F-9 Unit 2, Hot Scram Timing, Sequence B (2.91) 3-4-89 1

0.28 0.69 1.54 2.68 E-8 Unit 1, Hot Scram Timing, Accumulator Work (2.68) 7-7-89 89 0.28 0.65 1.38 2.47 H-8 Unit 1, Hot Scram Timing, Sequence A (2.75) 8-2-89 1

0.23 0.59 1.32 2.39 N-10 Unit 1, Hot Scram Timing, Check Valve (2.39) 115 and 126 and 127 Work 8-3-89 1

0.26 0.62 1.39 2.43 K-3 Unit 1, Hot scram Timing, Check Valve (2.43) 115 and 126 and 127 Work 8-5-89 1

0.28 0.64 1.39 2.42 P-7 Unit 2, Hot Scram Timing, WR Q75187 to (2.42)

Check 115 Valve and Replace 126 and 127 diaphram. Also replaced 126 valve seat.

8-9-89 1

0.30 0.67 1.46 2.58 H-7 Unit 2, Hot Scram Timing, Check Valve (2.58) 114 and 115 Work and Diaphram and Seat Replacement of 126 and 127 Work.

3-11-89 1

0.29 0.66 1.42 2.48 J-19 Unit 2, Hot Scram Timing, Check Valve (2.48) 114 and 115 Work and Diaphram and Set Replacement of 126 and 127 Work.

9-9-89 89 0.30 0.66 1.41 2.47 F-6 Unit 2, Sequence A, Hot Scram Timing (2.76)

Il-29-89 177 0.29 0.67 1.44 2.54 G-9 Unit 1, Sequences A and B, Start-Up Hot (3.27)

Scram Timing 0027H/0061Z5

1 4

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.

k 0027H/0061Z

- 1

-1 Q

QTP 300-S32 En Revision 2 I

QUAD CITIES REFUELING-October 1989 INFORMATION REQUEST

- 1.

Unit:

01 Reload:

10 Cycle:

11 2;

Scheduled date for next refueling shutdown:

10-6-90 3.

Scheduled date for restart following refueling:

12-11-90 4.

Will refueling or resumption of operation thereafter require a Technical Specification change or other license amendment:

NOT AR YET DLTERMINED.

U 5.

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

JULY 6, 1990 6.

Important licensing considerations associated with refueling, e.g., new i

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:

1537 i

8.

The prresent licensed spent fuel pool storage capacity and the size of any trc* ease in licensed storage capacity that has been requested or is t~

planned in nu:nber of fuel assemblies:

1 L

I-Licensed storage capacity for spent fuel:

3657 a.

b.

Planned increase in licensed storage:

0 9.

The projected date of the last refueling that can be discharged to the I

spent fuel pool assuming the present Ilcensed capacity:

2008 1,.

a APPROVED l'

e

]

(final)

W30W 14/0395t Q.C.O.S.R.

F_

f:

,f-QTP 300-532 Revision 2 QUAD CITIES REFUELING-October 1989 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:

5-5-90 4.

Will refueling or resumption of operation thereafter require a Technical Specification change or-other license amendment:

NOT AS YET DETERMINED.

5.

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

NOV.'.MBER 2,.1990 L

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:

1843 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.

Lict'ned 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 fuei pool assuming the present licensed capacity: 2008 APPROVED (finai) 00T 3 01969 14/0395t-O.C.O.S.R.

\\

1 2

Kg, )e,

. w..

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

Atmospheric Containment Atmospheric Dilution / Containment ACAD/ CAM

=

Atmospheric Monitoring-

-ANSI American National Standards Institute Average Power Range Monitor APRM ATHS Anticipated Transient Without Scram Bolling Water Reactor BWR CRD Control Rod Drive Electro-Hydraulic Control System EHC Emergency Operations Facility EOF GSEP Generating Stations Emergency Plan r

High-Efficiency Particulate Filter HEPA HPCI-High Pressure Coolant Injection System High Radiation Sampling System HRSS IPCLRT-Integrated Primary Containment Leak Rate Test IRM

. Intermediate Range Monitor Inservice Inspection ISI LER Licensee Event Report LLRT Local Leak Rate Test LPCI-Low Pressure Coolant Injection Mode of RHRS Local Power Range Monitor L

.LPRM l

MAPLHGR.

Maximum Average Planar Linear Heat Generation Rate l

MCPR' Minimum Critical Power Ratio MFLCPR Maximum Fracticn Limiting Critical Power Ratio MPC Maximum Permissible Concentration l

MSIV Main Steam Isolation Valve NIOSH National Institute for Occupational Safety and Health PCI Primary Containment Isolation PCIOMR Preconditioning Interim Operating Management Recommendations RBCCW Reacter >;ilding Closed Cooling Water System

'RBM:

Rod Block Monitor RCIC Reactor Core Isolation Cooling System RHRS Residual Heat Removal System rotection System RPS Reactor o Rod -Horti. clinimizer RWM SBGTS Standby Gas Treatment System Standby Liquid Control SBLC SDC Shutdown Cooling Mode of RHRS SDV Scram Discharge Volume SRM Source. Range Monitor Turbine Building Closed Cooling Water System TBCCH.

TIP Traversing Incore Probe TSC Technical Support Center 0027H/00 W.

c

-