ML19325C287

Monthly Operating Repts for Sept 1989 for Quad-Cities Nuclear Power Station,Units 1 & 2.W/891002 Ltr

ML19325C287
Person / Time
Site:	Quad Cities
Issue date:	09/30/1989
From:	Deelsynder L, Robey R COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
RAR-89-68, NUDOCS 8910160021
Download: ML19325C287 (25)

Text

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[7,9~;K^. . Qqad Cateos Nuclear Powur Station -

'22710 206 Avenue North jy;' O. . Cordova, lil6ncis 61242-9740 :.

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, Director of. Nuclear Reactor-Regulations.

U.' S. Nuclear Regulatory Commission Mail-Station Pl-137 Washington, 9. C. 20555 x

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1 Enclosed forfyour information is th'e Monthly Performance Report W . covering the operation of Quad-Cities Nuclear Power Station, Units .

One andLTwo.-duringlthe month;of' September, 1989. 3 sRespectfully, a t

' COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR' POWER STATION.

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. QUAD-CITIES NUCLEAR POWER STATION g:

UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT 6:

SEPTEMBER, 1989

'COMMONNEALTH EDISON COMPANY I

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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TABLE OF CONTENTS.

p h I. Introduction.

II. ' Summary of Operating Experience lb '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 Haintenance of Safety Related Equipment

IV. -Licensee-Event Reports V . .. Data Tabulations k A. Operating' Data Report B. Average Daily Unit. Power Level l- C. ~ Unit. Shutdowns.'and Power Reductions

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[ VI. Unique Reporting Requirements- i

.A. Main Steam Relief Valve Operations

.B. Control Rod Drive Scram Timing Data ,;

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,VII.. Refueling Information ,

..VIII . Glossary

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! I. INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Boiling Water ,

Reactors, each with a Maximum Dependable Capacity of 769 MWe Net, located in Cordova, Illinois. The Station is jointly owned by Commonwealth Edison

-Company and Iowa-Illinois Gas & Electric Company. The Nuclear Steam Supply ,

Systems are General Electric Company Bolling Water Reactors. The Architect / Engineer was Sargent & Lundy, Incorporated, and the primary.

construction contractor was United Engineers & Constructors. The Mississippi River is the condenser cooling water source. The plant is subject to license numbers OPR-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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SUMMARY

-OF OPERATING EXPERIENCE A."! Unit One

% , Unit One-began the month of September holding load at maximum attainable power while routine.surveillances were performed. Normal operational-activities occurred through September 9.

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_On_ September 9,' power'reduetions began in preparation for normal unit shut-down. On September 10, at 0058 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br />, the generator was taken off-line and at 0106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br />, the reactor was manually scrammed'for the purpose of h , -beginning the End of Cycle = Ten Refueling Outage for Unit One. At 0128 hours0.00148 days <br />0.0356 hours <br />2.116402e-4 weeks <br />4.8704e-5 months <br />, the reactor mode switch was placed to SHUTDOWN. At 0608 hours0.00704 days <br />0.169 hours <br />0.00101 weeks <br />2.31344e-4 months <br />, the reactor head vents were opened and normal unit shutdown procedures were completed, n

For-the remainder of the month, normal refueling activities occurred. On September 12 andi13, the reactor head and steam dryer were removed. On September 13, at 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, the mode switch was locked in REFUEL. On September 14, at 1530 hours0.0177 days <br />0.425 hours <br />0.00253 weeks <br />5.82165e-4 months <br />, core unloading was begun and was completed

..nt 1805 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.868025e-4 months <br /> on September 17.

On September 21, at 1410 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.36505e-4 months <br />, during the transfer of new fuel from the new fuel storage vault to the fuel pool, fuel bundle LYT 191 dropped from

~ the- refueling grapple and can'e to rest on top of the spent fuel racks.

' Operating Engineer, Lead Nuclear Engineer, and Radiation Protection Foreman (among others) responded. After determination of no apparent damage, the i bundle was moved to cell in fuel rack. No further fuel movement was done until an investigation was completed. A modification to the refueling bridge hoist raise circuitry was completed on September 27. Fuel moves were resumed on September 28 (Deviation Report Number 04-01-89-080). Normal refuel activities were performed for the remainder of the month.

B. ' Unit Two Unit.Two began the month of September with unit startup in progress. The reactor.was made critical at 2008 hours0.0232 days <br />0.558 hours <br />0.00332 weeks <br />7.64044e-4 months <br /> on August 31. The generator was synchronized to the grid at 0838 hours0.0097 days <br />0.233 hours <br />0.00139 weeks <br />3.18859e-4 months <br /> on September 1. A load increase 1 to full power was begun using control rods and recirculation pumps. At 1255 hours0.0145 days <br />0.349 hours <br />0.00208 weeks <br />4.775275e-4 months <br />, the unit was held at 416 MWe to perform routine testing and surveillances. On September 2, at 1005 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.824025e-4 months <br />,_ full power was achieved.

1 l On September 3, the reactor recirculation loop A flow indication differed by greater than 5% from the loop flow characteristic curve. Reactor recir-culation pump speed' oscillations were approximately 2% (88-90%). This flow fluctuation was an indication that there could be a problem with the 2A g reactor recirculation loop flow control instrumentation. This instability

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motor / generator set _ scoop' tube was locked up. -The M/G set-scoop tube wasJ

,taken.out;of the' locked-position per tha demands of the Load' Dispatcher-until repairs couldibe made. On September 8, at 2315 hours0.0268 days <br />0.643 hours <br />0.00383 weeks <br />8.808575e-4 months <br />, a power reduction.was taken'to perform routine control rod scram' timing. At

, 0530 hours0.00613 days <br />0.147 hours <br />8.763227e-4 weeks <br />2.01665e-4 months <br />, on September'9, the testing was successfully completed.

At-9 0820 hours0.00949 days <br />0.228 hours <br />0.00136 weeks <br />3.1201e-4 months <br />. a power increase to full load-was achieved.

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Upon investigation, it was determined that the 2A recirculation pump speed-

-oscillations were occurring due to a faulty' tach generator. Power levels were reduced,'the 2A recirculation pump'was taken'out of. service, and a new tach generator was installed into the system.- A test was successfully.

performed on the system to verify that reactor recirculation pump; flow and i motor. generator set speed did not oscillate. .The work was completed on

. September 15.

On September =15, at 0133 hours0.00154 days <br />0.0369 hours <br />2.199074e-4 weeks <br />5.06065e-5 months <br />, power levels were adjusted, and the unit-was placed in Economic Generation Control;(EGC). The unit remained in EGC or operated-near: full power until September 18. At.1048 hours0.0121 days <br />0.291 hours <br />0.00173 weeks <br />3.98764e-4 months <br />,_EGC was- .

-; -- tripped and the unit was taken to full power. -l For the remair. der of the month Unit Two operated in Economic Generation Control:or remained near full power. Normal plant operational activities-

-p _and routine surveillances were successfully performed and completed.

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N Bzis  ; III ~. LPLkNT OR PROCEDURE CHANGES TESTS EXPERIMENTS,-AND SAFETY'

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.y A.:l-Amendments to-Facility License or Technical Specifications

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. Specifications for-the reporting ~ period..

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' B '. - Facility or Procedure Changes Requiring NRC-Approval hc ,

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p- ,' C... Tests and Experiments Requiring NRC Approval

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M There were no tests'or Experiments' requiring 1NRC approval j for the' reporting period. ~ 1 R ID. Corrective' Maintenance of Safety Related Equipment I

?The following' represents a tabular summary of the major safety '

related maintenance. performed.on Units One and Two during the o reporting period.:. -This summary includes the following: Work  !

1 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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SUMMARY

A 30R_K REQUEST NO. Q69319, Q69321 a" .

. FR NUMBER: 80-013 COMPONENT: System 5700 - While performing procedure QOS 005-S14, Unit Two Equip-

' ment Attendants Operator's Surveillance / Turnover, the Equipment Attendant (EA) founi the Unit One A Core Spray room coolcc to be off and was not able to start it manually. The EA notified control roos personnel that the room cooler was inoperable. The 1A Core Spray systems and the Unit One kenctor Core Isolation C9oling (RCIC) systems were declared inoperable. Upon inspection of the cooler, the EA'found that one belt of the two had been broken, and the other one was off the pulley. The EA placed the loose belt back on the pulley and the cooler operated satisfactorily. The NRC was notified because the RCIC system had been declared troperable.

CAUSE OF MALFUNCTION: The cause of the event was an insufficient preventative maintenance program. The failure of the room cooler belt was due to normal end of life. The room cooler had been inspected by maintenance personnel during the Fall 1987 Unit One Refuel. Outage as part of procedure QMPM 5700-1 Emergency Air llandling Unit Maintena' ice and Inspection. The belts were visually inspected, determined to be acceptable and not replaced. If the belts had been replaced during the outage inspection, it would have prevented failure of the belts during operations.

RESULTS & EFFECTS ON SAFE 0PERATION: The safety of the plant and operating personnel was not affected dut-ing this event. The Unit One High Pressure Coolant injection (HPCI) and the Low Pressure Coolant Injection (LPCI) mode

.of Residual-Heat Removal (RHR) were successfully' tested after finding the 1A Core Spray room cooler inoperable and prior to removing it from service for repairs. Therefore, all other Emergency Core Cooling (ECC) systems including the B loop of Core Spray were operable throughout the event.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action consisted of re-installing the loose fan belt and operating the room cooler temporarily en one fan belt. Work requests were written to replace the belts for the balance of the :com coolers during the next available system outage. Work Requests Q693' and Q69321 were written to replace the fan belts on the Unit One

.HPCI and Unit One "A" RHR room coolers.

u - To prevent re,currence of this event, the refuel outage maintenance surveillance, QMPM 5700-1, will be revised to require replacement of all ECCS room cooler fan belts each refuel outage.

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WORK REQUEST No.: Q76866, Q76906 l

LER NUMBERt N/A.

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r COMPONENT , System 202 - The IB recirculation pump was being taken out of service  ;

to repair the mechanical seal. Upon attempting to close the IB recirculation <

a suction valve 1-202-4B. the Nuclear Station Operator (NS0) noticed that the-valve had stopped An the mid-travel position. The NSO then tried to open'the valve and it would_not fully open. The NSO tried to stroke the valve again but obtained the same results. The valve was taken out of service and Work Request Q76866 was written to remove the motor and check to the pinion gear. ,

The motor was reinstalled but failed to operate when tested. Work Request #

Q76906 was_ written to investigate the motor. A replacement motor was then f - installed on the motor operator.

• CAUSE OF MALFUNCTIONt- The apparent cause of the motor operator, MO 1-202-4B.

was a failure of the magnesium shorting ring. The shorting ring had separated from the rotor core and caused a decrease in rotor field. For this reason the p

motor did not have enough-torque to move tha valve. The separation of the shorting ring was attributed to Intergranular Stress Corrosion Cracking due to environmental degradation of the magnesium,

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RESULTS'& EFFECTS ON SAFE OPERATION: The safety of the public and plant personnel-t was not.affected by this event. The unit was in cold shutdown, and the 1A recircu-

[ lation pump was fully operational to provide proper circulation cooling during cold shutdown conditions. If the event had been at power, the IB recirculation puro could not have been isolated if a catastrophic failure of the mechanical sei. would have occurred. This would have created excessive amounts of water

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in the drywelli however, a double-ended shear of the recirculation line, which is far worse than the seal failure, has already been analyzed.

ACTION TAKEN TO PREVENT REPETITION: Electricol Maintenance (EM) personnel determined the motor to be the problem and replaced it. In addition, a list t

of all magnesium operators was compiled. Inspection / test methods were in progress

- to detect present and future rotor degradation.

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UNIT 2 MAINTENANCE

SUMMARY

WORK REQUEST NO.t. Q66584 LER NUMBER .N/A L '

COMPONENT! System 1200 - While investigating a samplo expansion for the on-going.In-Service Inspections (IS1), it was discovered that-a weld in penetration  ?

-X-14 on line number 2-1202-6"A was not previously identified in the ISI program.

Since this condition existed for a triple flued head penetration, a repair of  ;

Unit One and Unit Two triple. flued head penetrations was conducted.

CAUSE OF MALFUNCTION: The likely cause of the event was a management deficiency r that did not require complete and accurate documentation of actual conditions i

.during the construction'of the plant. Current programs and requirements are such that this type of condition would be adequately documented and controlled.

l L RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plant and its personnel were not affected by this event.

i ACTION TAKEN TO PREVENT REPETITION: The weld on line 2-1202-6"A was considered ,

, inaccessible. As a result, a revision to ISI relief request CR-4 was completed to include the additional inaccessible weld in penetration X-14. Work Request ',

Q66584 was written to conduct a weld overlay on line 2-1202-6"A near the flued head anchor.

WORK REQUEST NO.t Q76658 LER NUMBER: N/A COMPONENTt.. System 1700 - While the unit was operating at 100 percent of rated l

[ core thermal power, a half scram was received on the 902-5 panel, annunciator D-15 " CHANNEL B REACTOR SCRAM", and a half " GROUP 1..ISOL. Cil. TRIP" was 3

. received on the 902-5 panel, annunciator B-7. In addition, a Main Steam Line (MSL)' Channel "B" high radiation alarm annunciated. Other plant conditions i

' appeared normal. The operators immediately checked the MSL radiation anonitor, t

.L 2-1705-2B, located in the 902-2. panel. A failed input / output (10) board was found as a result of a self-diagnostic test of the monitor. Work Request Q76658 1 was. written to repair the radiation monitor. The monitor was replaced with i a tested spare, and the half scram and half Group I isolation were reset.

'CAUSE OF MALFUNCTIONt The cause of the event was random oquipment failure. [

'A failed K3 relay in the I/O board caused the upscale contact to fail open, giving a false high radiation signal.

! RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal since the Reactor Protection System and Primary Containment Isolation Systems were fully operable during the entire event.

ACTION TAKEN TO FREVENT REPETITION: Work Request Q76658 was written to repair the "B" MSL radiation monitor. A tested spare was installed to allow the system to return to service. The original monitor was repaired and returned to the 902-2 panel. Station records end NPRDS data show that the K3 relay failure was an isolated incident.

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f WORK REQUEST NO. Q77045 ,

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. COMPONENT System 6600 - The Unit Two Diesel Generator (DG) had-been started

-and loaded to Bus 24-1 for the Diesel Generator Monthly Load Test (QOS 6600-1).

The Control Room then received the " DIESEL GEN TROUBLE" alarm. The Operator ,

in the DG room determined that air was exhausting through the Air Start motors after noticing the local " LOW STARTING AIR PRESS./ COMPRESSOR LOCKED OUT" alarm. .

. The DG was declared inoperable and put to STOP. and Maintenance Work Request Q77045 was written to inspect both the Air Start Solenoid and the Air Start ,

Relay Valve.  !

CAUSE OF MALFUNCTION: The apparent cause of the failure of the DG Air Start System was corrosion in the Air Start Relay Valve. -There were signs of vster ,

E present inside the valve, and the top seat was not seating properly due to corrosion. Water is believed to have entered the system with the air, past  ;

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the cyclone type air dryer incorporated in the air start piping.

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I RESULTS 6 EFFECTS ON SAFE OPERATION: The safety implications of this event vere ninimal as the Unit One Half Diesel Generator and its associated safety ,

systems were fully operable as required by Technical '.ecifications.

ACTION TAKEN TO PREVENT REPETITION: The immediate ..,..cetive action was to investigate the cause of the air leak. When the cause could not be readily determined, the Air Starting System was isolated and the Diesel Generator declared inoperable. Work Request Q77045 was generated to investigate and ,

repair the oroblem, j

-The Air Start Relay Valves are cleaned and inspected every Refuel Outage per Station procedure QMMS 6600-1-S4, Diesel Inspection - Refueling Outage Checklist.

This procedure was revised to include more detail during inspections and part replacement.

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IV. LICENSEE EVENT REPORTS i

The following is a tabular summary of all.11censee event reports.for l 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 f Report Number Date Title of Occurrence I o

89-014 9-10-89 MSIV 2A and 2D j Exceeded Leak Rate j Test Limit 89-015 9-16-89 off Gas isolation Due to Off Gas Radiation [

Monitor Spiking High While  ;

Swapping Power Supply j

'89-016 9-21-89 New }uel Bundle Dropped in Fuel Fool During Vault to Pool Transfer i

, UNIT 2 - 89-004 8-11-89 Inability of ACAD to Perform with Loss of ,

Diesel Generator 5

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[. V. DATA TABULATIONS  :

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The following data tabulatter are presented in this report: .

A. Operating Data Report i V

I B. Average Daily Unit Power Level .

C. Unit Shutdowns and Power Reductions  ;

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' t APPENOlX C

- OPERATING DATA REPORT DOCKET 140. 50-254 UNIT one i OAfg October 6. 1989 l COMPLE780 SY Lynne Deelsnyder ]

I TELEPM0fgg 309-654-2241 l  !

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OPERAftf008?ATUS 0000 090339 093089 720

1. REPORTNet POR80ig 2400- OR0es MOURsIN RePORTiset Pere 00:

_2511 MAm. OsPetee. CAPACITY tesumsdesel. 769

3. SWIE0fuTLY Aufte0Rt380 POusem LevtL IgY 050000 SLSOfRerAL RAfttet Itfuuu.seuill  :

'- . S. PeguGA L8v0L TO WIMe0M RWTRICT80 (IP ANYI tesuus.seselt N/A j

4. RamanNs FOR ReBTRICte04 esp AleYh q THISRIOlffM VR 70 MTE CWisWLAftVS  !

5767.4 123309.6- l

3. suWesegl 0F MOWRB RSACTOR tune CRITICAL . . . . . . . . . . . . . . 217.1 O.0 0.O g4g i l S. REACTOR RGBORVE SMWTDOnse MOWNB . . . . . . . . . . . . . . . . . . .

5655.2 1 H314.4 T. MOURS OOWERATOR ON Litut . . . . . . . . . . . . . . . . . . . . . . . . . 217.909,2 0 0.0 0.0 l

8. Weelf ReggAVs seeWT90sule MOURS . . . . . . . . . . . . . . . . . . . . . .

12309578 253999657 l S. OR005 TMERIAAL SWOROY 08NGRAT80 lesawl ............. 426341 134377 3952822 82310415 f 106 OR006 GLOCTRICAL SNOROY 0848AATBD tessuwt . . . . . . . . . . . . .

123861 3768439 77111711 it. Nff SLOCTR6 CAL GNOROY 08NERAfte insuuMI . . . . . . . . . . . . . .

30.2 88.0 80.5 i

18. ' EACTOR SERVICE P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . . .

0.2 88.0 82.8

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

30.1 86.3 77.0

14. UtslT SSRVICE P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

86.3 7R.s

18. UNif AV AILASILITY P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . 30.1 22'4 -

74.8 65.7 ,

18. UNIT CAPACITY P ACTOR (Weine tt0Cl . . . . . . . . . . . . . . . . . . . . . '

21. 8 72.9 64.0

17. W8847 CAPACITY PACTOR tuene Osmen issuel . . . , . . . . . . . . . . . .

0.0 6.5 s.4

14. UNIT PORC80 0WTAGd Raft . . . . . . . . . . . . . . . . . . . . . . . . . .

19. SteWT00guset SCMSOULIO OvFR Nix 7 4 MONTM8 (TYPt. DAT8. A40 OVRATiON OP EACMH

20. IP SMUT 00 sun AT INO OF REPORT PEA 100. ESTIMATED CATE OP STARTUP:

POR4 CAST ACMifV90

21. UNITS IN TSST STATUS (PRIOR TO COMMERCI AL OPSRATIONh INITIAL CRITICALITY INITIAL ELSCTRICITY C0tateSRCIAL OPORAfl000 1.1H

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APPEN0lX C OPERATING DATA REPORT DOCKET NO. 50-265 UNIT h __

CATE October 6 1989 COMPLETED BY Lynne Deelsnycfer TELEPHONg 309-654-2241 l

[ OptAATING 87ATUS 0000 090189 093089 720 t.flep0RTustPORs00e 2400 SRom MOUR$ IN REPORTitle 70RIOS:

3. SWRRelWTLY AWTte0Re889 P0usem LSVEL lesuey 2511 MAX. 00F0000. CAPACITY elues 8esel, 769 0W8081 SL8CTRICAL RATosse lesuspeessi: /ev N/A 3 POWW LSVOLTO WesetM RWTRICT80 llP ANYl Induuteesth __ _

4. ftSANIS POR RWTRICTISBIIIP AsfYh TNas es08fTN YR TO DATE CutsWLATIVE

4. leWidthi 0F MOURO REACTOR nAS CRITICAL . . . . . . . . . . . . . . 7 20. O g . L 117291.A O.0 0.0 2985.8 l S. REACTOR ROBORv8 SMWTOOWN se0WRB . . . . . . . . . . . . . . . . . . .

711.4 6216.5 113948.2

. pe0WIIS 08N8 RAT 081080 Liteg 0.0 0.0 702.9

8. UMT RWORVE 9MUT90 sued MOURS, . . . . . . . . . . . . . . . . . . . . . .

13633294 244543567

9. OR000 TMERIAAL titetROY 08N8AAT80 leenMI ............. 1621073 4414085 78347556 {

1 0. GR005 GLOCTR6 CAL SNOROY OSNGR ATIO lt85fMl . . . . . . . . . . . . 52 7 798 504786- 4218608 73955185  ;

11. 888T SLOCTR6 CAL INEROY 088stRAftp lefuuMI . . . . . . . . . . . . .

95.7 77.3

12. '5 ACTOR esRv tCl P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . .100. 0 95.7 79.3 l

13. REACTOR AV AILAtiLITY P ACTOR . . . . . . . . . . . . . . . . . . . . . .100. 0

14. UMT SERVICE F Act 0R . . . . . . . . . . . . . . , . . . . . . . . . . . . . . 9 0

• O 90*9 75 2 .

I 90*9 75.6 it. UMT Av AILAtiLITY P ACTOR , . . . . . . . . . . . . . . . . . . . . . . . . 9 0

• O .

83.7 63.4  ;

to. Utelf CAPACITY P ACTOR lusing 080C1 . . . . . . . . . . . . . . . . . . . . 91. 2 81.6 61.8

17. Whlt CAPACITY PACTOR tunne Ossion tonel . . . . . . . . . . . . . . . . 88. 9 __

4.3 8.3 10s WMT PORC80 OUTAGE Raft . . . . . . . . . . . . . . . . . . . . . . . . . .

19. SMuf00nN8 SCee80WL80 CVEM NE AT 4 MONTN8 IfYPS. DA78. AND OWRAfl0N OP 8ACseh

30. IP SMUT 00nN AT $NO OF REPORT Pam:00. EsTIMAtt0 OATE OP STARTUP:

4

21. UMTS IN TEST STAWS IPRIOR TO COMMERCI AL OPERAT10cch POR8CA8T ACHfEV80 .

INITIAL CRITICALITY l IMTIAL 8LSCTRICITY C0tesSRCIAL OPSRAT9001 1.1M ,

.c . . - - - . _ . - - . . - - . . . - - . . - - , , - . - .

.. - - - - . _ - - . . . . ~ _

f

"' '

• APPENDIX S  !

AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-254 UNIT One ,

DATE nearnhar 2- 1989 COMPLETED BY iv=a De>1anvder  !

TELEPHONE 309-654-2241 MONTH scotember. 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe Net) (MWe-Net)  !

t 590 gy -7 ,

3 601 gg -7 3 605 is -8 4 583 30 -9 I 5 390 31 -5 [l 6 _

597 22 -4  ;

y 576 23

-4 e $82 -3 34 9 531 g -4 i 10* * ~3 ~4 33 it -14 27 -4 13 -10 g -4 g3 -10 gg -5 >

14 -9 30 -8 ,

l 13 -10 39 L

16 -7

• l l~

INSTRUCTIONS On this form, list the antage daily unit power lewl in MWe Net for each day in the reporting month. Compute to the penacst whole megawatt.

These figures will be med to plot a graph for cach reporting month. Note that when rmtimum dependable capacityis t uwd for the net electrical ratmg of the unit, there may be occasions when the daily average power level exceeds the 1001 line (or the entneted power Icwl line). In such cases, the average daily unit power output sheet should be footnoted to caplam the apparent anomaly.

1.16 4 es

. s.

APPENDIX B 3 1 - '

AVERAGE DAILY UNIT POWER LEVEL l l-n DOCKET NO. 50-265 UNIT Two

^

DATE October 2, 1989 COMPLETED BY 1,vnne Deelsnyder TELEPHONE 309-654-2241 MONTH September, 1989 .);

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

1 229 17 735 j.

_3 678 gg 742 3

572 g, 733 605 757 4 _ to 5 722 767' gg 743 f

• g 33 723 ,

y 762 729 23 744 718 8 24 690 723 3 g gg - - 615 g- 733 791 733 "

11 27 743 I 13 767 g 681 g, 740 13 14 674 30 700 13 818 33 13 671 INSTRUCTIONS On this form, list the averses daily unit power newl in MWe Net for each day in the reporting month Compute to the nearest whole megawatt.

These figures will be used to plot a graph for cach reporting month. Note that when rmnifrum dependable capacityis

- uwd tcr the net electrical rating of the unit, there may be occasions when the daily average power level exceeds the 100'A line (or the restricted power lewi line), in such cases, the average daily unit power output sheet should be footnoted to capiam the apparent anomaly.

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, , t p VI. UNIOUE REPORTING RE0VIREMENTS I

q The following items are included in this report based on prior commitments to the co.amission:

5 a

A. MAIN STEAM KELIEF VALVE OPERATIONS Relief valve operations during the reporting period are summarized in L. 'the.following table. The. table includes information as to which relief valve was actuated, how it was actuated, and the circumstances resulting l in its actuation.

[-

p ,

Unit One ir Date: September 9, 1989

-Valves Actuated No. & Type of Actuation 203-3A- 1 Manual 1-203-3B- 1 Manual 1-203-3C 1 Manual 1-203-3D. 1 Manual 1-203-3E 1 Manuc1 Plant Conditions: Reactor Pressure - 933 i: ,

Semi-Annual, Manual Operation of Electromatic-Description of Events:

Relief Valves (QOS 201-SI)

' B. CONTROL ROD DRIVE SCRAM TIMING DATA FOR UNITS ONE AND TWO The basis'for reporting this.date to the Nuclear Regulatory Commission are specified in the surveillance requirements of Technical Specifications 4.3.C.1 and 4.3.C.2.

t,i 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 grunter than 800 PSIG.

1.

0027H/0061Z

, 3,

HESULTS OF SCRAM TlWING EASUREENTS -

PEfF M M WIT 1&2 CMTML R00 DRIVES, ."R(Rf ' 1-1-89 TO 12 89 '* '

AVERAGE TIE IN SEC0pWS AT % MAX. TIE -

INSERTED FR(DI FULLY WITERAWN FOR 9(4 INSERTION DESCRIPTim MtABER - 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 Ur.it 1, Hot rcram Timing, Accumulator (2.68) Work 7-7-89 89 0.28 0.65 1.38 2.47 H-8 Unit 1. Hot Scram Timing, Sequenes A

-(2.75) 8-2-89 1 0.23 0.59 1.32 2.39 N-10 Unit 1, Hot Scram Timing, Chech 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 M-7 Unit 2, Hot Scram Timing, Check Valve (2.58) 114 and 115 *.;ork and Diaphram and Seat Replacement of 126 and 127 Work.

8-11-89 .

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 2 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) 0027H/0061Z

4-4 .

t' h

l' s s I

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

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

l l

1 l

l l

0027H/00612

4 QTP 300-S32 j

., Revision 1  !

QUAD-CITIES REFUELING March 1978 IKTORMATION REQUEST

1. Unit: 01 Reload: 9 _ Cycle: 10 l

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

3 Scheduled date for restart following refueling: 11-18-89 9

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

NOT AS ifET DETERMINED.

i 5 Scheduled date(s) for submitting proposed licensing action and supporting Information  ;

JUNE 10, 1989

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

'different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures NONE AT PRESENT TIME.

7 The number of fuel assemblies, a, . Number of assemblies in core: 0 j b. Number of assemblies in spent fuel pool: 2260 l 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:  ;

L a. Licensed storage capacity for spent fuel: 3657

b. Planned increase in licensed storage: 0 9 The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2008 AL' P P R () Nr E E)

APR 2 01978 Q.C.O.S.R.

1

. ~; _.

~

~ '

. .. ' " i QTP 300-532 l Revision 1 QUAD-CITIES REFUELING March 1978 INFORMATION REQUEST

1. Unit: 02 Reload 9 Cycle: 10

2. Scheduled date for next refueling shutdown: 2-3-90 1 3 Scheduled date for restart following refueling: 5-5-90

4. Will refueling or resumption of operation thereaf ter 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:

NOVEMBER 2, 1990

6. Important Ilcensing 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 PRESFNT TIME. 5 i

L 7 The number of fuel assemblies,

a. Number of astemblies in core: 724 l

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. Licensed storage capacity for spent fuel: 3897

b. Planned increase in licensed storage: 0 l 1 9 The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2008 '

l 4 P P R.O V E D APR 2 01978 Q.C.O.S.R.

1

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[?

VIII. GLOSSARY  ;

i The following abbreviations which may have been used in the Monthly Report,  ; '

are defined belos:

h l

ACAD/ CAM - Atmospheric Containment Atmospheric Dilution / Containment Atmospneric Monitoring ANSI - American National Standards Institute i' APRM -

Average Power Range Monitor ATWS - Anticipated Transient Without Scram i BWR - Boiling Water Reactor ,

CRD - Control Rod Drive EHC - Electro-Hydraulic Control System EOF - Emergency Operations Facility  ;

GSEP - Generatirg 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 i LER - Licensee Event Report .

LLRT - Local Leak Rate Test LPCI - Low Pressure Coolant Injection Mode of RHRS i LPRM - Local Power Range Monitor  ;

MAPLHGR - Maximum Average Planar Linear Heat Generation Rate i MCPR -

Minimum Critical Power Ratto  :

MFLCPR -- Maximum fraction Limiting Critical Power Ratio .

MPC -

Maximum Permissible Concentration  !

MSIV -

Main Staam Isolation Valve  ;

NIOSH - National Institute for Occupational Safety and Health t PCI - Primary Containment Isolation  !

PCIONR - Preconditioning Interim Operating Management Recommendations ,

RBCCW - Reactor Building Closed Cooling Hater System i RBM - Rod Block Monitor RCIC - Reactor Core Isolation Cooling System .' !

RHRS - Residual Heat Removal System RPS - Reactor Protection System  ;

RHM -

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 honitor TBCCH - Turbine Building C10 sed Cooling Water System TIP - Traversing Incore Probe TSC - Technical Support Center 0027H/00612

. _ . _. . _ _ __ _ . _ . - _ __