Text

Monthly Operating Repts for Oct 1989 for Quad-Cities Nuclear Power Station Units 1 & 2.W/891101 Ltr
	ML19327C116
Person / Time
Site:	Quad Cities
Issue date:	10/31/1989
From:	Deelsnyder L, Robey R; COMMONWEALTH EDISON CO.
To:	; Office of Nuclear Reactor Regulation
References
	0027H, 0061Z, 27H, 61Z, RAR-89-74, NUDOCS 8911200108
	Download: ML19327C116 (22)
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. o ound Cnos Nucket Power Station 22710 206 Avenue IJorth Cordova, llunos 61242 9740 Te4 phone 3094M2241 1

'l RAR-89-74 November 1, 1989 t

Dirtctor of Nuclear Reactor Regulations U. S. Nuclear Regulatory Comission Mall Station PI-137 Wasl61ngton, D. C. 20555 is

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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 October, 1989.

Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION j i

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R. A. Robey !

- , Technical Superintendent RAR/LFD/djb I Enclosure i

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[ QUAD-CITIES NUCLEAR POWER STATION [

[ UNITS 1 AND 2 [

MON 1HLY PERFORMANCE REPORT

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( OCTOBER, 1989 i COMMONWEALTH EDISON COMPANY i

AND :

IONA-ILLIN0IS GAS & ELECTRIC COMPANY

[

NRC DOCKET NOS. 50-254 AND 50-255 {

LICENSE NOS. DPR-29 AND DPR-30 f

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TABLE OF CONTENTS I, Introduction l II. Summary of Operating Experience i

A. Unit One :

B. Unit Two III. Plant or Procedure Changes, Tests, Experiments, and Safety i Related Maintenance ;

I A. Amendments to Facility License or Technical Specifications B. Facility or Procedure Changes Requiring NRC Approval '

C. Tests end 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 L

A. Main Steam Rellef Valve Operations [

B. Control Rod Drive Scram Timing Data .

i VII. Refueling Information -

i VIII. Glossary L t

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! I i^ .. INTRODUCTION <

p i i Quad-Cities Nuclear Power Station is composed of two Bolling Water !

l Reactors, each with a Maximum Dependable Capacity of 769 MWe Net, located in I 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 Boiling Water Reactors. The Architect / Engineer was Sargent & Lundy, Incorporated, and the primary :

construction contractor was United Engineers & Constructors. The Mississippi i River is the condenser cooling water source. The plant is subject to license j 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 critica11tles for Units One and Two, respectively were October 18, 1971, and April 26, 1972. Commercial generatio.) 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 The End of Cycle Ten Refueling Outage activities continued normally for Unit One during the month of October. The vessel was flooded, and the core was reloaded beginning on October 17 and completed on October 21. Core reload verification was begun on October 21. Three fuel bundles were dis-covered to be in the wrong position, and these discrepancies were corrected.

Core reload verification was successfully completed on October 22. Control rod friction testing was performed on October 22 thru October 24. One control .

rod drive failed the testing, and the fuel bundles surrounding this control rod were removed from the vessel. On October ?.7, the rod was then friction tested with no fuel and successfully passed the test. The fuel was then reloaded, the rod was again tested, and all CRD friction testing was successfully completed on October 27.

B. Unit Two Unit Two began the month of October operating in Economic Generation Control (EGC). Normal operational activities and routine surveillances were performed during the first week of the month.

On October 7, a power reduction was taken to make a drywell entry due to a packing leak on the high pressure drain valve for jet pump #16. The packing was replaced and power levels were adjusted according to the demands of the Chicago Load Dispatcher. On October 9. the unit was placed in EGC.

On October 12, at 0140 hours0.00162 days 0.0389 hours 2.314815e-4 weeks 5.327e-5 months . Unit Two received a full reactor scram. The first hit showed condenser low vacuum. The event typer showed main turbine stop valve closures. Actual condenser vacuum was normal. At the time of this event, Electrical Maintenance was working on the #2 Main Steam Valve limit switch. Upon investigation, it was determined that two leads were not lifted during the replacement of the limit switch on #2 stop valve.

This resulted in closure of stop valves 1, 3 and 4 when the limit switch was removed. Maintenance completed their work on the limit switch and the reactor was made critieni on October 13, et 2350 hours0.0272 days 0.653 hours 0.00389 weeks 8.94175e-4 months . On October 14, at 1225 hours0.0142 days 0.34 hours 0.00203 weeks 4.661125e-4 months , the generator was synchronized to the grid. A power ascent to 400 MWe was taken. On October 15, power ascent to full load was takea with control rods. On October 17, power levels were adjusted, and the unit

was placed in EGC.

For the remainder of the month, normal operational activities and toutine 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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III. P. ANT OR PROCEDURE CHANGES. TESTS EXPERIMENTS. AND SAFETY t R.: LATED MAINTENANCE >

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

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

-C. Tests and Experiments Requiring NRC Approval There were no Tests or Ir.periments 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 reporttag 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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SUMMARY

i WORK REQUEST NO.t Q67221

' LER NUMBER: N/A COMPONENT System 6600, 6700 - While performing monthly procedure QOS 6600-1 ,

(Diesel Generator Monthly Load Test) in preparation for the Diesel Ganerator '

monthly lmaintenace, the Unit One Nuclear Station Operator (NS0) attempted to '

y close the 1/2 Diesel Generator (DG) to Bus 13-1 breaker, but the breaker failed ;

y to c3ose. The breaker was racked out, inspected, and then racked back in with '

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no problems found. The breaker was then successfully closed. Several subsequent

, closures of the 1/2 Diesel Generator to Bus'13-1 were successfully completed. .

l To comply with Technical Specifications, procedure QOS 6600-3 (Shared Unit (1/2) !

(,, Diesel Generator Outage Report 7 Day Limitation) was. started. The 1/2 Diesel Generator was declared operational before the requirements for QOS 6600-3 could !

[, be completed, f

CAUSE OF MALFUNCTION: The cause for the 1/2 Diesel Generator's initial failure to close in to Bus 13-1 could not be determined. This was partially due to :

.the fact that Operating Personnel racked the 1/2 DG to Bus 13-1 breaker out t

before Electrical Maintenance personnel could inspect it in its failed position. j

.Upon completion of their inspection. the Electrical Maintenance personnel found <

( nothing that would cause the failure of the 1/2 DG to close in to Bus 13-1. 7 F .

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RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal due to the fact that all safety functions were operational in r addition to each of the unit Diesel Generators. In addition, the 1/2 DG was !

successfully closed in to Bus 13 1 in a timely manner. t ACTION TAKEN TO PREVENT REPETITION: The initial corrective action was to inspect I the 1/2 DG feed breakers to Bus 13-1 and-to Bus 23-1 under Work Request Q67221. ,

Since no definite cause was found for the failure, strip chart recorders were !

placed in the 1/2 DG logic to Bus 13-1 and Bus 23-1. As an interim measure, !

the Operating Department has been instructed to immediately contact the Electrical !

Maintenance Department if another closure failure occurs and not to rack out

~i the circuit breaker prior to Eleccrical Maintenance inspection. A supplemental 6 report will be written if a cause for the 1/2 DG not closing into Bus 13-1 l J s is found..

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LER' NUMBER: N/A {

, COMPONENT: System 1700 - While Operating at 100 percent of rated core thermal ,

power, a fault in the "C" Main Steam Line (MSL) radiation monitor caused a r channci "A" 1/2 scram and a 1/2 Group 1 isolation when the Nuclear Station !

Operator.(NS0) could not turn the monitor display on. Instrument Maintenance r was notified and a blown main chassis fuse was found. The fuse was replaced !

and the alarms were reset. The "C" MSL tadi6 tion monitor was then replaced I with a calibrated spare. After repair of the original MSL radiation monitor, ,

the monitor was reinstalled and returned to service. l t

CAUSE OF MALFUNCTION: The cause of the failure of the "C" MSL radiation monitor

.was a faulty low voltage power supply (LVPS) which blew the fuse causing the )

1/2 scram and 1/2 isolatian.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of this event l were minimal because the other three MSL radiation monitors were functional .

to perform their primary function. Also, the blown fuse resulted in a failure in the conservative direction. -

ACTION TAKEN TO PREVENT REPETITION: Work Request Q73541 was written to investi- ,.

gate and repair the "C" MSL radiation monitor. A calibrated spare was installed ;

while repairs were being made. Work Request Q73608 was written to fix the '

LVPS of the original MSL radiation monitor. Due to the high failure rate. [

the station has submitted some failed LVPS's to General Electric for analyris. t l

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UNIT 2 MAINTENANCE

SUMMARY

WORK REQUEST NO.: Q78045 LER NUMBER ' N/A i

COMPONENT 3. System 0203 - While performing QoS 250-8, MSIV Testing at Cold ,

. Shutdown, the 2-203-ID. Inboard Main Steam Isolation Valve (MSIV) failed to {

close completely when air.was bled down:from its operator during *.he fail safe -

test. Thef4A IB and IC MSIV's failed closed satisfactorily. Work Request j Q78045 was: written to have the Mechanical Maintenance (MM) Department investigate !

the proble's. . ' The valve was successfuly repaired and f ail-saf e tested.

l CAUSE OF MALFUNCTION: The cause of the event is unknown. The valve packing had recently been changed to live load packing opposed to the conventional ;

lantern ring type packing. '

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences were minimal ,

due to the fact that: '

1. 2D MSIV (outboard) was fully operable as a redundant valve to the r

.lD MSIV as the 2D and all the other outboards passed their fail- I safe testing. i 2.- ID MSIV was found to close within 1 - 1/2 inches of fully closed, j chich would still greatly limit steam flow during an accident t ondition and worst case conditions (i.e. 2D MSIV not closing 4 -

at all). Additionally, it is suspected that steam flow would f have assisted in closing the valve. :

t ACTION TAKEJ TO PREVENT REPETITION: Immediate corrective action was to write <

f Work Request Q78045 to investigate the problem. Follow-up actions included- l the repair and retest of the ID MSIV. Outboard MSIV's were then tested successfully.to complete QoS 250-8.

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I' IV. LICENSEE EVENT REPORTS !

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-The following is a tabular summary of all licensee event reports for [

r Qaad-Cities Units One and Two occurring during the reporting period, r pursuant to the reportable occurrence reporting requirements as set forth ;

in sections 6.6.B.1, and 6.6.8.2. of the Technical Specifications. j UNIT 1 !

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L Licensee' Event l

Report Number Date Title of Occurrence f r

[ 89-017 10-22-89 Semi-Annual Surveillance i

(QO3 500-3) Not Completed. [

! On Time i r !89-018 10-12-89 Potential DG Operation I l at Reduced Voltage r

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UNIT 2 :

..89-005 10-12-89 Reactor Scram from Turbine-

! Stop Valve Closure ,

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V.. DATA TABULATIONS ,

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The following data tabulations are presented in-this report:. :

A. ' Operating Data Report

8. . Average' Daily Unit Power Level ,

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, C. Unit Shutdowns and Power Reductions -

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. , s, ; .: . i APPENOlX C {

l OPERAT6MO DAYA REPOR7 00C8(ET NO. 50-254 UNIT One __

DATg November 6, 1989 C048PL8780 gy 1.ynne Deelsnyder TELEPH0 peg 309-654-2241 l

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080447000STA M 0000 100189 103189 745 ,

1.REPSImes0 Pens 00, 2400 GR0es w0VReiN RePORTwe0 Pene00: l man. Ospeest. CAPatsfy stesse.atseli 769 )

2511

& OWAReleTLY AUThefM890 POWOR L8v8L tesent.

l 00B00018648TeeCAL RATHet leflue.sesui N/A

& POUIOR LOYSL T$ quMe0N RWTRICT80 IIP ANY) meesel:

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4. IISANIS POR ItsTheC?telt ilp Alms i I

1 TMs6 as00ml YA TO DATO CWheW6AftV8 t

5767.4 1211no.6 .

8. leWbs0WI 0F NOURS RSACTOR 1248 CRITICAL . . . . . . . . . . . . . . . O. 0 -

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0. 0 0.0 3421.9_

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4. RSANT0fl RNGRvt SMWTOSHII te0URS . . . . . . . . . . . . . . . . . .

5655.2 119314.4 l f.HOWAB000e8AATOR000LiteE ......................... 0.0 0.0 909.2 S. Wess? RegeRye SMWTOOWR MOWRg . . . . , , . . . . . . . . . . . . . . , , 0. 0 :

0 12309578 253999657 l

9. GROSS TwenteAL siegAST GesetRAfte infeuMI ............. t 0 3952822 82310435 ;

16. OR000 SLOCTR6 CAL SNOROY 080etRAfgG teseR4. . . . . . . . . . . . .

3763107 77328381 l

11. let? ELOCTR6 CAL SNOROY 0$NGRAT80 ttfusMI . . . . . . . . . . . . . . -5332

0. 0 70.0 80.I

18. ' EACTOR 98RvlC8 P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . .

0. 0 79.0 82.4 ,

13. RSACTOR AvalLASILITY P ACTOR . . . . . . . . . . . . . . . . . . . . . .

0. 0 77.5 77.5 [

14. WMT 98Rvice P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

77.5 78.1 i 1.. W.T Av A,u..utv P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . . . o o 67.I 65.3 SS. Uself CAPACITv PACT 0ft IWeme 08051 .....................-0.9 f 65.4 63.7

17. U8 elf CAPAC.TV PACTOR IWeme genom teveel . . . . . . . . . . . . . . . . . -0,9 6.5 5.4 l

it. Uself PORC80 0bTAGE RATS . . . . . . . . . . . . . . . . . . . . . . . . . 0. 0

19. SNW700 ness SCM00WL40 OvtR Nax7 e asomTMs t7YPt. DATE. AND OURATION OP GACNt: f

30. IP SMUT DOWN AT 840 OF REPORT Ptml00. 88TIMAT90 DATE OP STARTUP: l

21. VielT5 IN TEST ST ATV6 IPRIOR TO CouwsmCI AL OPERAT10NI; POR0CA9T ACHItv80 . !

r INifl&L CRITICALITY 18elflAL 8L8CTfWC47Y r C00R8881CIAL OP8AAft006 1.1 M

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OPERATIN0 0ATA REPORT I

00 cutt NO. 50-265 j i

UNIT M --

l OAfg November 6, 1989 l COMPLET80 gy Lynne Deelsnyder _ l Tel.8PHONg 309-654-2241 l l t

OptRAflNG STATUS 0000 100189 2400 103189 GRees NOURelN REPORfnes pene00: 745 __

j g, py0RTeh0 M MAa.00Ptose. CAPACITY Rfluefesel; 769 l

. E cWRRessTty AUTMORe800 POWBOR L8v8L Y %. 2511 OmeGee GLOCTRICAL RATvese tasuureesel:

& POWOR LSVOL TO UnseCM RGBTRICTtt top ANYI tesuursessi:

N/A f a AsAstuse0RRestneCTien nP Amyli

m. t.0,,T., .R ,0 0AT. - ,.V. ;

6970.7 117920.6- +

8. DeWeethi 0F MOURS R4 ACTOR U1148 CRITICAL . . . . . . . . . . . . . . 698. 8 O.0 0,0 2985.8_ t

0. RSACT0fl ROBORVE SMWT90suel MOWAB . . . . . . . . . . . . . . . . . . .

6902.8 114634.5 j

7. M0WR$ 0t#4RATOR ON Lifet . . . . . . . . . . . . . . . . . . . . . . . . . 6 8 6. 3 0.0 0.0 702.9 l 8, Wesef R800 Rye SMWTecame Mr:WRs. ......... ,,........ ,.

1549855 15183149 246093422 j S. Omges TMcAMAL tee 0ROY 4844RATSO testuni . . . . . . . . . . . . . .

508962 4923047 78856518

10. GROSB SLSCTRICAL ENERGY 088stm Afte lesuent . . . . . . . . . . . . .

488795 4707403 74443980 j

11. leet OL8CTReCAL BNOROY OSNGRAft0 lbfuuMI . . . . . . . . . . . . . .

93.8 95.5 77 L ,

12. ' EACTOR $8mvict P ACTOR . . . . . . . . . . . . . . . . . . . . . .

93.8 95.5 79.4 [

13. REACTOR AV AILAtl8.iTY P ACTOR . . . . . . . . . . . . . . . . . . . . . . .

92*1 94.6 75.3 f

14. WMT SSRVICE P ACT OR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

f. 94.6 75.7 l L 19. Wesef AV AILAtlLITY P ACTOR . . . . . . . . . . . . . . . . . . . . . . . . 9 2. I l

85. 3 83.9 _

63.6

14. Weelf CAPACITY F ACTOR IWeens 800Cl . . . . . . . . . . . . . . . . . . . .

83. 2 81.8 61.9 J f 17. Weelf CAPACITY P ACT0ft 4Wesne Doulpi teveel . . . . . . . . . . . . . . . .

7.9 4.6 8.2

i. WMT PORC.0 0WT AO. T. . . . . . . . . . . . . . . . . . . . . . . . . . .

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19. SMW700 West 3CM80WL40 OvSR NEXY 8 M00sTMS (TYPE. CA78. A8ee DURATION OP 5ACMI:

30. IP SMWT 90svis AT tes0 0P R8 PORT PER100. ESTIMAT80 0 ATE OP STARTUP:

PORSCA87 ACMitv80 ;

21. UMTS 188 TEST ST ATW8 (PR60m TO COMMERCI AL OPERAfteseh leelflAL CRITICALITY laelTIAL ELSCTRICITY COMassRCIAL OPGRAfleep 1.1M ;

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'I APPENDIX S i AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-254 j UNIT On'

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DATE November 6 1989 {

l COMPLETED BY tvaa- Deelanvder !

1ELEPHONE 309-654-2241 i

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MONTH ,

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. t DAY AVERAGE D AILY 19WER LEVEL DAY AVERAGE DAILY POWER LEVEL U (MWe het) (MWe Net) l 1 -6 3, _3 l 3

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3 -7 19 _

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gg. . -3 -3 36 11 -3 gy -3 l 13 -3 3 -3 l 13

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INSTRUCTIONS On this form. list the avorop daily unit power level in MWeNet for each day in the reporting month. Compute to the nealsst whole megawatt.

Thew figures wdl be used to plot a graph for cach reporting month. Note that when trutimum dependable capacity is !

uad for the net electrical rating of the unit. there may be occasions when the daily averer power level exceeds the ;

1001 hoe (or the rntracted power level line). In such cases, the average dely umt power output sheet should be :

I footnoted to caplam the apparent anomaly.

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.' AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-265 :

UNIT Two l

DATE Novernher 6, 1989 l COMPLETED BY tynne Deelsnyder f TELEPHONE 309-654-2241 f

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MONTH 1989 I

i DAY AVERAGE DALLY l9WER LEVEL DAY AVERAGE DAILY POWER LEVEL !

(M #e h et) (MWe Net) l 1

728 17 760 l

3 727 gg 724 !

I 3 751 jg ,

767 4 75o 30 726 j g 723 gg ggy [,

6 743 698 33 747

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7. _ gg 7tg 8 417 34 746 I 9 782 35 732 l 10"

733 36 739 l

11 745 ty 720 )

i 18 720 j 13- 3 13 -11 33 780 f 14 149 3p 726 !

l I 15 572 33 803

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INSTRUCTIONS On this form, list the awrses daily unit power level in idWe Net for each day in the reporting month. Compute to the 7 nesisst wiions melawatt.

' Thew figures will be used to plot a ysph for each reporting month. Nute that when trenimum dependable capucity is uwd for the net electrical rating of the unit, there may be occasions when the daily average power level escoeds the 1001 line (or the restricted power level line). In such cases, the average daily unit power output sheet should be footnoted to caplaus the apparent antmely. '

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.II/5A APPEWIID QTF 300-513 UNIT SNUTDOWS AND FUWER REDUCTIONS Revision 6 DOCKET NO. 50-265 August 1982 IRIIT NAfE QUAD-LITIES UNIT TWO CofM'ETED BT L. DEELSNTDER DATE November 2, 1989 IIEPORT MONTN October. 19d9 TEMPNONE 309-654-2241

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h$% g h LICENSEE @ @@"

DLEIATION R EVENT $0 $

u NO. DATE (NOURS) REPORT FO. CORRECTIVE ACTIONS /ColWENTS 89-21 891008 F 0.0 H 5 VALVEX Power Reduction Taken To Repair Leaking Valve Packing in Drywell 89-22 891012 F 58.7 H 3 89-005 INSTRU Reactor Scrr.mmed Due to Personnel Error When Performing Work on #2 Main Steam Valve Limit Switches - Missed Two Leads j on Drawing in Work Package When Replacing Limit Switch 89-23 891021 F 0.0 H 5 VALVEX Power Reduction Taken to Furmanite Valve in Drywell APPROVED AUG161982 (final) yco3g

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.J 'O e VI. UNIOUE REPORTING REQUIREMENTS ,

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

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

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

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l' VII.

REFUELING INFORMATION

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

,, O'Brien to C. Reed, et al., titled "Dresden, Quad-Cities, and Zion

' Station--NRC Request for Refueling Information", dated January 18, 1978.

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OTP 300-532 !

Revision 2 0'JAD CITIES REFUELING October 1989 i INFORMATION REQUEST l

1 Unit: 01 Reload: __

9 Cycle: 10 f

\

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

3. Scheduled date for restart following refueling: 11-18-89 )'

4. Will refueling or resumption of operation thereafter require a Technical t Specification change or other license amendment: 3 i

NOT AS YET DETERMINED.

5. Scheduled date(s) for submitting proposed licensing action and supporting information: f JUNE 10. 1989 f

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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: i NONE AT PRESENT TIME.

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7. The number of fuel assemblies. *

a. Number of assemblies in core: 724 :

b. Number of assemblies in spent fuel pool: 1536 !

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

a. Licensed storage capacity for spent fuel: 36s7 i

b. Planned increase in licensed storage: 0 l, 9. The projected date of the last refueling that can be discharged to the !

spent fuel pool assuming the present licensed capacity: 2008 '

f APPROVED i

(final) 14/0395t g 34 g C).C.O.S.R.

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QTP 300-532 l l

Revision 2 !

QUAD CITIES REFUELING October 1989 ;

INFORMATION REQUEST i

1. Unit: 02 Reload: 9 Cycle: 10 l

2. Scheduled date for next refueling shutdown: t

, 2-3-90 i

3. Scheduled date for restart following refueling: '

s-s-90

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

NOT AS YET DETEIO!!NED.

5. Scheduled date(s) for submitting proposed licensing action and supporting information: l NOVEMBER 2, 1990

6. Important licensing considerations associated with refueling, e.g., new or different fuel design or supplier, unreviewed design or performance '

analysis methods, significant changes in fuel design, new operating procedures:

NONE AT PRESENT TIME.

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7. The number of fuel assemblies.

4. 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. 1.icensed storage capacity for spent fuel: 3897

b. Planned increase in licensed storage: 0 -

9. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2008 P

APPROVED ,

(final) 14/0395t OCT 3 01969 O.C.O.S.R.

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VIII. GLOSSARY .)

! The following abbreviations which may have been used in the Monthly Report, .

are defined below: l ACAD/ CAM - Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring -

ANSI -

American National Standards Institute ;

e APRM - Average Power Range Monitor l ATHS - Anticipated Transient Without Scram ,

c ; BWR - Boiling Hater Reactor !

! CRD -

Control Rod Drive ,

I, EHC - Electro-Hydraulic Control System i EOF - Emergency Operations Facility ;

GSEP- - , Generating Stations Emergency Plan HEPA. -- High-Efficiency Particulate Filter -

HPCI - High Pressure Coolant Injection System f HRSS - High Radiation Sampling System IPCLRT .- Integrated Primary Containment Leak Rate Test IRM - tr.Drmediate Range Monitor j

!SI - Inse.*vice Inspection !

LER - Licensee Event Report :

LLRT -

Local Leak Rate Test :

LPCI - Low Pressure Coolant Injection Mode of RHRS l LPRM - Local Power Range Monitor :

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

Minimum Critical Power Ratio i MFLCPR - Maximum Fraction Limiting Critical Power Ratio :

MPC' -

Maximum Permissible Concentration MSIV -

Main Steam Isolation Valve 'i l' NIOSH - National Institute for Occupational Safety and Health PCI - Primary Containment Isolation !

PCIOMR - Preconditioning Interim Operating Management Recommendations !

RBCCH - Reactor Building Closed Cooling Water System RBM - Rod Block Monitor RCIC. - Reactor Core Isolation Cooling System ;