ML20064K911

Monthly Operating Repts for Jan 1983

ML20064K911
Person / Time
Site:	Quad Cities
Issue date:	02/03/1983
From:	Beth Brown, Buss R COMMONWEALTH EDISON CO.
To:
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ML20064K903	List: ML20064K902 ML20064K911
References
NUDOCS 8302150048
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Text

QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT JANUARY 1983 CCNMONWEALTH EDISON COMPANY AND-IOWA-ILLINOIS CAS & ELECTRIC COMPANY NRC DOCKET-NOS. 50-254 AND 50-265 LICENSE NOS. DPR-29 AND DPR-30 l

l 1

l f

8302150048 830203 PDR ADOCK 05000254 PDR R

O TABLE OF CONTENTS I.

Introduction II.

Summary of Operating Experience A.

Unit One B.

Unit Two III. Plant or Procedure Changes, Tests, ExperLnents, and Safety Related Maintenance A.

Amendments to Facility License or Technical Specifications B.

Facility or Procedure Changes Requiring NRC Approval C.

Tests and Expertnents 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.

INTRODUCTION Quad-Cities Nuclear Power Station is compased 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 Canmonwealth 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 condenser cooling method is a closed cycle spray canal, and the Mississippi River is the condenser cooling water source. The plant is subject to license numbe rs DPR-29 and DPR-30, issued Octdber 1,1971, and March 21,1972, respectively, pursuant to Docket Numbers 50-254 and 50-265. The date of initial reactor criticalities for Units 1 and 2 respectively were October 18, 1971, and April 26, 1972.

Commercial generation of power began on February 18, 1973 for Unit 1 and March 10, 1973 for Unit 2.

This report was compiled by Becky Brown and Randall Buss, telephone number 309-654-2241, extensions 127 and 181.

II.

SUMMARY

OF OPERATING EXPERIENCE A.

UNIT ONE January 1-31: Unit One began the month increasing load af ter a load reduction for maintenance.

On January 4, load was reduced rapidly to approximately 320 MWe during the unit's flow drop test, followed immediately by a load increase to 820 MWe which was completed by January 6.

Three times during this month the load was decreased to approximately 700 MWe to perform weekly Turbine tests. An average load of approximately 820 MWe was maintained at all other times.

On January 26, load was decreased to 700 MWe to change Condensate pumps.

Load was then increased to 825 MWe by 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />.

On January 27, at 2150 hours0.0249 days <br />0.597 hours <br />0.00355 weeks <br />8.18075e-4 months <br />, load was decreased 100 MWe to 720 MWe due to a steam leak in the IB Off Gas Recombiner. At 2230 hours0.0258 days <br />0.619 hours <br />0.00369 weeks <br />8.48515e-4 months <br />, load was increased at 50 MWe/ hour for two hours.

On January 29, at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, load was decreased 100 MWe/ hour to 700 MWe for weekly Turbine tests.

Load was then increased to 831 MWe by 1420 hours0.0164 days <br />0.394 hours <br />0.00235 weeks <br />5.4031e-4 months <br />.

B.

UNIT TWO January 1-10: Unit Two began the month holding load at approximately 780 MWe.

On January 5 load was decreased to approximately 590 MWe by 2150 hours0.0249 days <br />0.597 hours <br />0.00355 weeks <br />8.18075e-4 months <br /> to change the Steam Jet Air Ejectors due to a steam leak.

At 2310 hours0.0267 days <br />0.642 hours <br />0.00382 weeks <br />8.78955e-4 months <br />, load was dropped to 500 MWe to return the original Steam Jet Air Ejectors to service af ter the repairs were completed. A load increase was begun at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> on January 6, beginning at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, and the load reached approximately 772 MWe at 1415 hours0.0164 days <br />0.393 hours <br />0.00234 weeks <br />5.384075e-4 months <br />.

Load was then held due to Condensate Demineralizer problems. At 1520 hours0.0176 days <br />0.422 hours <br />0.00251 weeks <br />5.7836e-4 months <br />, load was dropped to 650 MWe.

This was necessitated by high Reactor water conductivity caused by a Condensate Demineralizer failure.

On January 7 load was dropped to 550 MWe for control rod pattern adjustment. At 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br />, load began increasing at 5 MWe/ hour to 790 MWe on January 9 January 11-16: On January 11, at 1344 hours0.0156 days <br />0.373 hours <br />0.00222 weeks <br />5.11392e-4 months <br />, the unit scrammed on an erroneous high steam line flow signal and subsequent main steam isolation valve closure caused by contractor personnel Jarring an instrument rack. The unit was on line at 0023 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> on January 12 and load was increased to 773 MWe by January 13 On January 14 load was dropped 200 MWe to backwash and precoat a Condensate Demineralizer.

Load then increased to 744 MWe.

On January 15, load was decreased to 500 MWe to reverse flow through the main Condenser in order to reduce high backpressure.

Load was then increased to approximately 780 MWe on January 16.

B.

UNIT TWO:

(continued).

January 17-31: On January 17, load was decreased to 328 MWe to investigate Primary Containnent leakage. This leakage was a result of a 2A Recirculation pump loop valve packing' leak.

Following adjustments, the load was increased to 668 MWe by 2220 hours0.0257 days <br />0.617 hours <br />0.00367 weeks <br />8.4471e-4 months <br />.

Unit load was held here due to high Condenser backpressure. On January 18 load was dropped to 520 MWe to take the 2C Circulating Water pump out of service to repair a casing leak.

Load was then increased.to approximately 650 MWe.

On January 28, load was reduced at 100 MWe/ hour in preparation for a scheduled Maintenance Outage to repair the 2C Circulating Water pump. At 1830 hours0.0212 days <br />0.508 hours <br />0.00303 weeks <br />6.96315e-4 months <br />, the Generator was taken off line. The Reactor was manually scrammed at 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br />. The Maintenance Outage continued through the end of the month.

III.

PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A.. Amendments to Facility License or Technical Specifications On December 15, 1982, the NRC issued Amendment 83 to License DPR-29 This Amendment provides changes to the.

License and Technical Specifications required to operate Unit One with the fuel load for Cycle 7 Unit One began operating in Cycle 7 on December 22, 1982.

On December 23, 1982, the NRC issued Amendments 84 and 77 to Licenses DPR-29 and DPR-39, respectively. This Amendnent adds requirements for verifying that the scram discharge volume drain and vent valves are open every month.

B.

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

C.

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

D.

Corrective Maintenance of Safety Related Equipnent The following represents a tabular summary of the safety related maintenance performed on Unit One and Unit Two during the reporting period. This summary includes the following headings: Work Request Numbers, LER Numbers, Components, Cause of Halfunctions, Results and Effects on Safe Operation, and Action Taken to Prevent Repetition.

._m.-

UNIT Of1E MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q21639 RV-1-4711 The relief valve seat Local Leak Rate Test The valve seats were was worn, causing was found to leak at relapped and the valve to lift at low greater than 10.36 li f ting pressure was p res s u re.

scfh.

adj usted.

i Q23549 TIP #5 Ball The ball valve seats The ball valve sticks.

The ball valve was Valve were binding against Containment isolation replaced and cycled the ball.

was maintained at all satisfactorily.

times.

Q23721 Unit I Diesel Leaky Y strainer.

Ai r leaking f rom "Y" on Replaced Y strainer.

Gene ra to r starting air header.

(1-6601)

Diesel Generator operability was un-a f fected.

Q23024 TIP #4 Ball The valve seats TIP #4 ball valve was The valve internals were Valve were di rty, sticking.

Containment cleaned, and the spring isolation was maintained tension was adjusted.

at all times.

Q23925 83-5/03L llPCI Signal A resistor in the The automatic flow Replaced the flow Converter flow controller controller was in-controller circui t 1-2306 power supply had operable, but IIPCI amplifier, and resistor burnt out.

could be operated in -15 VDC power supply.

manually.

i

UNIT TWO MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q23332 33-2/03L 2-703 TIP #4 The valve internals

1. 4 TIP ball valve will The valve was cleaned were dirty, and the not close. The manual and the spring tension spring tension on the ball valve was closed was adjusted.

solenoid operator was to maintain containment out of adjustment.

isolation.

Q22241 2B RHR lieat The expansion joint Leakage from the crack The expansion joint on Exchanger on the drain f rom to the service water did the drain line was 2B-1003 the tube side had not exceed the release replaced.

cracked.

limits. The heat exchanger was still operable.

Q23837 2-2001-16 The solenoid operated The valve will not open The solenoid valve and pilot valve was worn. when given open signal.

pipe elbow were replaced.

I

IV.

LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for Quad-Cities Units One and Two occurring during the reporting period, pursuant to the reportable occurrence reporting requirements as set forth in sections 6.6.B.1. and 6.6.B.2. of the Technical Specifications.

UNIT ONE Licensee Event Report Number Date Title of Occurrence 83-1/03L I-04-83 1/2 Diesel Generator Failure to Run 83-2/03L 1-09-83

' A' RHR Heat Exchanger Leak 83-3/03L I-18-83 1-263-58A Level Switch Out of Calibration 83-4/03L I-19-83 Unit One HPCI Speed Changer 83-5/03L 1-15-83 Unit One HPCI Motor Gear Unit Failure to Stay at High Speed Stop UNIT TWO 83-1/03L 1-02-83 Of f Gas Moni to r Fai l ure 83-2/03L I-06-83 Reactor Water Conductivity Greater than 10 umho 83-3/03L 1-10 TIP #4 Ball Valve Failure 83-4/03L 1-16-83 2B RHR Heat Exchanger Out of Service for -Tube Leak

V.

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

A, Operating Data Report B.

Average Daily Unit Power Level C,

Unit Shutdowns and Power Red 6ctions t

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OPERATING DATA REPORT

_ _ ____ _ ___.._ _ _.. UNIT ONE DATEFebroorv 01 1983 COMPLETED BYRondall D Buss

.-._ _ -.____TELEP. HONE 309-654-224ix181._....

.m OPERATING STATUS

.__0000_010183_ _ ___ _____-_,_.__. _._, _.

___z__..

1.. Reporting period 2400 013183 Gross hours in reporting periodi 744

2. Currently.outhor.1 zed. power le v el _( MW t ) _2511..Ho x. De p e n d._ c ap ac i t y. _ _._.

.i._ m _

(MWe-Net): 769* Design electrical rating (MWe-Net): 789

3.. Power level to_which restricted (if_ony)(MWe-Net.)__NA___

4.

Reasons for restriction (if any).

Yr.to'Date Cumulative This Month

5. Number of. hours reactor _was_criticol___

744,0-.

744.'O.,

75915.2 _

6, Reactor reserve shutdown hours 0.0 0.0.

3421.9

..__-_..__._.._.-.____-._1 7.

Hours generator on line 744.0z 744,0 72830.6

_8.

Unit reserve sh u t d own_. h o ur s...

0. 0 _.

0.0 _

909.2_._

9.

Gross thermal energy generated (MWH) 1816962

-1816962 148029953 10 ', Gross electrical energy generated (MWH) 598456 590456 47720337 ii. Net electrical energy _ generated (MWH). _

563374.

563374 44392282.,_.

12. Reactor service factor 100.0 100.0 80.7

13. Reactor ovallobility factor i00.'0- '

~

84.'4

~

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i00.0 L4.. Unit service factor 100.0 100.0_

77.5 i

15. Unit ovallobility factor 100.0 100'0 78.4

16. Unit copocity factor (Using MDC) 98.5 98.5 "Ei 4

~

~

17. Unit copocity factor.(Using Des.MWe).

96.0_

9 6. 0. __

59.8,,

18. Unit forced outoge rote 0.0 0.0 6.7

~

over next 6 months (Type,Date,and Duration of each ):

19. Shutdowns scheduled

[

20..If shutdown at end'of report period,estinated date of stortup ___NA________

• The MDC nay be lower than 769 MWe dering perleds of high anblent tenperature due Vo the thernal perfernance of the spray canal.

SUN 0FFICIAL COMPANY NUMERS ARE USED IN THIS REPORT

OPER ATING DATA REPORT DOCKET NO.

50-265 UNIT TWO DATEFebruarv 01 1983

~

COMPLETED BYRondoll D Buss

. TELEPHONE 309-654-2241xiB1 OPERATING STATUS 0000 01C183

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Reporting period:2400 013183 Gross hours in reporting period:

744 2.

Currently authorized power level._(MWt): 2511.Mox. Depend copocity (MWe-Net): 769* Design electrical rating (MWe-Net): 789 3.

Power level to which restricted (if any)(M_We-Net): NA 4.

Reasons for restriction (if any):

~~'Th's Month Yr.to Date Cumulative i

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Number of hours reactor was critical 659,7 659.7 72923,1 6.

Reactor reserve shutdown hours 0.0 0.0 2985.8

~~

7.

liours generat or on line 655,9 655,9.

70244,0 8.

Unit reserve shutdown hours.

0,0 0.0 702.9 9.

Gross thernal energy generated (MWH) 1447122 1447122 146038616

10. Gross electricc1 energy generated (MWH) 450236 450236

~ 46487771' ti. Het electrical energy generated (MWH) 422541 422541 4360610'd _

L2, Reactor service factor 88,7 88,7 78,3

13. Reactor availability factor 88,7 88.7 81,5 L4. Unit service factor 88,2 88.2 75.4 15, Unit availability factor 88,2 88,2 76,2

16. Unit copocity factor (Using MDC) 73,9 73.9 60.9

17. Unit capacity factor (Using Des.MWe) 72.0 72,0 59,4 tu. Unit forced autoge rote 11,9 11.9.

9,2

19. Shutdowns scheduled over next 6 non;hs (Type,Date,ond Duration of each):

20. If shutdown at end of report period estimated date of startup __];i_3jL_____

3

• The MDC ncy be lower than 769 MWe during perieds of high onblent temperature due to the thernal perfernance of the spray canel.

• LH0FFICIAL COMPANY NUMBERS ARE USED IN THIS REPORT

~

APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-254 l!_

~. _..DATEFebruarv 01 1983.

COMPLETED BYRandall D Buss

~. - _ _.

_ MONTH January 1983 DAY AVERAGE DAILY POWER LEVEL' DAY AVERAGE DAILY POWER LEVEL

_.. _._ _ _ M W e -- Ne t ).

(

(MWe-Net).

i.

696.5 17, 777.9 2.

754.0 18.

772.3 3.

768.6

__ _ ___ _ _.__._19..

764.8 4,

700.1 20.

772.6

_~

6, 766.2 22.

776.8 7.

772.0 23.

726.7 9.

735.5

_f _,_._ __ _2 5..

769.0 10.

752.7 26.

'767.7

.12.

754.4

_._..2 8.,_

755.1 13.

-783.0 29.

768.1 15.

777.1

_. 31, 779.9 16, 732.0

~

INSTRUCTIONS On this fern, list the overage daily snit peuer level in liWe-Net for each dot in the reporting nenth.Conpite to the nearest whole negawatt.

- -These figures will be used to plot a graph for each reporting nenth. Note that when notinen dependable capacity is-used for the net electrical rating of the snit there ney be occasions when the daily everage power level exceeds the 1981 line (or the restricted power level line),.In such cases,the average daily snit power estpet sheet sheeld be festnoted to explain the opperent onenely m

a-

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APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-265

_ _ _ _ _ _ __. DATEFebt vor v 01 1 9 8 3 __.

COMPLETED BYRondoll'D Boss

~

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TELEPHONE 309-654-2241xi81~

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MONTH.

Jonoorv 1983 DAY AVERAGE DAILY POWER LEVEL

. DAY-AVER AGE D AILY POWER LEVEL j

(MWe-Net)

____ (MWe,-Ne t )_.

i.

743.0 17.

579.1

. _ _ _ _ _. _ ~. _

3.

730.7

_19 _

590.3 4,

742.9 20, 615.0 22.

641.1 6.

606.4 7.

640.3 23.

637;4 g.

526.5

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64 ~~

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711.9

_ _ _ 2 5._

638.0

_j 10, 735.3 26, 648.0 ii.

411.8~

~^~~~~727. '

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~~

639.5

12.

484.8 28, 394.0 13, 653.4 29.

-7.7 SS T

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-5;8~~

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-5.4 15, 640.1 31..

16, 718.5 On this forn, list the overage dolly enit pour level in lille-Net for each day in the reporting nonth Compute to the nearest whole negowett.

These figures will be used to plot a graph for each reporting nonth. Note that when no inen dependable capacitv.is.

osed for the net electrical rating of the unit there noy be occasiens when the daily overage powr level exceeds the 100% line (or the restricted power level line),.In such cases,the overage daily snit pour actput sheet should be festnoted to explain the opperent onenoly

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p ID/SA APPENDIX D QTP 300-S13 UNIT S!!UTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO. _Q10-254 August 1982 UNIT NAME _ Quad-Citles Unit _0ne COMPI.ETED BY Randali Buss DATE Feb 1. 1983 REPORT MONTil January 1933 TEI.EPil0NE 309-654-2241 m

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DURATION d

EVENT g

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NO.

DATE (110URS)

REPORT NO.

CORRECTIVE ACTIONS / COMMENTS A

G3-1 830104 S

0.0 B

5 CB ZZZZZZ Load reduction in preparation for performance of Recirculation Flow Drop Test 03-2 830109 s

0.1 B

5 HA XXXXXX Reduced load to perform weekly Turbine tests 03-3 330116 5

0.0 8

5 liA XXXXXX Reduced load to perform weekly Turbine tests 83-4 030123 S

0.0 0

5 IIA XXXXXX Reduced load to perform weekly Turbine tests 83-5 030126 F

0.0 B

5 HH PuttPXX Load reduction to change Condensate pumps 8 3~-6 830127 F

0.0 B

5 t1B RE C0f tB Load reduction due to steam leak in IB Off Gas Recombiner 33-7 830129 5

0.0 B

5 llA XXXXXX Reduced load to perform weekly Turbine tests APPROVED AUG 1 G 1982 l

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- (final) ygg3g

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ID/5A APPENI)IX D QTP 300-S13 UNIT SIIUTUOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO. 050-2_6_5 August 1982 UNIT NAME Quad-Cities Unit Two COMi>LETED BY Randal1 Buss DATE Feb I.

1983 REPORT HONTil January 1983_

TELEPil0NE 309-654-2241 5

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$w h$

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EVENT

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DURATION d

g u

NO.

DATE (il0URS)

REPORT NO.

CORRECTIVE ACTIONS / COMMENTS a

ca 33-1 830105 F

0.0 B

5 llc XXXXXX Reduced load to change Steam Jet Air Ejectors due to steam leak.

83-2 830105 F

0.0 11 5

llc XXXXXX Reduced load to change Steam Jet Air Ejectors af ter repair of steam leak 83-3 830106 F

0.0 B

5 HG DElll:1X Load reduction due to Condensate Demineralizer problems 83-4 830106 F

0.0 A

5 83-2/03L llG DEMlHX Load reduction due to high Reactor

\\later Conductivity 83-5 830107 5

0.0 H

5 RB CONROD Load reduced to perform Control Rod pattern adjustments 83-6 830111 F

10 7 H

3 ZZ ZZZZZZ Reactor isolation signal and subsequent scram on erroneous High Steam Line Flow signal due to contractor personnel jarring an instrument rack APPROVED AUG 101982 (final)

VCObH a

M n

n n

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M M

M BUI M

M RWR9 n

n n

p ID/5A APPENI)lX 1)

QTP 300-S13 UNIT SiltlTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO.

050-265 August 1982 IINIT NAME Quad-Cities Unit Two COMPI.ETED BY Randall Buss DATE Feb 1, 1983 HEPORT MONTil January 1983 TEl.EPil0NE 309-654-2241 N

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d I.ICENSEE

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mo p., o P

DURATION EVENT

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U NO.

1) ATE (llotlRS)

REPORT No.

CORRECTIVE ACTIONS /CorlHENTS o

ca 83-7 830114 F

0.0 B

5 liG DErtlHX Load reduction due to Condensate Demineralizer problems 83-8 830115 F

0.0 H

5 llc ZZZZZZ Load reduction due to high Condenser backpressure 33-9 830115 F

0.0 li 5

liC ZZZZZZ Load reduction due to high Condenser backpressure 83-10 830117 F

0.0 B

5 CB VALVEX Load reduction to perform maintenance on a Recirculation Loop Crosstie Equalizer valve 83-11 830118 F

.0.0 0

5 HF PUtiPXX Reduced load to take 2C Circulation Water Pump out of service 83-12 830128 S

0.0 B

5 llF PUttPXX Load reduced in preparation of unit itaintenance Outage to repair 2C Circula-tion Pump casing 83-13 830128 5

77.5 B

2 FIF PUtiPXX Unit shutdown for flaintenance Outage APPROVED AUG 1 G 1982

-l-(final) ygg3g

VI, UNIQUE REPORTING REDUIREMENTS The following items are included in this report based on prior cannitments to the conmission:

A.

MAIN STEAM RELIEF VALVE OPERATIONS There were no Main Stean Relief Valve Operations for the reporting period.

B.

CONTROL ROD DRIVE SCRAM TIMING' DATA K)R UNITS ONE AND TWO There ns nn Control Rod Drive Scram Timing Data for Units One and Two for the reporting period.

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VII.

REFUELING INFORMATION The following information about future reloads at Quad-Cities Station was requested in a January 26, 1978, licensing memorandum (78-24) from D. E. O'Brien to C. Reed, et al., titled "Dresden, Quad-Cities, and Zion Station--NRC Request for Refueling Information",

dated January 18, 1978.

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QTP 300-S32 Revisicn 1

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QUAD-CITIES REFUELING M2rch 1978 i

INFORMATION REQUEST 1.

Unit:

Q1 Reload:

6 Cycle:

7 2.

Scheduled date for next refueling shutdown:

9-6-82

'[

3 Scheduled date'for restart following refueling:

12-18-82 F

4.

Will refueling or resumption of operation thereafter require a technical specification change or other license amendment:

Yes f

5.

Scheduled date(s).for submitting proposed IIcensing action and supporting Information:

8-19-82: Tech. Spec. changes submitted to the NRC.

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:

a) All 7x7 fuel assemblies vill be removed from the core.

b) MAPLHGR curves for fuel types in the core are being extended to h0,000 MWD /ST.

c) MCPR limits will be detemined by GE's ODYN computer code.

,(s d) The vessel pressure safety limit is being modified to accommodate the m'

potential for higher reactor pressures as c' alculated by ODYN.

7 The number of fuel assemblies.

a.

Number of assemblies I'm core:

724 l

b.

Number of assemblies in spent fuel pool:

800 d,

S.

The present licensed spent fuel pool storage capacity and the size of any l

Increase in licensed storage caoacity that has been requested or is planned in number of fuel assemblies:

0 l

a.

Licensed storage capacity for spent fuel:

3657 b.

Planned increase in IIcensed storage:

0 9.

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

[

spent fuel pool assuming the present licensed capacity: 2003 a

WPPROVED s,

c 3 APR 2.01978 a

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QUAD-CITIES REFUELING Harch 1978 i

INFORMATION REQUEST e

v 1.

Unit:

Q2 Reload:

6 Cycle:

7 m

2.

Scheduled date for next refueling shutdown:

9-11-83 3

Scheduled date'for restart following refueling:

11-20-83 F

4.

Will refueling or resumption of operat!on thereaf ter require a technical L

specification change or other license amendment:

Depending upon the Licensing analyses, a MCPR limit change may be needed.

5 Scheduled date(s) for submitting proposed licensing action and supporting I"I I* "I "*

8-22-83 (if necessary) 6.

Important IIcensing 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:

NFS intends to apply 10CFR50.59 to the Q2R6CT reload unless MCPR Technical Specification change is required.

E 7

The number of fuel assembites.

a.

Number of assemblies in core:

724 b.

Number of assemblies in spent fuel pool:

1140 n

L 8.

The present IIcensed 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 IIcensed storage:

0 9

The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2003 L

APPROVED r-d APR 2.01978 Q. c. o. S. R.

=.. -

w-w.~ =_ p : =+ :

= z = = ---

+ := - -

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 American National Standards Instinate ANSI APRM Average Power Range Monitor ATWS Anticipated Transient Without Scran BWR Boiling Water Reactor Control Rod Drive CRD EHC Electro-Hydraulic Control System EOF Bnergency Operations Facility GSEP Generating Stations Emergency Plan High-Ef ficiency Particulate Filter HEPA High Pressure Coolant Injection System HPCI HRSS High Radiation Sampling System IPCLRT Integrated Primary Containment Leak Rate Test IRM Intermediate Range Monitor ISI Inservice Inspection LER Licensee Event Report LLRT Local Leak Rate Test LPCI Low Pressure Coolant Injection Mode of RHRS Local Power Range Monitor LPRM MAPLHGR Maximum Average Planar Linear Heat _ Generation Rate Minimum Critical Power Ratio MCPR MFLCPR Maximum Fraction Limiting Critical Power Ratio Maximum Permissible Concentration MPC MSIV Main Steam Isolation Valve NIOSH National Institute for Occupational Safety and Health PCI Primary Containment Isolation PCIOMR Preconditioning Interim Operating Management Recommendations RBCCW Reactor Building Closed Cooling Water System RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling System RHRS Residual Heat Removal System RPS Reactor Protection System RWM Rod Worth Minimizer Standby Gas Treatnent System SBGTS Standby Liquid Control SBLC SDC Shutdown Cooling Mode of RHRS SDV Scram Discharge Volume SRM Source Range Monitor TBCCW Turbine Building Closed Cooling W ter System a

TIP Traveling Incore Probe TSC Technical Support Center

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