Text

Listing of Changes,Tests & Experiments Completed During Mar 1990 & Summary of Safety Evaluations,Per 10CFR50.59 & 10CFR50.71(e).W/900402 Ltr
	ML20042E418
Person / Time
Site:	Quad Cities
Issue date:	03/31/1990
From:	Robey R; COMMONWEALTH EDISON CO.
To:	; Office of Nuclear Reactor Regulation
References
	RAR-90-32, NUDOCS 9004200751
	Download: ML20042E418 (29)
	v • d • e

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Quad Citie6 Nuclear Power StLtion O Commonwealth Edison 22710 206 Avenus North oordove. Illinoit (1542 Tol6 phone 3D$/6541241 RAR-90-32 April 2, 1990 Director of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Mall Station PI-137 Washington, D. C. 20555 Enclosed please find a listing of those changes, tests, and experiments completed during the month of March,1990, for Quad-Cities Station Units I and 2. DPR-29 and DPR-30. A summary of the safety evaluations are being reported in compliance with 10CFR50.59 and 10CFR50.71(e).

Thirty-nine copies are provided for your use.

Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION

h-R. A. Robey Technical Superintendent RAR/LFD/ekb Enclosure ,

cc: R. Stols T. Watts /J. Galligan 1

90042007Ds 900402 DR ADOCKOboog4 0027H/00612 [,f I g.

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Modification M-4-1/2-86-6 Description Provide a non-safety related commercial grade facility to accommodate Illinois Departmeat of Nuclear Safe'.y (IDNS) stack sampling equipment, complete-with air compressor and diesel generator for backup power. Sample lines tie.

into existing stack strut,ture.

Evaluation

1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety.as previously evaluated-in the Final Safety Analysis Report is not increased because the modi-fication does not degrade systems or attucture important to safety and does not result in new or increased radiological consequences from previously evaluated accidents. ' Dose estimates, as a result of a tubing rupture at ground IcVel, have been shown by calculation to be within 10CTR20 limits.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because no new accidents or malfunction of a type necessary to be evaluated in-the FSAR are postulated to occur.

3. The margin of safety..as defined in the basis forLany Technical Spect-fication, is not reduced because the modification does not involve any safety-related system or structure or its associated basis, as defined in Tech Specs. Tubing loading on the. safety-related chimney-is insignificant. Doses resulting from a postulated rupture are only fractional percentages of allowable limits and'do not significantly affect the RETS or their associated bases.

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Modification M-4-1/2-87-22D.

Description This modification relocated existing DG2 mo'ters on 902-8 in order that the DG2-meters ifne-up with the equivalent DGl/2 meters on 902-8.~ No new compo- :

nents were installed by this partial modification.

Evaluation The probability of an occurrence or the consequence of an accident,

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or malfunction of equipment important to safety as previously evaluated- ,

in the Final Safety Analysis. Report 11s not increased'because this partial- I mod only includes the relocation of existing DG2 meters on 902-8.- No new meters are added by this partial mod. Therefore, no Single Failure Event or Design Basis Analysis, as evaluated in the FSAR, is affected by this partial mod.

2. The possibility for an accident or malfunction of a different type than any previously eveluated in-the-Final-Safety Analysis Report'is not. created because no new interfaces with existing systems are created by this partini mod since only existing metc.ss are relocated. The.

only interaction is the-relocation of the existing meters near other safety-related equipment on 902-8. . This interaction is mitigated through the seismic mounting of the relocated meters.

3. The margin of safety, as defined in the basis for any Technical'Spe'ci-fication, is not reduced because'this partial mod does notichange the-configuration of the existing DG2 system, but only relocates existing DG2 meters within the same panel (902-8). Therefore, the' margin of safety defined in the basis-for Tech Specs, Section 3.9, is not reduced.

No other systema discussed in the Tech Specs are affected by this partini mod.

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Modification M-4-1/2-87-22G Description Installation of red and green indicating lights (start and stop) on 902 to provide indication that the DG2 cooling water pump is running. This partial ,

mod also terminated spare conductors on existing cables and installed new fuse-blocks at switchgear 29 and panel 902-8.- New cable has been routed between local-panel 2252-98 and 902-8. Seismically supported conduit was installed.

Evaluation

1. The probability of an occurrence or the_ consequence of-an accident,.

or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because the new lights have no adverse effect on the existing panel. This partial-mod does not affect the control of the~DG2 cooling' water pump and hence the ability of the Unit 2 diesel generator to operate and provide emer-gency power. Therefore, no~ Single Failure Event or Design Basis Analysis, as evaluated in the FSAR, is affected by this partial mod.

2. The possibility for an accident or malfunction of a different type.

than any previously evaluated in the Final Safety Analysis Report'is-not created because an existing divisional cable with. spare conductors is used and is isolated from the non-safety related indicating lights circuit. The new lights, fuseblocks and_ conduits are seismically mounted to mitigate failure. The new lights circuit does not interact with the DG2 cooling water pump controls.

3. The margin of safety, as defined in the basis =for'any Technical Speci-fication, is not reduced because-the existing and new cables have been verified to be within the capabilities.of the existing; fire. detection.

and suppression systems. Existing fire stops are-rescaled with approved firestop material.- The new lights circuit doesLnot affect the operation of the DG2 cooling-water pump controls and hence the availability of the Unit 2 diesel generator. .Therefore, the margins of safety defined in the basis for. Tech Spec, Sections 3.9-and 3.12, are not reduced. No other systems discussed in the Tech Specs are affected by this partial mod.

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Safety Evaluation.#90-213- ;

Repairs to Piping Supports Units 1 and 2 The purpose of these repairs was to bring the affected piping and supports into compliance with FSAR design criteria. In 1987, it was determined that.

there were inconsistancies between the as-built and as-analyzed geometry lof; l a Mark I Large Bore'model. As a result, the piping configuration verificatien ,,

program was initiated to identify similar discrepancies on otherLlarge bore t Mark 1 lines. Most-deviations were resolved via analytical techniques, however,'

some required repairs due to technical or economic considerations. Early;in the program, operability analyses were conducted and all affected lines were determined to be operable. Repairs typically included support demolitions. A weld upgrades, support reinforcement, tee reinforcement, etc.

1. The probability of an occurrence or the consequence of'an accident, or malfunction of. equipment important to safety as previously evaluated'--

in the final Safety Analysis Report is not increased because the. purpose of these repairs is to resolve discrepancies between the as-built con-figuration of the Mark 1 large bore piping and supports and theJas-analyzed geometry. The repairs bring _the piping into compliance with FSAR criteria. Therefore, the probability of piping failure due to hydrodynamic, seismic and other loads is decreased.

2. The possibility for an accident or malfunction of a differentztype than any previously evaluated in the Final Safety Analysis Reportjis

not created because no system operating parameters or design criteria is being changed. The purpose of these repairs is to. bring the piping and supports into compliance with existing FSAR criteria.

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! 3. The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because the margin of safety is. increased.

The purpose of the repairs is to bring the piping and supports into-compliance with existing FSAR criteria. Technical Specification bases have not been impacted or changed.

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Safety Evaluation #90-224 Relocation of EQ Thermocouples TE-2-5741-430 & 43E Description EQ valves M0-2-202-5B. MO-2-2301-4, and MO-2-3706 temperatures are not 4- monitored by existing thermocouple locations. Thermocouples TE-2-5741-43C and 43E have been relocated in accordance with this: Minor Design Change.

Evaluation

1. The probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated-in the Final Safety Analysis Report is not increased because temperature sensing-has no control on the functioning of equipment.

2. The possibility for an accident or malfunction of-a different type than any previously evaluated in the Final Safety Analysis Report is not created because FSAR did not address temperature indication for the above mentioned' valves.

3. The margin of safety, as defined in the basis for any-Technical Speci :

fication, is not reduced because temperature' sensing of drywell, environ-ment is not mentioned in Tech Specs.

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'Sofety Fvaluation #90-227- l Support for Valve A0-2-1301-35 -l

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Description i i

-This Minor Design. Change modified existing pipe support on valve A0-2-1301-3$' ;

to prevent valve stem from cor.tacting the pipe-support in the:close position. La Evaluation

1. The probability of an occurrence or the consequence of an accident.

or malfunction of. equipment important to safety as.previously-evaluated in the Final Safety Analysis Report is not increased because this_ change <

will not adversely-affect operation of valve A0-2-1301-35, _ therefore, the probability of an accident is not increased.

2. The. possibility for an' accident or malfunction of a different. type than any previously' evaluated in the Finni Safety Analysis. Report'is not created because the change'to.the support will not affect.the-structural integrity.of the valve or support.

3. The margin of safety. as defined in the basis for any Technical Speci--

fication, is not reduced because the function of the support.has not' changed, therefore the margin of safety is unaffected.

't Safety Evaluation #90-230 Minor Design Change Concerning the 125 Vdc Distribution Fanel Breaker Mountings Description Mounting configuration for molded case circuit breakers within.the reactor building 125 Vdc distribution panels has changed due to a vendor upgrade of the breakers; breakers are one-half inch smaller. ..

Breaker's load side mounting bracket was modified to accommodate the new' style breaker.

Evaluation

1. The probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because the load side mounting brackets of the distribution panel breakers have no influence on the 125 Vdc system with respect to the FSAR.. This' change does not alter the design, function, or method in which the.125-Vdc system functions, as defined in'the FSAR. Catastrophic failure of the bracket would not impede or degrade the function of=this. system.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety,Analysik-Report is-not created because the failure mode analyzed in the FSAR is that;of multiple grounds. Catastrophic failure of.the bracket would notLground-the 125 Vdc bus.

3. The margin of safety, as defined in the basis for any Technien1 Speci--

fication, is not reduced because the modification to the-breaker brackets do not affect any set points, operational' limits, or special' conditions that prescr1be the margin of safety.

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'N Safety Evaluation f90-240' Changes to C-Model:

Description Routine .ED-$, ED-6 and ED-26 vill calculate true values of activity, con--

tainment activity, release rate, end release concentration. Also', a new program .

module ED-31 will'be added. This module is only applicable to PWR plants.l-Evaluation ,'

1.- The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated-in the Final, Safety Analysis Report is not. increased because all QA1 required testing has been. completed. The changes will'make the programs; casier to use.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because this program is only used for determining post accident release conditions. Adequate testing has been-performed.'

3.. The. margin of safety, as defined in the basis for any Technical-Speci-fication, is not reduced because C model program is ot,part of any Tech Spec basis.

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Safety. Evaluation.f90-241_ .

Point Trend Programs'insta11ation- '

Description Point trend has been changed to incorporate human factors improvements.;

Evaluation

1. The probability of an' occurrence _or the consequence _of an accident,~

or malfunction of equipment important to safety as previously evaluated' in the Final Safety Analysis _' Report is not increased because thel functionality of the program has been improved.' All of the originali capabilities have been retained.

2. The possibility for an accident or malfunction of-a different type than any previously evaluated in the Final Safety Analysis Report isi not created because the changes onlygimprove the' usability. .The changes have been tested in accordance-with Qp 3-54.

3. - The margin of safety, as defined-in the basis for any Technical Speci-fication, is not reduced because point trend is not mentioned in the basis for any Tech Specs.

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4 Safety Evaluation #90-242 Safety Parameter Display System Installation

- Description Various changes have'been made to meet human' factors requirements and. allow SPDS to run on the Tektronix terminal instead of RAMTEK.-

Evaluation

1. The. probability of an occurrence or the consequence of.an accidente:

-or malfunction-of equipment important to rafety as:previously-evaluated. ,

in the Final Safety Analysis Report'is not increased because no changes to the basic display or calculations have been made.

2. The possibility for an accident or malfunction of a different type:

than any previously evaluated in the Final Safety Analysis Report is not created because all changes have been tested in accordance'with QP 3-54,

3. The margin of safety, as defined in the basis for any Tcchnical Speci-fication. is not reduced because SPDS does not. appear in the basis.

for any Tech Spec.

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Safet'y Evaluation #90-258, Constant Spring Can Pipe Support ISI #3001D-W-102-Description ;

The main steam line constant spring can support,had damage to the, support steel which caused the pipe support strut to exceed'the six degree swivel litait.

The engineering evaluation by Sargent and Lundy requires removal of,the pipe ,

support. j l

Evaluation

1. The probability of an occurrence-or the consequences of'an accident or malfunction of. equipment important'to safety as previously evaluated in the Final Safety Analysis Report is not increased because this change will not affect seismic analysis, therefore, the prohability of an accident is not increased.

2. The possibility for an accident or malfunction of'a different type than any previously evaluated in the Final Safety Analysis Report is not created because the deletion of.this support will not-affect the structural integrity of this line or.other pipe supports on same line.

3. The margin of safety, as defined in the basis for.any Technical Speci--

fication, is not reduced because this work is not covered in the_ Tech Specs, so margin of safety is not reduced.

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. i r Procedure Change QAp 300-2, Revision 26 Conduct of Shift Operations Deat ription This. procedure change located the Standby' Liquid Control Injection key.

in th' control. room, removed statement about procedures use, and provided guidance on initialing _ surveillance _ sheets.

. Evaluation

1. The probability of an_ occurrence or the consequence of an accident, or malfunction of equipment _important to safety;as previously. evaluated- '

in the: Final-Safety Analysis Report is not increased because all changesi are administrative in nature and don't affect the consequence or-probability- <

of an accident.

2. The possibility for an accident.or malfunction of a different type; than any previously evaluated in the. Final Safety ~ Analysis Report _is '

not created because these changes do not effect: any plant condition that would alter the plants response'to an event.'Therefore,;no new accident is created.

3. The margin of safety, as defined in the basis for any Technica1LSpeci-fication, is not reduced because all changes are administrative in nature.

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Procedure Change QOA 010-5. Revision 9 plant Operation With the Control Room Inaccessible-Description Assign SCRE to report to the TSC during execution of this pros:edure.

Evaluation 1.- The probability of an occurrence or the-consequence.of an accident, or malfunction of equipment important to safety.as previously-evaluated-in the Final Safety Analysis Report is not increased because the change-clarifies the role and assignment of;the SCRE during this evolution.

No change in equipment operation or system response to any plant levolu--

tion has been made.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because the function of the SCRE will still be the same-but by locating him in the'TSC access to the computer is available to allow proper monitoring of both units. .No new possibility for'an.

accident or malfunction can be created by this change.

3. The margin of safety as defined in the basis for any Technical:Speci-fication, is not reduced.because stationing the SCRE in the.TSC will provide overlapping indication with plant operator indications which will permit better coordination of activities and administrative controls which will increase the crew performance. This will increase.the margin; of safety.

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.= 5 Procedure Change QGs 1300-1, Revision 12 and QOS 1300-3, Revision 6 RCIC Monthly and Quarterly Test and-RCIC Motor Operated Valve Operability Test Description Change wording from " suppression' chamber" to." torus". Add' quarterly _ freedom; of movement testing of the RCIC turbine trip and throttle valve and' delete stepsf ,

o which allow RCIC to be placed in abnormal line-up (e.g. 48 closed'and 49 open).

Evaluation ]

1. The probability of an occurrence or the consequences of an accident, or malfunction of equipment important.to safety as previously evaluated in the Final Safety Analysis Report is not increased because the basic' e method of testing the RCIC system remains unchanged. New-steps are added to require quarterly testing'of the trip and throttle valve.

This should fmprove overall system reliability,<therefore, the probability of an occurrence or malfunction of equipment is'not increased.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because changes to the valve line-up ensure normal system configuration at all times (e.g. 48 valve open and 49 valve closed)..

Testing of the trip and throttle valve will not affect system availa-bility, therefore, the possibility'for'an accident han not been created.

3. The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because procedures fulfill all operability requirements of Technical Specifications.

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i..;,e o-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 section 6.6.B.l. and 6.6.B.2..

of the Technical. Specifications.

UNIT 1

r. Licensee Event Report Number Date. Title of Occurrence

.90-04 3-10-90 Generator Trip / Reactor. ,

Scram 90-05. 3-13-90 RCIC 1301-61 (Steam Inlet:

. Valve) Would Not Operate -

RCIC Inoperable 90-06 3-13-90 Tornado Touchdown

.On Site-90-07 3-18-90 Reactor Vent and-Control-Room Vent Isolation 90-08 3-21-90 Fire Loading' Exceeds

Appendix.R - 0utside Design' Basis UNIT 2 90-05 3-19-90 . Group'11' Isolation Reactor Building Vent!

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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 Reductionst

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4. krasons For Kestriction (If any):

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5. husber of Hours Rea: tor has (ritical 676.8= " :.2045;6i t !26209.l4l d ,

6. Reactor Reserve thatdown Hours 0.0 -; ' f l 0soi , .! 3421.9 : ( f-

7. hours 6enerator Dn lint = .662.9- 52021.1f 1122111.76 ~ ,w' B. Unit Reserve Shutdown Hours .

70.0 .,

, . .; 0.01 3 ; 909.2 %

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9. 6ressTheraalEnergyGenerated(Mhl- 1504767.0" 1477j19$.01 L260453$46.03

10. 6ress Elettrical toergy Generated (Wh) ; , 52373tAl -1566750;04 84420307.0 1 , ,

11. hit Electrical Inergy 6enerated TMhl: ;504653.0) ' (1504)40.03 M 79345226.0 f ,

12.ReactorServiceFactor. - 91.'0. : 94.7' : 10.12 ,

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13.Rea:torAvailabilityFactor'- : 91.0F , .I94 72 ' , 82.3 4 '

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-14. Unit Service Factor - ; 9911Te '93.61 t I 77.5 i 15.-UnitAvailabilityFatter '189.11 193.6; l

[76.11 ^ '-

16. Unit Captity Fatter 10 sing MDC) .

83.2 =- 90.63 65.5" m

" Unit Captity Fatter (Ostra Design He)-

. 196.0j ' 1 8R.32 < ~ 63.If - ,

1 16.UnitForcedOutageRate i 110.9 ? + 13;91 -

- 5,41 -

~y ,, n it. Shutdeans Scheduled Over hert 6 Months (Type, Date, Land Duration of techh? ,

--j-20.'11 Shut Doun at End of Report Period. Estiaated Date of Startup: 1, - ,_

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,APPENDl!t=

OFERAtlN6DATAREPORii.

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DocietNo.!50-265 L Unit: Two? ~

,lateJ April 6, 1990t .

Cospleted Py 'Lynne Deellnyder!

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1. Reporting Period 1421 111190 GrcssHours.inReccetFerlcdt144 , :c s >

2. Currently Authorized Pewer Level 1MWti lill Man.1 Depend. Capacity _ (MWe-Wetli M Design tiettrical Rating (MWe het): 'M

3. Fener tevel te Which Restricted (11 Any) (MWe-hetit M - ,

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'4 Reasons For Restriction lli any):

THISMCNTHI N TO Daft $ ! COMULAtlVE!

5. hunter of Hours Reactor Was Critical . 0. 0 . ~ 771' 2 ' 120!!5.8

6. Featter Reserve Shutdeun Hours' 10.0 -

0.0 2985.8 '

7. Hours 6enerater On Line. 0.02 1757.2; '116855.73

8. Unit Reserve Shutdown Hours .

0.04 :n : 0.0 ' . '702.9:

9. Gross iternal Energy Generated thwh) 0.0 ? 1512552.0~ 1250909169.0 10.6resstiettricalEnergyBecerate((MWh) LO 0m 490629.0- , 80429713.0?

!!. Net tiettrical. Energy Generated (MWhl -5581.0 ; ; 458148.0 ; > 75937797.0 "

'!?. Restter Service Factor' O.0 E ~ 35.71 ' 77.0 -

13. Reactor Availatility. Factor 0. 0 ' < ;35.7- 79.0-

14. Unit Service Factor 0.0 ; 35.1 - 74.9:

15. Unit Availatility Factor 0.0 ; l 35.1: 75.41

16. Unit Capacity Factor (Using PDC) . -l.02 5 27.6 i-63.31

17. Unit Capacity Factor (Using Dest;n Muel -1.0 L 1 26.9? , , = 61.7 3

18. Unit Forced Dutage Rate 0.0 J 0.01 7 8.1.

19. Shutdeans $thetuted Over Next 6 Months (Type, Date, and Duration of. Eachl .

20. If Shut Denn at'End of Report Fertoo.- Esticated Date of Startups ; __

21. Units in fest Status Grier to Coseercial Operationit Forecast -Achieved' InitialCrititslity. _

initial Electricity __, J tessertialOperation _

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APPEND 118

' AVERAGE tiAILY t*li FDWER LEVEL ., - c.

, s Do:tet No. : 50-254- ~

iUnit One. w

,jj;,

, .-, .'DateiApril6(1990E

' Cospleted Pyi Lynne Leelsnyder4 Telephone'309654-22411 ,

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' MONTH ' P,A,XB :c

i DAYAVERA5EDAILYPOWERLEVEL 'LAYAVERA6iDAILYPOWERLEVEL1 <

(hWe Net) f(MWeNet)- ,

1 795 ;17 769. , .

2 783 . .18 , 757e .

3 757 19' - 784 7 ,

4 763' '201 i 7861 <

.5 - 763 - '21; '786{ ,

cQ 6 766' s. -- 221, e767" ~

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-7 796 ~23 I 782 -

8 . 777; '24 :4 162 i '

9= 768 ' 25 :- 74th 10- 33- - 26 77B ?

11: -5 : 27 1777; 12 - .6 .26, 1 780 13 158' 29 < 1788 s 14 . 623 : 30; .799 . . . ,

!$ - 792 31; i771; ,

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. a On this fort, list the average daily unit poser. level in Nhe-Wet for e'a th dy in the reporting aanthP:

Costste to the cratest .whole segasatt.

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These figures mill be used to plot a graph for ea:h reporting sonthf Nets that'ahe;. saximus'decendable "

caps:ity is ssed for the net.lelettrical rating of.the; unit,' there say be etcasions when the daily average; pewer level exceeds the 1001'llne for the restricted peser level linef. ~ln such cases, the averige daily; unit pener output sheet srculd be footnoted to explain the apparent anosaly.s

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AVERA6E DAILY DWlf POWER LEVEL '4 Dotlethe.r30-265 -

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.+

l These figures will be tied to plot a graph for ea:h reperting sonth.5-Note. thatShen _sarisua dependab'le M

- capacity is 'uset for the het electrical rating of._ the unit,_ there say be occasions'when the _dsfly average; j

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VI. UNIQUE' REPORTING REQUIREMENTS 1he1following-items are included in _this report based. on prior: commitments: to '

the commission:

A. Main Steam Relief-Valve-Operations Relief' valve' operations during the reporting period-are:gummarizedlini thet following table. The table includes information;as to_which relief: valve.

was actuated, how it was. actuated and the circumstances resulting in its; actuation.

E Units one Date March 13,=1990 ,, ,

Valves Actuated No. 6-Type of Actuation 1-203-3A 1 Manual

= 1-203-3B- 1 Manual --

1-203-3C ' l Manual:

1-203-3D' l Manual.

1-203-3E 'l Manual .

Plant Conditions: Reactor Pressure. 5924 ,

Description of Events: Routine surveillance and Post Maintenance: Test:

using:the Semi-Annual', ManualL0peration of-Electromatic Relief 1Valvesf(QOSL201-SI):L:

Tech'Spect Ref. 3.5/4.5.D.l.a B. Control Rod: Drive Scram Timing Data--for Units One and'Two There.vas no Control Rod' Drive ScramiTiming Data for. Units One and Two for the reporting _ period.

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.VII.y REFUELING'INFORMATIONi The fol. lowing information about future: reloads.at Quad-Cities Station-was--

requested in a January 26. 1978,ilicensing-memorandum (78-24) from D.-E.

-O'Brien to C. Reed,:et al.,.titied "Dresden,-Quad-Cities,<and Zion Station--NRC Request.for Refueling:Information", dated January-18,'19.78..-

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L '0027H/00612 .,..

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QTP 300-532 Revision 2 '

QUA0 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: ,

r NOT AS YET DETERMINED. 'I 5.

Scheduled date(s) for submitting proposed licensina action and

! supporting information: +

JULY 6, 1990 6.

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

' or different fuel design or supplier, unreviewed design or performance analysts methods, significant changes in fuel design, new operating procedures:

NONE AT PRESENT TIME.

b

7. The number of fuel assemblies.

4. Number of assemblies in core: 'n4

b. Number of assemblies in spent fuel pool: 1537

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: 3657 j

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 i

L ,

APPROVED ,

(fi"*"

l 1

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

L-__ . . - .- - -.. .. --

I

' QTP 300-S32 l Revision 2 .

QUAD CITIES REFUELING October 1989 INFORMATION REQUE5T ,

1. Unit: 02 Reload: 4 Cycle: 10

2. Scheduled date for next refueling shutdown: 2-3-90

3. Scheduled date for restart following refueling: 4-21-90

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

NOT AS YET DETERMINED.

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

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.

t

7. The number of fuel assemblies,

a. Number of assemblies in core: 0

b. Number of assemblies in spent fuel pool: 2735

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: o i

9. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2008 APPROVED-I l'4/0395t C).C.0,3.R.

'i

a

.,.-s.- .

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

ACAD/ CAM -

Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring ANSI -

American National Standards-Institute APRM -

Average Power Range Monitor ATHS - Anticipated Transient Without Scram BWR - Bolling Water Reactor CRD -

Control Rod Drive EHC -

Electro-Hydraulic Control System EOF -

Emergency Operations facility GSEP -

Generating Stations. Emergency Plan HEPA. -