ML19340B508

Monthly Operating Repts for Oct 1980

ML19340B508
Person / Time
Site:	Quad Cities
Issue date:	11/01/1980
From:	Tubbs R COMMONWEALTH EDISON CO.
To:
Shared Package
ML19340B505	List: ML19340B504 ML19340B508
References
NUDOCS 8011110193
Download: ML19340B508 (26)

Text

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e QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT OCTOBER 1980 COMMONWEALTH EDISON COMPANY AND 10WA-lLLIN0lS GAS & ELECTRIC COMPANY NRC DOCKET NOS. 50-254 AND 50-265 LICENSE NOS. DPR-29 AND DPR-30

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TABLE OF CONTENTS 1.

Introduction 11.

Summary of Operating Experience A.

Unit One B.

Unit Two I!I.

Plant or Procedure Changes, Tests, Experiments, and Safety Related Maintenance A.

Amendments to Facility License or Technical Specification B.

Facility or Procedure Cnanges Requiring NRC Approval C.

Tests and Experiments Requiring NRC Approval D.

Corrective Maintenance of Safety Related Equipment IV.

Licensee Event Reports V.

Data' Tabulations VI.

Unique Reporting Requirements A.

Main Steam Relief Valve Operations B.

Control Rod Drive Scram Timing Data i

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Refueling information Vi!!. Glossary b

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INTRODUCTION quad-Cities Nuclear Power Station is composed of two Boiling Water Reactors, each with a Maximum Dependable Capacity of 769 MWe net, located in Cordova, Illinois. The Station is jointly owned by Commonwealth Edison Company and Iowa-Illinois Gas & Electric Company. The Nuclear Steam Supply Systems are General Electric Company Boiling Water Reactors. The Architect / Engineer was Sargent & Lundy, incorporated and the primary construction contractor was United Engineerr - & Constructors. The con-denser cooling method is a closed-cycle spray ce,al, and the Mississippi River is the condenser cooling water source. The plant is subject to license numbers DPR-29 and DPR-30, issued Octcber 1, 1971, and March 21, 4

1972, respectively, pursuant to Docket Numbers 50-254 and 50-265 The

+

i date of initial reactor criticalities for Units 1 and 2 respectively were October 18, 1971, and April 26, 1972.

Commercial generation of power 4

began on February 18, 1973 for Unit I and March 10,1973 for Uni t 2.

This report was compiled by Becky Brown and Robert Tubbs, telephone number 309-654-2241, extensions 245 and 174.

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SUMMARY

OF OPERATING EXPERIENCE A.

UNIT ONE October 1-31: Unit One remained shutdown for the entire reporting period for End of Cycle Five Refueling Outage.

5.

UNIT TWO October 1-4: Unit Two began the reporting period holding a load of 777 MWe. Load was held until 1730 on October 2, when a Steam Jet Air Ejector suction valve was found to be enly cracked open. Upon opening the valve load rose from 788 to 822 MWe.

Load was held at this level for the rest of this four day period.

October 5-8: At 0001 on October 5, load was reduced at 100 MWe/ hour to 400 MWe to time MSIVs.

The reactor was manually scrammed at 1326 due to a fire in the drywell. The fire was caused by oil, leaking from the speed adjusting valve on this MSIV, flashing upon contact with the hot valve body.

On October 6, at 0543, the reactor was brought critical and at 1032 the generator was brought on line.

Load was increased at various rates until a load of 800 MWe was held at 2200 on October 8.

October 9-13: Over this five day period, load was held at an average 1

of 815 Mwe, utnli 2330 on October 13 when a drop at 100 MWe/ hour was begun.

j October 14-16:

Load was reduced at 100 MWe/ hour until a load of 600 MWe was reached. A defective steam tunnel PCI temperature sensor was replaced and load was increased at 1530.

Load was increased at 50 MWe/ hour for three hours, then at 5 MWe/ hour until a load of 815 MWe was reached at 031,0 on October 16.

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October 17-24:

In preparation for a scheduled outage, load was l

reduced starting at 1200 at a rate of 100 MWe/ hour until the generator was tripped off line at 2043 on October 17 The reactor was then manually thutdown on October 17 at 0216 and remained subcritical until 0643 on October 21.

This three day shutdown was used to perform Unit One Battery Tests, install TMI-modification related piping, and other miscellaneous maintenance items. The generator was put on line at 0729 on October 22 and brought through various rates of increase until a load of 726 MWe was reached at 0325 on October 24.

October 25-27: At 0000 on October 25, load was reduced at 100 MWe/ hour 4

to 500 MWe to reverse condenser flow and perform control rod moves. At 0520, load was then ' increased at 5 MWe/ hour until 0855 on October 27, when a load of 830 MWe was held.

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October _28-31: The Unit held an average load of 812 MWe over this four day period, and ended the reporting period holding a load of 811 MWe.

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Ill. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFE 1Y RELATED MAINTENANCE A.

Amendments to Facility License or Technical Specifications There were no amendments to 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 experiments requiring NRC approval for the reporting period.

D.

Corrective Maintenance of Safety Related Equipment The following represents a tabular summary of the safety related maintenance performed on Unit One and Unit Two

- during the reporting period. The headings indicated in this summary include Work Request Numbers, LER Numbers, Components, Cause of Malfunctions, Results and Effects on Safe Operation, and Action Taken to Prevent Repetition.

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UNIT ONE MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION QO3960 Unit 1 Refuel Cables required Replaced festoon Installed festooning Bridge (833) replacing.

cables.

system on main grapple and on monoraII.

Per-formed refuel interlock test.

Q07952 Unit 1 Diesel Faulty heat Bad coolant water Replaced heat exchangers Gene ra tor exchangers, leak (8" in 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />).

with like-for-like and tested.

.Q04558 Unit 1 Diesel Faulty oil seal.

Lube oil scavenging Replaced oil seal and Generator pump leaks oil around tested.

1-6601 shaft.

Q07911 Unit 1 Diesel Movable contacts Pump failed to auto-Replaced movable contacts Generator Cooling faulty in control start on initiation 7-77 with contacts from Water Pump switch.

from Control Room.

5-57 and tested.

1-3903 QO2847 Unit i Diesel Tubes needed Tube leak in heat Plugged 1 tube on South Generator

repair, exchanger.

H-X and 4 tubes in Cooling Water North H-X and tested.

Heat Exchanger 1-6601 408261 80-26/03L 1/2 Diesel The motor in-Pump motor burned.

Hooked up temporary Generator sulation had motor, and tested.

1 1 Cooling Water degraded.

Pump 1/2-3903 Q07411 80-20/03L Electromatic The pilot valve Valve would not close Repaired pilot valve Relief I-203-3B disc and seat after opening for and tested.

were severely testing.

steam cut.

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UNIT ONE MAINTENANCE

SUMMARY

I CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERnTION PREVENT REPETITION Q07929 Unit 1 Diesel Needed cleaning and During surveillance, Cleaned and inspected Generator Day switch inspection.

tank level at 75%;

switches.

Tank 1-6600 100% is usual.

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1 UNIT TWO MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q08444 LPRM 32-33 Unit 2 Vessel mounting nut Water was leaking past Torque changed from 50' Reactor was loose, sealing surface.

pound to 100' pound.

Qo3595 Scram Pilot Valve requ' Ired Rebuild 31 scram pilot Rebuilt and tested 31 Valves 2-302-rebuilding.

valves (117 and 118).

scram pilot valves.

117 & 118 Q06399 Protective Relays NRC Commitment Calibrate & test RPS Calibrated and tested for RPS MG Sets MG 2A & 2B for over-sets 2A and 2B.

2A & 2B (8000) voltage & undervoltage

& under-frequency relays.

Q07736 80-21/03L 2A Core Spray Faulty relay 3A valve would not Replaced auxiliary Suction Valve contact in motor open after closing contacts on the closed 2-1402-3A controller.

during operability contactor, test.

Q08225 80-23/03L Inboard RHR Limits were out Valve would only open Reset limits so valve inj. Shutoff of adjustment.

far enough to give would ride on the by-to A Recirc dual indication.

pass longer to get off Loop 2-1001-29A of the seat.

Tested.

Q08109 MSIV A0-2-203-2B Bad limit switch.

Did not get full Replaced limit switch closed signal,during and the seal tight.

timing surveillance.

Tested.

Q061Q2 IRM Ch 11 Faulty connector.

Spikes high.

Cleaned connector and remated.

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UNIT TWO MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NbMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q08226 80-24/03L Suppression Limits were Valve would only open Adjusted limit switches Chamber Test 2 out of adjustment.

far enough to give and torque ~ switch.

Spray 2-1001-34A dual indication.

Cleaned C.S. In Control Room. Tested.

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Q08493 Electromatic Limit switch arm.

Relief valve had open Bent arm on' limit switch.

Relief Valve Indication.

Tested.

2-203-3C Q08427 80-25/03L RCIC Discharge The limit switch Did not go full open.

Re-adjusted 'both the Valve M0-2-1301-49 was out of adjust-closed and open limits, ment.

Tested.

Q08453

'B' Line Feed-Faulty seal ring.

Valve leaks through Replaced seal ring.

Water Check seal ring.

Valve 2-220-62B Q08450 SRM 22 Detector Bent detector.

The detector would

. Replaced detector S/N not move in or out of 6, 615, 202; made up the core.

connector.

Q07398 HPCI 2-2340-1 Faulty controller The MGU would not come Replaced controller and MOS-FET.

off of.the HES when and MOS-FET 3Q5 Tested.

running surveillance.

Q07649

.Feedwater Check Faulty seal ring.

Valve blowing steam Replaced seal ring; Valve 2-220-62A through holes on the tested.

bull ring.

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i Q05429 SRM Ch 21 Faulty detector.

Readings erratic when Replaced detector; moving noisy detector, tested.

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UNIT TWO t:AINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION T/; D: iD llu: BER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT RE:3Ti-~!C!i Q08428 80-26/03L Electromatic Excessive steam Valve failed to open Replaced valve with Relief 2-203-3C leakage past the during surveillance, a spare overhauled valve retainer valve.

plug.

Q07401 RCIC M0-2-1301-Old torque switch Valve has dual in-Replaced torque 16 had crack in dication when open.

switch; tested.

mounting.

Q08633 Reactor Building The solenoid pilot isolation damper / valve Repaired valves Vent isolation valve was binding.

would not close on like-for-like parts; Damper / Valve isolation signal.

tested.

2-5742A Q08619 80-27/01T 2C RHR Pump A limit switch to Pump will not start; Reset limits; tested.

2-1001-2C the pump Interlock (auto trips),

was out of adjust-ment.

Q084ji CRD 2-38-51 Unknown; will over-CRD would not insert Removed CRD S/N-1289 (K-13) haul & Inspect on to "00".

and replaced with new work request.

CRD S/N-186.

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IV.

LICENSEE EVENT REPORTS The following is a tabular summary of all license 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.I. and 6.6.B.2. of the Technical Specifications.

UNIT ONE Licensee Event Report Number Date Title of Occurrence 4

80-26/03L 10-11-80 1/2 Diesel Generator Cooling Water Motor Trip 80-27/0lT 10-15-80 EDS Nuclear I & E 79-14 Seismically Unqualified Systems 80-28/03L 10-23-80 1/2 Emergency Diesel Generator Cooling Water Pump Covered with Water UNIT TWO 80-22/03L 10-05-80 MSIV Closure Time Less Than Three Seconds 80-23/03L 10-10-80 RHR Inboard Isolation Valve M0-2-1001-29A Failed to Open 80-24/03L 10-10-80 RHR Suppression Chamber Test and Spray Valve M0-2-1001-34A Failed to Open 80-25/03L 10-17-80 RCIC Check Valve MO 1301-49 would not Fully Open 80-26/03L 10-17-80 Electromatic 2-203-3C Failed to Open

Y IV.

LICENSEE EVENT REPORTS (continued)

UNIT TWO 80-27/01T 10-24-80 2C RHR Pump would not start i

80-28/03L 10-24-80

.. Wrong surveillance for 1/2 Diesel Generator Inop i

80-29/03L 10-27-80 Vacuum Breaker 2-1622A instrument Setpoint-Drift 80-30/03L 10-31-80 Slow Cicsure of Valve 2-1601-63 4

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DATA TABULATIONS i:.

The following data tabulations are presented' in this report.

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i A. -Operating Data' Report

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Average Daily Unit Power Level i-C.

Unit. Shutdowns and Power-Reductions 3

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j IMAGE EVALUATION NNNN TEST TARGET (MT-3) 1.0 EDM BA

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' OPERATING DATA REPORT DOCKET NO.

50-254

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UNIT ONE DATE November 1, 1980

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COMPLETED BY R. C. Tubbs

- _. _. TELEPHONE 309-654-2241, ext. 174 OPERATING STATUS

.0000 100180 1.

Reporting period:2400 103180 Gross hours in reporting period:

745 2.

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

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

4.

Reasons for restriction (if any):

This Month Yr.to Date Cunulative 5.

Number of hours reactor was critical 0.0 5725.1 60439.2 6.

Reactor reserve shutdown hours 0.0 0.0 3421.9

7. Hours generator 'on line

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0.0 5623.i 57664.0 8.

Unit reserve shutdown hours.

0.0 0.0 909.2 9.

Gross thernal energy generated (MWH) 0 11394718 115938699

10. Gross electrical energy generated (MWH) 53 3662811 37285865

11. Net electrical energy generated (MWH)

-9563 3360883 34776419

12. Reactor service factor 0.0 78.2 81.3

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13. Reactor avo11obility factor O.0 78.2 85.9

14. Unit service factor.

0.0 76.0

-77.6

15. Unit availability factor 0.0 76.8 78.8

16. Unit copocity factor (Using Des.MWe)

-i.7 59.7 60.9

_.17. Unit copocity factor (using MDC)

-i.6 58.2.

59.3

18. Unit forced outage rate 0.0 3.6 7.8 i

19. Shutdowns scheduled over next 6 nonths (Type,Date,and Duration of each):

20. If shutdown at end of report period,estinated date o f s t a r t u p _12_-4-80_ ______

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• The MDC nay be lower than 769 We dering periods of high unblani temperature due to tne thernal perfornance of the spray canal.

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OPERATING DATA REPORT DOCKET NO.

50-265 UNIT TWO D ATE November 1,1980 COMPLETED BY R. C. Tubbs TELEPHONE 309-654-2241. ext. 174 OPERATING STATUS 0000 100180 1.

Reporting period:2400 103180 Gross hours in reporting period:

745 2.

Currently authorized power level (MWt): 2511 Max. Depend capacity (MWe-Net): 769* Design electrical roting (MWe-Net): 789 3.

Power level to which restricted (if.ony)(MWe-Net): NA 4.

Reasons f or restriction (if any):

This Month Yr.to Date Cumulative

.5.

Number of hours reactor.was critical-652.3 4433.8 57562.7 6.

Reactor reserve shutdown hours 0.0 0.0 2985.8 7.

Hours generator on line 617.1 4279.1 55070.9 8.

Unit reserve shutdown hours.

0.0 0.0 702.9 9.

Gross thernal energy generated (MWH) 1367840 9546745 112562854

10. Gross electrical energy generated (MWH) 443666 3000736 35863786

. 11. Net electrical energy. generated (MWM) 421353 2817269 33559794

12. Reactor service factor 87,6 o0.6 78.4

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13. Reactor avo11obility factor 87.6 60.6 82.5

.14. Unit service factor 82.8 58.5 75.0

15. Unit ovallobility factor 82.8 58.5 76.0

16. Unit capacity factor (Using Des.MWe) 73.5 50.0 59.5 t

-17. Unit capacity'foctor (Using MDC) 71.7 48.8 58.0

18. Unit-forced autoge rate 3.3 5.4 9.0

19. Shutdowns scheduled over next 6 nonths (Type,Date,and Duration of each):

. 20..If shutdown at end of report period,estino'ted date of stortup NA

$The MDC nay be lower than 769 MWe during periods of high anhiant temperature due to the thernal perfernance si the spray canal.

APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-254 UNIT ONE DATE Nove.aber 1,1980 COMPLETED BY R. C. Tubbs TELEPHONE 309-654-2241, ext. 174 HONTH October-1980 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 1.

-17.2 17.

-15.1 2.

-16.6 18.

-5.1 3.

-18.0 19.

-2.7 4.

-17.9 20.

-2.7 5.

-16.5 21.

-6.6 6.

-16.8 22.

-17.4 7.

-15.8 23.

-15.4 8.

-16.1 24.

-15.9 9.

-17.2 25.

-16.2 10.

-18.8 26.

-17.5 ti.

-15.8 27.

-6.5 12.

-15.6

-28.

-2.4 L3,

-17.5 29.

-2.3 L4.

-15.7 30.

-3.0

-16.0 31.

-2.2

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-16.0 INSTRUCTIONS On this forn, list the average daily unit power level in MWe-Net for each day in the reporting nonth.Cepete to the nearest whole nec u tt.

These figures will te ssed to plat a graph for enth reporting month. Note that when marinen de;endable capacity is used for the net electrir-1 rating of the sait there nay be occasions when the daily overage pcwer level excetes the il6% line (or the restritted pcwer level line).,In soth cases,the average daily snit power output sheet shseld be festnoted to trplain the apparent ansnaly

-AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-265 UNIT TWO

.DATE November 1, 1980 COMPLETED BY, R. C. Tubbs

. TELEPHONE 309-654-2241, ext. 174 HONTH October 1980 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 1.

744.7 17, 520.6 2.

755.0 18.

-7.2

._. u 3.

792.0

_ 19.

-6.2 4,

797.8 20.

-6.0 5,

204.3 21.

-12.7 6..

201.8

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199.5 7.

536.5 23, 570.0 B.

688.7 24, 658.9 9.-

770.7

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532.1 10, 770.6 26.

677.4 ii.

776.9 27.

760.6 12.

792.5

- - 28,

756.3 13, 785.3 29.

763.9 14.

527.9 30.

806.5 l

15.

716.7 31, 707.5 16.

773.8 i

INSTRUCTIONS On this Fern, list the average dailt unit psver level in fWe-Net for ecch dat in the reporting nonth. Compute to the nearnt while ner awatt.

Then figeres wif'l be osed to plot a graph for ecch reporting nonth. Note that when notinen dependable capacity is used for the net electrical rating of the snit there may be occasions when the daily average power level exceeds the iHI line (or the restricted power level line).,In such uses,the average daily snit power output sheet should be festnoted to explain the apparent one %1y

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(' 3, APPENDIX D QTP 300-S13 UNIT SilVTDOWNS AND POWER REDUCTIONS Revision 5 DOCKET NO, 50-254 g

jg7g Qu d-Cities Unit One UNIT NAME p

November 1, 1980 OATE REPORT HONTil October 1980 TELEPHONE 309-654-2241, ext.

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8 LICENSEE yg gg

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DURATION EVENT y8-g: 8 (HOURS)

NO.

DATE u-y53 REPORT NO.

8 CORRECTIVE ACTIONS / COMMENTS R

13 800831 S

745.0 C

4 NA RC F 'JELXX Continuation of End of Cycle Five Refueling Outage 1

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(3 APPENDlX D QTP 300-S13 UNIT SilVTDOWNS AND POWER REDUCTIONS Revision 5 DOCKET NO.

50-265 March 1978 UtilT NAME quad-Cities Unit Tw COMPLETED BY R. C. Tubbs DATE November 1, 1980 REPORT HONTH October 1980 TELEPHONE 309-654-2241, ext.

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LICENSEE pg 8 g.

p{a" yo DURATION EVENT y8

@:8 (HOURS)

NO.

DATE u.

y! $ g REPORT NO.

8 CORRECTIVE ACTIONS /CCnHENTS R

15 801005 S

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5 80-22/03L CD VALVEX Power reduction to time Main Steam isolation Valves.

16 801005 F

21.1 H

2 CD VALVEX Reactor raanually scrammed due to fire caused by o!1 leaking from speed adjusting valve on MSIV, and flashing to fire when contacting hot valve body.

17 801014 F

0.0 A

5 SD INSTRU Power reduction to replace faulty temperature switch in Primary Containment Isolation circuit.

18 801017.

S 106.8 B

1 5cheduled outage to perform battery tests and miscellaneous maintenance items.

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3 VI.

UNIQUE REPORTING REQUIREMENTS The following 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 summarized in the following table. The table includes information as to which relief valve was actuated, how it was actuated, and the circumstances resulting in-its actuation.

VALVES NO. & TYPE PLANT UNIT DATE ACTUATED ACTUATIONS CONDITIONS DESCRIPTION OF EVENTS 2

10-17-80 2-302-3A 1 Manual Rx Press Surveillance T.S.

2-302-3B 1 Manual 980 4.5.D.l.b.

2-302 3C Failed To Open 2-302-3D 1 Manual 2-302-3E I Manual 2

10-21-80 2-302-3C 1 Manual Rx Press Test after failed to i

445 open on 10-17-80 (Replaced Valve)

B.

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

The following table is a complete summary of Units One and Two Control Rod Drive Scram Timing for the reporting period.

All scram timing was performed with reactor pressure greater than 800 psig.

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RESULTS OF SCRAM TIMING HEASUREMENTS PERFORMED ON UNIT 1 s 2 CONTROL ROD ORIVES, FROM l-1-80 TO 12-31-80 AVERAGE TIME IN SI$0NDS AT %

Max. 1seu INSERTED FROM FULLY WITHDRAWN For 90%

Insertion DESCRIPTION NUMBER 5

20 50 90 Technical Specification 3 3.C.1 s DATE OF RODS 0.375 0.900 2.00 3.5 7 sec.

3.3.C.2 (Average Scram insertion Tin 10-20 1

0 32 0 56 1.06 1.82 K-13 Unit Two Rod K-13 was replaced 1.82 during weekend outage

' Cold Scram Time 10-22 88 0.32 0 71 1.49 2.59 J-12 Seq A Hot Scram Timing Unit 2 2.87 e

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REFUELING INFORMATION' The following information about future reloads at quad-Cities Station

.was requested in a January 26, 1978, licensing memorandum (78-24). from

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D. E. O'Brien to C. Reed, et. al., titled "Dresden, quad-Cities, Zion Station - NRC request' for refueling information" dated January ~ 18, 1978.

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QTP 300-S32 Revision 1 QUAD-CITIES REFUELING March 1978 j(-

INFORMATION REQUEST 1.

Unit:

1 Reload:

6 Cycle:

7 2.

Scheduled date for next refueling shutdown:

9-12-82 (Shutdown E0C6) 3 Scheduled date for restart following refueling:

12-5-82 (Startup BOC7) 4.

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

No, Plan 10CFR50 59 reloads for future cycles of Quad Cities Unit 1.

The review will be conducted in August, 1982.

5 Schedoled date(s) for submitting proposed licensing action and supporting info'mation: August, 1982 for 10CFR50.59 related changes es 90 days prior to shutdown.

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:

New fuel designs:

7 The number of fuel as,semblies.

a.

Number of assemblies in core:

724 b.

N aber of assemblies in spent fuel pool:

820 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:

1460 b.

Planned increase in licensed storage:

None 9

The projected date of the last refueling that can be discharged to the spent fuel oool assuming the present licensed capacity: September, 1985 (end of batch discharge capability)

)

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

QTP 300-S32 Revision 1 QUAD-CITIES REFUELING March 1978 INFORMATION REQUEST 1.

Unit:

2 Reload:

5 Cycle:

6 2.

Scheduled date for next refueling shutdown:

8-30-81 (Shutdown EOC5) 3 Scheduled date for restart following refueling:

12-20-81 (Startup BOC6) 4.

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

No, Plan 10CFR50 59 Reloads for future cycles of quad Cities Unit 2.

The review will be conducted by early August, 1981.

5 Scheduled date(s) for submitting proposed licensing action and supporting Information: Early August, 1981 for 10CFR50 59 related changes a90 days prior to shutdown.

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 c,,arating procedures:

New Fuel Design:

1.

Barrier Fuel 2.

Control Cell Core

[-

flq 7

The number of fue' a,ssemblies, a.

Number of assemblies in core:

724 b.

Number of assemblies in spent fuel pool:

672 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:

1460 b.

Planned increase in licensed storage:

None 9

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

September, 1984 (End of batch discharge capability)

(p A.'P P R O V E D APR 2 01973 Q. c. c. S. R.

s.

Vill.

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

J CRD Control Rod Drive Systsm SBLC Standby Liquid Control System MSIV Main Steam Isolation Valve RH RS '

Residual Heat Removal System RCIC Reactor Core Isolation Cooling System HPCI High Pressure Coolant Injection System Source Range Monitor SRM Intermediate Range Monitor IRM Local Power Range Monitor LPRM Average Power Range Monitor APRM TIP Traveling incore Probe RBCCW Reactor Building Closed Cooling Water System TBCCW Turbine Building Closed Cooling Water System RWM Rod Worth Minimizer Standby Gas Treatment System SBGTS HEPA High-Efficiency Particulate Filter Reactor Protection System RPS IPCLRT Int. grated Primary Containment Leak Rate Test LPCI Low Pressure Coolant injection Mode of RHRS RBM Rod Block Monitor Boiling Water Reactor BWR ISI In-Service inspection MPC Maximum Permissible Concentration Primary Containment isolation PCI SDC Shutdown Cooling Mode of RHRS LLRT Local Leak Rate Testing MAPLHGR Maximum Average Planar Linear Heat Generation Rate RO Reportable Occurrence DW Drywell RX Reactor EHC Electro-Hydraulic Control System MCPR Minimum Critical Power Ratio PC10MR Preconditioning Interin Operating Management Recommendations i

LER Licensee Event Report ANSI American National Standards Institute NIOSH National Institute for Occupational Safety and Health ACAD/ CAM -

Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring l