ML20004F137

Monthly Operating Repts for May 1981

ML20004F137
Person / Time
Site:	Quad Cities
Issue date:	06/01/1981
From:	Tubbs R COMMONWEALTH EDISON CO.
To:
Shared Package
ML20004F134	List: ML20004F133 ML20004F137
References
NUDOCS 8106160543
Download: ML20004F137 (27)

Text

_. ~

i 1

QUAD-CITIEF NUCLEAR POWER STATION UNITS 1 AND 2 MON ~iHLY PERFORMANCE REPORT MAY 1981 COMMONWEALTH EDISON COMPANY AND ,

IOWA-ILLINOIS GAS F,ELECTP.lc COMPANY NRC DOCKET NOS. 50-254 AND 50-265 LICENSE NOS. DPR-29 AND DPR-30 i

8106160543

TABLE OF CONTENTS

1. Introduction

~

11. Summary of Operating Experience A. Unit One B. Unit Two lit. Plant or Procedure Changes, Tests, Experiments, and Safety Related Maintenance A. Amendments to Facility License or Technical Specifications B. Facility or Procedure Changes Requiring NRC Approval C. Tests and Experiments Requiring NRC Approval D. Corrective 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 Requiremer.ts

. A. Main Steam Relief Valve Operations B. Control Rod Drive Scram Timing Data.

Vll. Refueling Information Vill. Glossary m -- -- -

- ~ ~ . y - - - --

' )

l. 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 lowa-Illinois Gas & Electric Company. The Nucleat Steam The Supply Systems are General Electric Company Boiling Water Reactors.

Architect / Engineer was Sargent & Lundy, incorporated and the primary construction contractor was United Engineers & Constructors. The con-denser cooling method is a closed-cycle spray canal, and the Mississippi

' River is the condenser cooling water source.- The plant is subject to license numbers DPR-29 and DPR-30, issued October 1,1971, and March 21, The 1972, respectively, pursuant to Docket Numbers 50-254 and 50-265 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 I and March 10, 1973 for Unit 2.

This report was compiled by Becky Brown and Robert Tubbs, telephone

~

number 309-654-2241, extensions 245 and 174.

eM* - .__ _

e

• aw e

• Nm

• eg# e r-w-- -,,i+ .-

p y y g .w--e-, , - - . ,,-y _-.y- -

-py -..p---y ,,e---y,. .pp+.- -, ,

.---,,-.9-,yy-

. 1 1

1 1

• ~

II.

SUMMARY

OF OPERATING EXPERIENCE A. UNIT ONE May 1-9: Unit One began the reporting period holding a load of 815 MWe. Over these initial nine days, an average load of 814 MWe was l held, except on May 3 On that day load was dropped to 700 MWe for the weekly turbine test. After the tests were completed, load was held for an additional three hours, per the Load Olspatcher, resulting in a load of 768 MWe for the day.

May 10-12: On May 3 at 2330, load was dropped at 100 MWe/ hour to perform the weekly turbine tests. At 2335, an alarm for the 8 Recirc Pump Motor Lower Lube Oil HI/Lo Level was received. Seven Hundred HWe was reached and held at 0035 on May 10 and the tests were Load was held at 700 MWe until 0515 At that completed at 0145 time load was increased at various rates until 0800 when it was .

dropped back to 400 MWe at a rate of 100 MWe/ hour. The Drywell was deinerted and an entry was made to add oil to the Recirc Pump Motor Lower Bearing in both loops. At 1415 load was increased at 100 MWe/ hour for two-hours, then at 5 MWe/ hour until 0700 on May

12. The resulting daily average load for May 10,11, and 12 was 614, 735 end 809 MWe, respectively.

May 13-21: With the exception of May 17, load was held at an average

- of 809 MWe. On May 17 a load of 708 MWe was achieved due to per-formance of the weekly turbine tests and load reductions per the Load Dispatcher's_ request. .

May 22-27: Load was held at 812 MWe on May 22 until 1900, when it was dropped in preparation for a Maintenance Outage to repair a leaking seal ring on I A Feedwater Check Valve. The Unit was tripped off-line at 2337 and the reactor was manually scrammed at 0050 on May 23 Other work items performed during this outage were: IA and IB Recirc MG Sets were rebrushed, Pilot Valves were replaced in B, C, l', and E Electromatic Relief Valves, and miscellaneous valves were repaired.

The reactor was pulled critical at 2003 on May 24, and the generator was put on line at 0349 on May 25 Load was increased until 0500, when it was held at 230 MWe to perform scram timing on 88 control rods. At 1100 load was increased to 400 MWe and special rods maneuvers were performed for the Nuclear Engineer. Load was then increased to 500 MWe for a Xenon soak. At 1130 on May 26 load was increased at various rates until it was held at 2400 on May 27

--r

May 28-31: On May 28 load was held at 803 MWe. At 0145, on May 29, load was dropped to 750 MWe to switch reactor feed pumps and was then increased to 807 MWe at 092J. Load was held until 0430 on May 31.

At that time load was dropped to 700 MWe to perform the weekly turbine tests, however, the Load Dispatcher requested that the Unit drop to 650 MWe. Load was held until 0615 when it was then increased at various rates until 804 MWe was reached and held. The Unit ended the reporting period in that state.

B. UNIT TWO May 1-8: Unit Two began the reporting period holding a load of 783 MWe. Although load was held, with the exception of weekly turbine tests on May 2, the average load was 766 MWe. This gradual drop in maximum load is attributed to a limiting control rod pattern.

May 9-11: At 0000 on May 9, load was dropped to 600 MWe to perform rod moves 'for the Nuclear Engineer. Load was increased at 5 MWe/ hour unitt 2l00 on May 10. On May 10 the coastdown for End of Cycle Five began. At 2220 the 2E condensate demineralizer was taken out of service due to high post strainer D.P. This necessitated dropping load on May 11 to backwash and precoat 2G and 2F demineralizers.

The average daily load for May 9,10, and 11 was 648, 755, and 762 MWe.

May 12-15: Load was held over this four day period at an average of 761 MWe. Operational occurrences during this period included restoration of the 2E condei.; ate demineralizer on May 12.

May 16-20: On May 16 load was dropped for the weekly turbine tests, then held an additional four hours for the Load Dispatcher. Load

- was again dropped for the Load Dispatcher on May 17 for four and one hal f hours. The resulting daily average load was 729 MWe on May 16, and 705 MWe on May 17 On May 18,13, and 20 load was held at an average of 746 MWe.

May 21-22: On both of these days the Load Dispatcher requested load to be dropped for three and one half hours each day. The resulting daily average loads were 709 and 719 for May 21 and 22 respectively.

May 23-27: Lead was held for this five day period with the exception of the weekly turbine ?-sts on May 24. However, due to deratings, the tests had little etrect on load. The average load over this period was 718 MWe. ,

(

_ y

May 28-31: On May 28, at 0115, load was reduced to 550 MWe at the request of the Load Dispatcher. Load was later increased,average starting at 0600, at various rates until it was held at 1445 The However,

~ due load for the day was 666 MWe, Load was held on May 29 to all lift' pumps tripping off at 0930, and the plant going on full river operation, a large rise in power occurred due to the using of the cooler river water. The average load for the day was 712 .1We.

On May 30 at 0030, power was dropped to 500 MWe for the Nuclear Engineer. At this time the turbine weekly tests were performed and condenser flow was reversed. At 0335 load was increased at 5 MWe/ hour.

The load increase continued through May 31 and the Unit ended the reporting period a'; 745 MWe and increasing at 5 MWe/ hour.

O l

l l

f e.-, - - - - - - - -- e ,4r, - . - - , ,.,,-yw--.-, - -,,.my--r... r w w ... ,,,. ,,--., m--- - , , --

111. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A. Amendments to Facility License or Technical Specifications On April 16, 1981, Amendments 66 and 60 were issued to DPR-29 and DPR-30 respectively. These Amendments consist of changes in the Technical Specifications for each of the two units which change setpoints for certain system settings. These changed setpoints are for: 1) Turbine Condenser Low Vacuum Scram; 2) Main Steamline Low Pressure isolation; 3) Main Steamline High Flow' Isolation; 4) ECCS-ADS Interlock, and

5) ECCS Fill System High Pressure Alarm. These changes in instrument and system setpcint have been nade to reduce the number of nuisance alarms a,d spurious trips caused by set-point drift.

On April 20, 1981 Amendments 67 and 61 were issued to DPR-29 and DPR-30 respectively. These Amendments eliminate the requirement to reduce reactor power to below 50 percent of

' rated power when the main steam isolation valve closure time verification is performed.

On April 20, 1981, Amendments 68 and 62 were issued to DPR-29 and DPR-30 respectively. These Amendments eliminate the requirement for continuous monitoring of the primary contain-ment inerting system make-up as a means of monitoring the containment for gross leakage.

On April 24, 1981, Amendments 69 and 63 were issued' to DPR-29 and DPR-30 respectively. These Amendments remove reactor water cleanup isolation valve MO-1201-80 from Table 3 7-1 of the Technical Specifications and excludes the valve from the

-surveillance requirenent described in Section 4.7.D.

On May 13, 1981, Amendment 71 was issued to DPR-29 This Amendment extends the MAPLHGR curve for a mixed-oxide fuel bundle to 50,000 MWD /ST planar average exposure. This w!Il

[

enable the completion of a high burnup fuel experiment in the present core.

B. Facility or Procedure Changes Requiring NRC Approval l

l There were no Facility or Procedure Changes Requiring NRC approval for the reporting period. ,

f 1

i

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

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

C. Tests and Experiments Requiring NRC Appreval There were no Tests and 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 Hal functions, Results and Effects on Safe Operation, and Action Taken to Prevent Repetition.

i

.w am

• e *** e =* e

• W M* M N - - - - -- ,- _ - ,'v ~',.

UNIT ONE MAINTENANCE

SUMMARY

J CAUSE RESULTS & EFFECTS .

OF ON ACTION TAKEN TO U.R. IER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION NUMBER NUMBER

' 1/2 Diesel Gener- The field flash The field ground alarm Lubricated the field Q12240-ator relay was sticking came up. Diesel Gener- flash contactor; closed, ator operability _was tested operable, not affected.

l-1001-65C The pump internals The pump had high Replaced shaft Q12159 sleeves, bearings, IC RHR Service were worn. bearing vibration, Water Pump The other three RHR seal and balanced 4

Seivice Water Pumps impeller.

were operable.

,]

IB RHR Pump Me,ch5nical seal on Water was leaking Char.3ed cartridge--

j Q12192 type mechanical seal.

pump shaft was from the seal. The pump was operable.

4 worn.

lA-1002 RHR Seal was worn. Water was leaking Replaced mechanical ~

Q12191 seal and broken stud.

Pump from the seal. The pump was operable, 10-1002 RHR Seal was worn. Water was leaking Replaced mechanical Q12193 seal.

Pump from the seal. The pump was operable.

l-302-19A Wire had pulled The solenoid did not Found wire on coil f Ql264b energize when voltage pulled out of wire nut.

! . Back-up Scram loose from the Solenold Valve coll leads. was applied to it. Reconnected lead with during initial testing. wire nut.

The RPS system was fully operable.

}

}

i i

UNii ONE MAINTENANCE

SUMMARY

4 CAUSE RESULTS 6 EFFECTS- ,

OF ON ACTION TAKEN TO W.R. LER PREVENT REPETITION COMPONENT MALFIINCTION SAFE OPERATION NUMBER NUMBER 1-220-62A Feed- Steam leak in Steam was leaking in Replaced pressure seal Ql2518 the MSlV room. No ring and installed new water Check Valve pressure seal ring.

abnormal releasa. of " tie" wire on hold down radioactive material to bolts for trim.

the environs occurred.

Control circuit Old cable had been Replaced cable from Q11606 M0-1-202-5A operator to 3rywell cable has shorted spliced as a temporary wires--5 conduc- fix. The valve was penetration--verified i

tor cable #12507 fully operable, pump trips on valve closure.

The auxiliary The valve would not Replaced auxiliary 1-1001-19A RHR Q11421 System Crossrie contacts were close from the control contacts on circuit Valve sticky, switch. Failure was breaker.

In the safe direction; LPCI operability was not af fected.

1-2001-16 The air cylinder The valve would not Replaced air cylinder l Q12651 was worn causing open completely. The on valve operator.

isolation capability air to leak by the

'. piston. was not affected.

t e

b UNIT TWO MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS OF ON ACTION TAKEN TO W.R. LER PREVENT REPETITION MALFUNCTION SAFE OPERATION ___

NUl'.BER NUMBER COMPONENT l

A0-2-203-2A--2A There was a steam Failed to get 1/2 scram initially the 10 percent Qll701 81-07/03L during testing. The closed switch was Outboard MSIV leak in the vicin- exercised and it worked ity of the limit other HSIVs functioned as designed, satisfactorily. Then switch from the during the short RCIC testable outage, the limit check valve. switch assembly was changed.

IRM was failed down- Replaced detector and IRM Number 17 The detector was checked connectors.

Q10879 failed. scale. The other

, IRMs in that channel

• were operable.

Calibrated pressure 2-1001-18A The pressure The valve did not Q12314 switch was out of automatically open switch PS-I-1001-81A.

RHR Hinimum Flow Valve calibration. during the flow test.

LPCI was operable.

,I I The valve did not Calibrated pressure 2-1001-18B The pressure Q12313 automatically open .ch PS-1-1001-818.

RHR Minimum swtich was out of

"., ] Flow Valve calibration. during the flow test.

LPCI was operable.

?

1 Water was leaking from Changed mechanical seal.

2A RHR Pump The mechanical Q12209 seal was worn. the seal. The pump j was operable.

i Water was leaking f rom Changed mechanical seal, 28 RHR Pump The mechanical Q12210 seal was worn. the seal. The pump was operable.

4 4

I i

UNIT WO HAINTENANCE

SUMMARY

.i 2

• CAUSE RESULTS & EFFECTS OF ON ACTION TAKEN TO W.R. LER MALFUNCTION SAFE OPERATION PREVENT REPETITION NUMBER NUMBER COMPONO!T Changed mechanical seal, 2C RHR Pump The mechanical seal Water was leaking from -

Q12211 the seal. The pump was l was worn.

operable.

A voltage control Lost voltage control to _ Replaced rectifier.

Q12546 Unit 2 Diesel Generator rectifier burnt the generator. The 1/2 l

Voltage out. Diesel Generator and-j Regulator

- associated ECCS components 1

and containment cooling -

i mode of RHR were .

demonstrated operable.

Two off-site lines capable of supplying 345 KV power were available.

~The circuit The valve would not open Replaced auxiliary Q12549 81-12/03L MO-2-1001-78 contacts on circuit RHR Pump Suction breaker auxiliary , from the Control Room, l contacts were LPCI and containment breaker.

Valve

! sticking. cooling modes of RHRS

)

were still operable.

I Spurious breaker The 1/2 Diesel Generator Replaced 4 wires in l Q12519 81-10/03L Unit 2 Diesel was operable. Two off- cubicle 1, Bus 24-1.

3

~

Generator closure ccused the generator to site _ lines supplying Replaced wiring and i motorize, burning 345 KV power were potential transformers some wiring and available. The low in generator.

i potential trans- pressure ECCS and

{  ! formert. containment cooling

,i mode =, of RHR associated with the operable Diesel

' Generator were proven 4

operable.

}

i i

• i j i * *

)

i .

m

f 8 4

UNIT _TWO HAINTENANCE

SUMMARY

.{

I

' CAUSE RESULTS & EFFECTS I, ON ACTION TAKEN TO OF W.R. LER PREVENT REPETITION

} COMPONENT MALFUNCTION SAFE OPERATION

NUMBER NUMBER 2-2001-16 Drywell Air cylin, der was Valve would not open Replaced e'r cylinder Ql2635

' Equipment Drain worn, cauhing from the Control Room. on operator.

i Sump Discharge sluggish bperation. The valve failed on Valve the isolated condition.

The redundant Isolation

- valve was fully operable.

4 )

4!

1  !

h a

l -

j i

1 j e i

i i

i j  !

j b

3 i1 -

I 1

i <

= -

e IV. LICENSEE EVE!4T 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 Speci fications.

UNIT ONE Licensee Event Title of Occurrence Nurober Date There were no Licensee Event Reports for the reporting period for Unit One UNIT TWO 5-4-81 MO-2-1001-378 Breaker 81-9/03L tripped.

5-15-81 Unit Two olesel SI-10/03L Generator inoperable f -

5-18-81 M0-2-1001-7B RHRS 81-12/03L . Valve failed to open l

- . w e-.- .. www .me. _

n m. m

, -----.-,e ., ..._.m ,_ _.-_ ._ . , . ~ , ,. __--- - . . _ _ ..#,,-. ,,, . , , , , - , , - . , . , _ , , . . . . ,_,-.-_ ~ -y . . . - . , , , , , , , , . . , , - , _ ., . . . , , , ..

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 Reduct'ons

.-.n. . - - . - - -

. OPERATING DATA REPORT

~ ~ - ' ~ ~ ~ ~

~ ~

DOCKET NO. 50-254 UNIT ONE - - ~ - ~ -

DATEJune t 1981

~"'

COMPLETED BYRobert C'Tobb s TELEPHONE 309-654-2241Xi74_ -

OPERATING STATUS 0000 050181 744"-

1. ' Rep oFtiiig ~ period : 2400 053181~~GFoss hours in reporting period:

2. Currently authorized power leeel (MWr): 2511 Max. Depend capacity ^' ~ '- '~

~ ~ ~ 1 MWi--Nit ) i~~~769* Design ~e1~ec t r ical ~ra t ing (MWe-Net): 789

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

i<A _

4. Reasons for restriction (if any):

~

~ ~ ~ ~ - ~ ~ ' ~ ~

~

This Month Yr.to Date~~ Cumulative ~

700,8_ 3410.0 64117.1

5. Number of hours reactor was critical 0.0 0 . 0. 3421.9

6. Reactor reserve shutdown hours

~

691.8 3T47.8 "61231 5

7. Hours generator on line 0.0 0.0 909.2

8. Un i.t. reser ve shu td own ho urs.

1622874 7770335 124012423

9. Gross thernal energy generated (MWH) 530156 2554270~ 39933177 CO .~ Gross electrical energy generated (MWH)

~

492672. 2379914 37236193 ii. Het electrical energy generated (MWH) 94.2_ 94.1 20.8 12.' Reactor service factor

~

~

94.2 94 i_ ~ - ~~ 8 5'~ t

~ ~ ~' t.3 . Recctor avriilobility fac t or~

92.4 77,1 93.0_

14. Unit service factor 93.0 92.4 78.3 15.-Unit availability factor 86,i ~ 85.4 ~ ~~~ 61~0~

~

16.' Unit' capricit y 'f ac t or -(Using MDC)

~

83.9_ 83.2_ 59.4

17. Unit capacity Factor (Using Des.MWe) _ _ __

i.8 7,5 0 . 0_

10. Unit forced outage rate

19. Shu'tdowns scheduled over next 6 months (Typ.e ,Date ,and Duration of e~ach ) F

__,___NA

20. If shutdown at end of report period,estinated date o f s t ar t u p ,_____,

sihe !CC M7 he lower than 769 MWe during perids of high ebicat tegarature due is tne therol perfernnte er the sorcy ccul.

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

- OPERATING DATA REPORT

~~

DOCKET NO. 50-265 UNIT Tuo - - -

DATEJune i 1981 COMPLETED BYRobert C Tubbs TELEPHONE 309-654-9.241Xi74 - - - - -

CPERATING STATUS 0000 050181 ~

744

1. Reportlig'pdFiod:'2400 ~ 05318i.~G'r o sis h o u r #~'in r e p o r t in g p e r i o d :-

~

2. Currently authorized power level /.MWt): 2511 Max. Depend capacity

~ ~~ ~ ~ ~ ~

~~~~ ~

~

~

(hGe~he~t):~ ' ~769* Design" elaEtrical rc't ing '(MWe-Ne t ) : 789

~ ~

NA

. . . . 3' Power level to_which restricted (if any)(MWe-Net):

4. Reasons f or restriction (if any):

744.0. 3527.6 62360.4

5. Nunber of hours reactor was critical 0 . 0. 0.0. 2985.0

.6. Reactor reserve shutdown hours ~

~ ~ -

744.0. '3501.8 "59783TO~

Hours generator on line-~~

~

~

~ '7.

0 . 0_ 0.0 702.9 S. Unit reserve shutdown hours. , ,_

1701924 8138137_ 12333854_5_,

'9 . G[oss thermal energy generated (MWH)

~

- 543830 2591717_ 39313268-

~

~f0. GFos~s diectiical energy generatdd(MWH) 519000 2459614 36816566

11. Het electrical energy generated (MWH) 100.0 97.4 79.5

&2. Reactor service factor +

~"~

~~~

~ 8 3','3_~

100.O ~ ' '97 i4

~ ~i3'. Re5~ctor cvsil'ab11'ity' factor ~ ~~

~~~ ~

100.0_ 96.7_ 76.2 14 . Unit _ service factor. ,,

100.0 96.7. 77.1

5. Unit availability factor

~~ ~ 6170' 90.7 88.3

16. Unit capacity factor (Using MDC) 88.4 86.0. 59.5

17. Unit capacity factor (Using Des.Mue) 0.0 1.4 8.7

1. 3 . Unit forced outage rate

' ~ ~ 19l~ShuYdowns' scheduled over next ~6~ nonths (Type,Date,and Durntion of each):-

20. If shutdown at end of report period,estinated date of star tup __NA _ _ _ _ _ _ , , _

Ce PDC sy be lower then 76? IfJe during periods of high anbient tercerature due u tr.a t$trM1 perforncace of the spray centi. , _

. , _ _ . . ... ,,7

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

APPENDIX B AVERAGE DAILY UNIT POWER LEVEL

' DOCKET NO. 50-254

' ~~

UNIT ONE DATEJune i 1981 -~~~

COMPLETED BYRobert C Tobbs p

MONTH Mau 1981 .

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

(MWe-Net)

17. 659.0

i. 762.0

' ' 18. ~~ 754.6

. 757.'8

17. 7U6.7 3., 717.5 _

20. 756.0

4. < 753.6 764,4 756.0 21.

'5 .

22. 666.8

6. 765.0

23. -31.6

7. 760.2

24. -26.6

9. _ _ _

765.0

25. 269.0

9. 762.1

26. 497.4 to. 567.4

~

~~

27. 667.8

~ti. 602.4 ~

28, 772.9

. _12 , 754.5

29. 732.4 13, 760.7

30. 747.3

- 14 .

754.0 I' 719.3 756.6 .-

31. --.-. .

_ _ _ _ 1_5 ._~

L6. 748.8 INSTRUCTIONS Ca this fors, list the citrage daily unit p:ver level in f.We-Het for ecch dcy in the reporting eenth.Corpote to the nearest whole negawatt.

. Thest figures will be estd to plat a grooft (gr ecch reparting d far the net elettrical rating t nenth. Nste that when notinnn d

.ok+h i

line lor the restricted oower fevel line),In such [ eses,the overage detly unit power $utput sheet shield te fistnoted te erslalit the apparent anonaly

?

._. ___.c.... . _ . _ _ . - __

, . - - - - - -. ., ~ - . , ..__w

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

,e... . ., or ,.

APPENDIX B

- AVERAGE. DAILY UNIT FOWER LEVEL DOCKET NC. 50-265

" - ~~-~~~ ~

UNIT TWO DATEJone i 1981 - _ _ __ _

COMPLETED BYR ober t C Tubbs TELEPHONE 309-654-2241XT7T ~~~

tiONTH Mriv 1981 _ .

DAY AVERAGE DAILY POWER LEVEL LAY AVERAGE DAILY POWER LEVEL (HWe-Net)

(MWe-Net) 17, 672.6

1. 755.4

~ ~ " ~ ~ ~

~ ~

' 18.

~

~ 727.8 730'.~4

~

2.

19. 710.1 3, 739.0 _

~

20, 698.8

4. 734.4

~ 21. ~ ~ 67 6 . 4 ' ~ --" ~ ~ ~ ~

5. 725.7

22. 685.9

6. 728.5

23. 606.3

7. 720.3 ~

24.

~

' 6 9 0 . 7 ' ' ---~ ~~- ~

3. 720.2

25. 692.3 P. 616.8 _

26, 678.5 10, 721.7

~ -

27. 675.~2

.11.

727,3 28, 655.0

.me12,4e 731.1 -. . ..

=

ep-29, 658.5

13. _

732.9

30. 543 . 8- ~ - -

L4. 721.4

31. 652.4 719.8 _ ..-. ..

__ ._ 15._ .

16. 696.1 INSTRUCTIONS 3 this forn, list the overege daily unit power level in Mile-Net Far ecch day in the reparting nonth.Corpite to th neerest uhele sqcwatt.

Dest figeres Will be used to plat a graph fer tech reporting conth. Note that when tvel ettttds natinen thedependogle re d isr the net tiectrical rating of the gnit there ecy be occasions unen the delly 07er09t DCuer ihl

53 line ter tne restricted power level line),.In such ceses,the average dily onit power output sheet shield be f 2stnoted to explain the ap;crent onantly .

pwa w v- -- u 9- o g- w,- 4 -w-- .- -+g-arr- m yy ,-eg y-- m -- -pw p ee3 - q- es em --- 4 -- - we,,-  :--wv-

3 p . .:

gm.) p .m .3 g- s $ g r .sg  ; .5 p ..c3 y~ ; g-~ . >i t~5 -1 ;o q m c  : o

, .q APPENDIX D QTP 300-013

' UNIT SHUTDOWNS AND l'0WER RECUCTIONS Revisicei 5 March 1978' DOCKET NO. _ 50-254 ,

Qu d-CItles Unit One p UNIT NAt1E June 1, 1981 TELEPHONE 309-654-2241, DATE REPORT HONTH gay 1981 -,

e.x t . 174 m 8 -

5 m = 86 x gg gg E c: 8 $ $ LICS . ,

go DURATION $ g{a d Eh 08 g:8 '

NO. DATE u-(110VRS) a: y5g REPOR1 ,

• g CORRECTIVE ACTIONS / COMMENTS O

n

} _

F 0,. 0 8 5 CD HOTORX Load reduction to add oil to Recirculatio'n 81-9 810510 Pump Hotor Bearings

~

8 2 ZZ ZZZZZZ Maintenance Outage to repair leaking seal on

. 81-10 810522 S 52.2 Feedwater Check. Other items worked included; electromat:c pilot valves, MG set brushes and scram ti .Gng during ascension 1

i  !

', i, .

1 .

i  !

i

, i .

a .

i t

( f in al')

g.. g rq p .) g- .--) e. ; y~ ,) p- ~ 3 pe q m m m ,  ; y - . _

7 3 4. .. g... . z p... S , , .

1

• 'h ,

APPENDIX D QTP 300-S13 UNIT SilVTDOWNS AND POWER REDUCTIONS Revisioi 5 50-265 March 1978 .

DOCKET NO.

Quad-Citles Unit Two COMPLETED BY Robert C Tubbs UNIT N/#.C

""* I' I3 ' REPORT MONTH.

TELEPHONE 309-654-2241, DATE MAY 1981 ext. 174

~

w $ e r = Eb c a .

E ej :8 g-6 LICENSEE g8 5g . ,

po DURATION  ;$ pg* EVENT *8 m.

g:8 g CORRECTIVE ACTIONS / COMMENTS

'-

• y8g REPORT NO.

NO. DATE (HOURS) i R RB CONROD Load reduction to perform Turbine Tests and 81-10 810509 5 0.0 9/H 5 perform special rod maneuvers ZZ ZZZZZI Start of coastdown to End of Cycle Five 81-11 810510 S 0.0 'H 5 Refueling .. t RB CONROD Load reduction to perform Turbine Tests and 81-12 810530 S 0.0 B/H 5 perform special rod maneuvers

' 1 j .

i l

k 4

i i .

1 (finai)

l VI. UNIQUE REPORTING REQUIREMENTS l

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

A. Main Steam Relief Valve Operations .

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

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

The following table is a complete sumary 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.

9 l

t p

i i

i

, y e,y- - *' -*-*v-m--e N*w--+-"N'- - ' " '"* * " " '-** " ' - ' -N" '* 'e'**P ' ~ -

• RESULTS OF SCRAM TIMING ttEASUREMENTS PERFORMED ON UNIT 1 & 2 CONTROL ROD DRIVES, FROM I~I-8I TO 12-31-81 AVERAGE TIME IN SECONDS AT % Max. Time ,

INSERTED FROM FULLY WITHDRAWN For 90%

insertion DESCRIPTION 20 50 90 Technical Specification 3 3.C.I &

NUMBER 5 0.375 0.900 2.00 3.5

~

7 sec. 3.3.C.2 (Average Scram insertion Time 1 DATE OF RODS 88 0.29 0.66 1.42

• 2.49 2.83 (E-8) Unit 1 "A" Sequence Hot 5-25 4

o G

~-.-*.=-w-- .- ,

Vll. REFUELitG INFORMATION .

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

1 0'Brien to c.. Reed, et. al . , titled "Dresden, quad-Cities, and Zion Station--

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

i i

=

l l

i l

• QTP 300-532 R2 Vision 1 <

, Harch 1978 QUAD-C! TIES REFUEllHG I!! FORMATION REQUEST h, ,

s 6 2 Reload: 5 Cycle:

1. Unit:

8-30-81 (Shutdown EOC5)

2. Scheduled date for next refueling shutdown:

12-20-81 (Startup BOC6)

Scheduled date for restart following refueling:

3

4. Will refueling or resumption of operation thereaf ter require a technical specification change or other license amendment: No, Plan The review will be 10CFR50 conducted 59 Reloads by for future cycles of quad Cities Ur.it 2.

early August,1981.

5 Scheduled date(s) for submitting proposed licensing action and supporting Information: Early August, 1981 for 10CFR50.59 related changes a/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, nea operating procedures:

New Fuel Design: 1. Barrier Fuel g

2. Control Cell Core G- ..

D ~_* * * *

I,

• 3 7 The number of fuel a,ssembtles.

724 i a.  !!umier of assemblies in core:

672

b. Number of assemblies in spent fuel pool:

iI

'l 8. The present licensed spent fuel pool storage capacity and l

! In number of fuel assemblies: .

1460 i  :

a. Licensed storage capacity for spent fuel: '

None

b. Planned increase in licensed storage:

9 The projected date of the last refueling that can licensed be discharged capacity: September.to the 1984 spent fuel pool assuming the present j

(End of batch discharge capability) ./? y) P F1 C) \/ EE ,!.

. e.

i 1,

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

i m me _ _ _ _ --+

e *E me ee ==ew

• eooo

. QTP 300-532 Revision 1 flarch 1978

, QUAD-CITIES REFUEll!!G

~

stlFORMATIOff REQUEST 3(* ,

A Reload: 6 Cycle: 7

1. Unit: 1

_9 82 (Shutdown E0C6)

2. Scheduled date for next refueling shutdown:

12-5-82 (Startup BOC7) 3 Scheduled date for restart following refueling:

4. tilli refueling or resumption of operation thereaf ter require a technical tio, Plan 10CFR50 59 reloads specification change or other license amendment:The review will be conducted in for future cycles of Quad Cities Unit 1.

August, 1982.

5 Schedule 3 date(s) for submitting proposed licensing action and supporting

. Information: August,1982 for 10CFR50.59 related changes ~ 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 fuei design, new operating procedures:

tiew fuel d.esigns:

7 The number of fuel assemblies.

724

a. ilumber of assemblies in core:

820

b. Plumber of assemblies in spent fuel pool:

8. The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requasted or is planned

• in number of fuel assemblies:

140

a. Licensed storage capacity for spent fuel:

ficne

b. Planned increase in licensed storage:

f 9

The projected date of the last refueling that con licensed capacity: be discharged Sep temberto, 1985 the spent ft 2 pool assuming the present (end of batch discharge capability) XPPROVEC APR 2 01973 Q. C. c. S. R.

.-.._....__...._._._7...___.

\

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

ACAD/ CAM - Atmospheric Containnent Atmospheric Dilution / Containment Atmospheric Monitoring ANSI -

American National Standards Institute APRM - Average Power Range Monitor ATWS - Anticipated Transient Without Scram BWR -

Boiling Water Reactor CRD -

Control Rod Drive EHC - Electro-Hydraulic Control System EOF -

Emergency Operations Facility

- GSEP - Generating Stations Emergency Plan HEPA -

High-Efficiency Particulate Filter HPCI -

High Pressure Coolant injection System HRSS -

High Radiation Sampling System IPCLRT - Integrated Primary Containment Leak Rate Test IRM - Intermediate Range Moni tor 151 - In-Service inspection LER - Licensee Event Report LLRT -

Local Leak Rate Test LPCI - Low Pressure Coolant injection Mode of RHRS

LPRM - 8.ocal Power Range Monitor

. MAPLHGR* - Maximum Average Planar Linear Heat Generation Rate MCPR -

Minimum Critical Power' Ratio MPC -

Maximum Permissible Concentration MSIV -

Main Steam Isolation Valve NIOSH - Nation.1 Institute for Occupational Safety and Health PCI - Primary Containment isolation PCIOMR - Preconditioning interim Operating Management Recommendations RBCCW - Reactor Scilding 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 SBGTS

- Standby Gas Treatment System SBLC - Standby Liquid Control SDV

- Shutdown Cooling Mode of RHRS SDV

- Scram Discharge Volume SRM - Source Range Monitor TBCCW - Turbine Building Closed Cooling Water System TIP - Traveling incore Probe TSC - Technical Support Center

-*=ameng e -* w- me *e _ _ - -M y  %.., , . ==

+m*y-m-s---+- w wm s 3, e- 1r y mer w w v vw- --- iwv --h-