ML20004F137
| ML20004F137 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/01/1981 |
| From: | Tubbs R COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20004F134 | List: |
| References | |
| NUDOCS 8106160543 | |
| Download: ML20004F137 (27) | |
Text
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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
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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 --
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l.
INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Boiling Water located Reactors, each with a Maximum Dependable Capacity of 769 MWe net, in Cordova, Illinois. The Station is jointly owned by Commonwealth Edison The Nucleat Steam Company and lowa-Illinois Gas & Electric Company.
The Supply Systems are General Electric Company Boiling Water Reactors.
Architect / Engineer was Sargent & Lundy, incorporated and the primary The con-construction contractor was United Engineers & Constructors.
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, 1972, respectively, pursuant to Docket Numbers 50-254 and 50-265 The date of initial reactor criticalities for Units 1 and 2 respectively were October 18, 1971, and April 26, 1972. Commercial generation of power began on February 18, 1973 for Unit 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.
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II.
SUMMARY
OF OPERATING EXPERIENCE A.
UNIT ONE Unit One began the reporting period holding a load of 815 May 1-9:Over these initial nine days, an average load of 814 MWe was MWe.
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 completed at 0145 Load was held at 700 MWe until 0515 At that 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 IA and IB Recirc Other work items performed during this outage were:
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.
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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, starting at 0600, at various rates until it was held at 1445 The average load for the day was 666 MWe, Load was held on May 29 However, due
~
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 average load for the day was 712.1We.
the cooler river water.
On May 30 at 0030, power was dropped to 500 MWe for the Nuclear At this time the turbine weekly tests were performed and Engineer.
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.
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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
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enable the completion of a high burnup fuel experiment in the present core.
B.
Facility or Procedure Changes Requiring NRC Approval l
There were no Facility or Procedure Changes Requiring NRC l
f approval for the reporting period.
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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.
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UNIT ONE MAINTENANCE
SUMMARY
J CAUSE RESULTS & EFFECTS U.R.
IER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q12240-1/2 Diesel Gener-The field flash The field ground alarm Lubricated the field ator relay was sticking came up.
Diesel Gener-flash contactor;
- closed, ator operability _was tested operable, not affected.
Q12159 l-1001-65C The pump internals The pump had high Replaced shaft IC RHR Service were worn.
bearing vibration, sleeves, bearings, Water Pump The other three RHR seal and balanced Seivice Water Pumps impeller.
4 were operable.
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j Q12192 IB RHR Pump Me,ch5nical seal on Water was leaking Char.3ed cartridge--
pump shaft was from the seal. The type mechanical seal.
worn.
pump was operable.
4 Q12191 lA-1002 RHR Seal was worn.
Water was leaking Replaced mechanical ~
Pump from the seal. The seal and broken stud.
pump was operable, Q12193 10-1002 RHR Seal was worn.
Water was leaking Replaced mechanical from the seal. The seal.
Pump pump was operable.
f Ql264b l-302-19A Wire had pulled The solenoid did not Found wire on coil Back-up Scram loose from the energize when voltage pulled out of wire nut.
Solenold Valve coll leads.
was applied to it.
Reconnected lead with during initial testing.
wire nut.
The RPS system was fully operable.
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UNii ONE MAINTENANCE
SUMMARY
4 CAUSE RESULTS 6 EFFECTS-W.R.
LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFIINCTION SAFE OPERATION PREVENT REPETITION Ql2518 1-220-62A Feed-Steam leak in Steam was leaking in Replaced pressure seal water Check Valve pressure seal ring.
the MSlV room.
No ring and installed new abnormal releasa. of
" tie" wire on hold down radioactive material to bolts for trim.
the environs occurred.
Q11606 M0-1-202-5A Control circuit Old cable had been Replaced cable from cable has shorted spliced as a temporary operator to 3rywell wires--5 conduc-fix. The valve was penetration--verified i
tor cable #12507 fully operable, pump trips on valve closure.
Q11421 1-1001-19A RHR The auxiliary The valve would not Replaced auxiliary 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.
l Q12651 1-2001-16 The air cylinder The valve would not Replaced air cylinder was worn causing open completely. The on valve operator.
air to leak by the isolation capability piston.
was not affected.
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b UNIT TWO MAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS OF ON ACTION TAKEN TO W.R.
LER NUl'.BER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION l
Qll701 81-07/03L A0-2-203-2A--2A There was a steam Failed to get 1/2 scram initially the 10 percent Outboard MSIV leak in the vicin-during testing. The closed switch was ity of the limit other HSIVs functioned exercised and it worked switch from the as designed, satisfactorily. Then RCIC testable during the short outage, the limit check valve.
switch assembly was changed.
Q10879 IRM Number 17 The detector was IRM was failed down-Replaced detector and failed.
scale. The other checked connectors.
IRMs in that channel were operable.
Q12314 2-1001-18A The pressure The valve did not Calibrated pressure RHR Hinimum switch was out of automatically open switch PS-I-1001-81A.
Flow Valve calibration.
during the flow test.
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2-1001-18B The pressure The valve did not Calibrated pressure Q12313 swtich was out of automatically open
.ch PS-1-1001-818.
RHR Minimum
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Flow Valve calibration.
during the flow test.
1
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Q12209 2A RHR Pump The mechanical Water was leaking from Changed mechanical seal.
seal was worn.
the seal.
The pump j
was operable.
i Q12210 28 RHR Pump The mechanical Water was leaking f rom Changed mechanical seal, seal was worn.
the seal. The pump was operable.
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UNIT WO HAINTENANCE
SUMMARY
.i 2 CAUSE RESULTS & EFFECTS W.R.
LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONO!T MALFUNCTION SAFE OPERATION PREVENT REPETITION l
Q12211 2C RHR Pump The mechanical seal Water was leaking from -
Changed mechanical seal, the seal. The pump was was worn.
Q12546 Unit 2 Diesel A voltage control Lost voltage control to
_ Replaced rectifier.
Generator rectifier burnt the generator. The 1/2 l
Diesel Generator and-Voltage out.
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.
Q12549 81-12/03L MO-2-1001-78
~The circuit The valve would not open Replaced auxiliary l
RHR Pump Suction breaker auxiliary
, from the Control Room, contacts on circuit Valve contacts were LPCI and containment breaker.
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sticking.
cooling modes of RHRS were still operable.
l Q12519 81-10/03L Unit 2 Diesel Spurious breaker The 1/2 Diesel Generator Replaced 4 wires in I
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Generator closure ccused was operable. Two off-cubicle 1, Bus 24-1.
3 the generator to site _ lines supplying Replaced wiring and motorize, burning 345 KV power were potential transformers i
some wiring and available. The low in generator.
i potential trans-pressure ECCS and
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formert.
containment cooling
,i mode =, of RHR associated with the operable Diesel Generator were proven operable.
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UNIT _TWO HAINTENANCE
SUMMARY
4 I
CAUSE RESULTS & EFFECTS I,
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W.R.
LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Ql2635 2-2001-16 Drywell Air cylin, der was Valve would not open Replaced e'r cylinder 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.
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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 Date Nurober There were no Licensee Event Reports for the reporting period for Unit One UNIT TWO 81-9/03L 5-4-81 MO-2-1001-378 Breaker tripped.
SI-10/03L 5-15-81 Unit Two olesel Generator inoperable f
81-12/03L.
5-18-81 M0-2-1001-7B RHRS Valve failed to open l
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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
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OPERATING DATA REPORT
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DOCKET NO.
50-254 UNIT ONE
- - ~ - ~ -
DATEJune t 1981
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COMPLETED BYRobert C'Tobb s TELEPHONE 309-654-2241Xi74_
OPERATING STATUS 0000 050181
- 1. ' Rep oFtiiig ~ period : 2400 053181~~GFoss hours in reporting period:
744"-
- 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
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- 3. Power level to which restricted (if any)(MWe-Net):
4.
Reasons for restriction (if any):
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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
- 7. Hours generator on line
~
691.8 3T47.8 "61231 5 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)
CO.~ Gross electrical energy generated (MWH) 530156 2554270~
39933177
~
492672.
2379914 37236193 ii. Het electrical energy generated (MWH) 94.2_
94.1 20.8 12.' Reactor service factor
~ ~ ~' t.3. Recctor avriilobility fac t or~
~
94.2 94 i_ ~ - ~~ 8 5'~ t
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93.0_
92.4 77,1
- 14. Unit service factor 93.0 92.4 78.3 15.-Unit availability factor
~
16.' Unit' capricit y 'f ac t or -(Using MDC) 86,i
~ 85.4
~ ~~~ 61~0~
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- 17. Unit capacity Factor (Using Des.MWe) 83.9_
83.2_
59.4 0. 0_
i.8 7,5
- 10. Unit forced outage rate
- 19. Shu'tdowns scheduled over next 6 months (Typ.e,Date,and Duration of e~ach ) F
- 20. If shutdown at end of report period,estinated date o f s t ar t u p
__,___NA sihe !CC M7 he lower than 769 MWe during perids of high ebicat tegarature due is tne therol perfernnte er the sorcy ccul.
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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
~ 5318i.~G'r o sis h o u r #~'in r e p o r t in g p e r i o d :-
- 1. Reportlig'pdFiod:'2400
~
0
- 2. Currently authorized power level /.MWt): 2511 Max. Depend capacity (hGe~he~t):~ ' ~769* Design" elaEtrical rc't ing '(MWe-Ne t ) : 789
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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
~ '7. Hours generator on line-~~
~
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744.0.
'3501.8 "59783TO~
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0. 0_
0.0 702.9 S. Unit reserve shutdown hours.,
'9. G[oss thermal energy generated (MWH) 1701924 8138137_
12333854_5_,
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~f0. GFos~s diectiical energy generatdd(MWH)
- 543830 2591717_
39313268-
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- 11. Het electrical energy generated (MWH) 519000 2459614 36816566 100.0 97.4 79.5
&2. Reactor service factor
+
~ ~i3'. Re5~ctor cvsil'ab11'ity' factor ~ ~~
~"~
100.O
~ ' '97 i4
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100.0_
96.7_
76.2 14. Unit _ service factor.
100.0 96.7.
77.1
- 5. Unit availability factor
- 16. Unit capacity factor (Using MDC) 90.7 88.3
~~ ~ 6170'
- 17. Unit capacity factor (Using Des.Mue) 88.4 86.0.
59.5 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) i.
762.0 17.
659.0
. 757.'8
' ' 18.
~~ 754.6 717.5 17.
7U6.7 3.,
753.6 20.
756.0 4.
'5.
756.0 21.
764,4 6.
765.0 22.
666.8 7.
760.2 23.
-31.6 24.
-26.6 9.
765.0 25.
269.0 9.
762.1 to.
567.4 26.
497.4
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602.4 ~
~~
27.
667.8 754.5 28, 772.9
. _12,
13, 760.7 29.
732.4 30.
747.3
- 14.
754.0 I'
719.3 31.
756.6
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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
. Thest figures will be estd to plat a grooft (gr ecch reparting nenth. Nste that when notinnn d nearest whole negawatt.
t
.ok+h line lor the restricted oower fevel line),In such [ eses,the overage detly unit power $utput sheet shield te d far the net elettrical rating ifistnoted te erslalit the apparent anonaly
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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) 755.4 17, 672.6 1.
2.
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730'.~4
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~ 727.8 3,
739.0 19.
710.1
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698.8 4.
734.4 20,
~ ~ 67 6. 4 ' ~ --" ~ ~ ~ ~
5.
~
725.7 21.
6.
728.5 22.
685.9 7.
720.3 23.
606.3
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3.
720.2 24.
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692.3 P.
616.8 25.
678.5 10, 721.7 26,
.11.
727,3 27.
675.~2
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28, 655.0 731.1 ep-
.me12,4e 658.5
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29, 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 Dest figeres Will be used to plat a graph fer tech reporting conth. Note that when natinen dependogle re neerest uhele sqcwatt.
tvel ettttds the d isr the net tiectrical rating of the gnit there ecy be occasions unen the delly 07er09t DCuer ihl line ter tne restricted power level line),.In such ceses,the average dily onit power output sheet shield be
- 53 f 2stnoted to explain the ap;crent onantly pwa w
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,.q APPENDIX D QTP 300-013 UNIT SHUTDOWNS AND l'0WER RECUCTIONS Revisicei 5 50-254 March 1978' DOCKET NO. _
Qu d-CItles Unit One p
UNIT NAt1E June 1, 1981 TELEPHONE 309-654-2241, DATE REPORT HONTH gay 1981 e.x t. 174 8
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CORRECTIVE ACTIONS / COMMENTS y5g REPOR1 (110VRS) a:
NO.
DATE u-On
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81-9 810510 F
0,. 0 8
5 CD HOTORX Load reduction to add oil to Recirculatio'n Pump Hotor Bearings 81-10 810522 S
52.2 8
2 ZZ ZZZZZZ Maintenance Outage to repair leaking seal on
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Feedwater Check. Other items worked included; electromat:c pilot valves, MG set brushes and scram ti.Gng during ascension 1
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- 'h APPENDIX D QTP 300-S13 Revisioi 5 UNIT SilVTDOWNS AND POWER REDUCTIONS March 1978 50-265 DOCKET NO.
Quad-Citles Unit Two COMPLETED BY Robert C Tubbs UNIT N/#.C TELEPHONE 309-654-2241,
""* I' I3 '
REPORT MONTH.
MAY 1981 DATE ext. 174
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EVENT
- 8 g:8 CORRECTIVE ACTIONS / COMMENTS (HOURS) y8g REPORT NO.
m.
g NO.
DATE R
i 81-10 810509 5
0.0 9/H 5
RB CONROD Load reduction to perform Turbine Tests and perform special rod maneuvers 81-11 810510 S
0.0
'H 5
ZZ ZZZZZI Start of coastdown to End of Cycle Five Refueling t
81-12 810530 S
0.0 B/H 5
RB CONROD Load reduction to perform Turbine Tests and perform special rod maneuvers 1
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l VI. UNIQUE REPORTING REQUIREMENTS The following items are included in this report based or. prior commitments l
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.
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- 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 NUMBER 5
20 50 90 Technical Specification 3 3.C.I &
DATE OF RODS 0.375 0.900 2.00 3.5 7 sec.
3.3.C.2 (Average Scram insertion Time
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1 5-25 88 0.29 0.66 1.42
- 2.49 2.83 (E-8)
Unit 1 "A" Sequence Hot 4
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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.
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QTP 300-532 R2 Vision 1 <
Harch 1978 QUAD-C! TIES REFUEllHG I!! FORMATION REQUEST h,
s 6
1.
Unit:
2 Reload:
5 Cycle:
8-30-81 (Shutdown EOC5) 2.
Scheduled date for next refueling shutdown:
12-20-81 (Startup BOC6)
Scheduled date for restart following refueling:
3 Will refueling or resumption of operation thereaf ter require a technical 4.
specification change or other license amendment: No, Plan 10CFR50 59 Reloads for future cycles of quad Cities Ur.it 2.
The review will be conducted by early August,1981.
Scheduled date(s) for submitting proposed licensing action and supporting 5
Information: Early August, 1981 for 10CFR50.59 related changes a/90 days prior to shutdown.
Important licensing considerations associated with refueling, e.g., new or different fuel design or supplier, unreviewed design or performance analysis 6.
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
!!umier of assemblies in core:
i a.
672 Number of assemblies in spent fuel pool:
b.
iI The present licensed spent fuel pool storage capacity and
'l 8.
In number of fuel assemblies:
l 1460 Licensed storage capacity for spent fuel:
i a.
None b.
Planned increase in licensed storage:
The projected date of the last refueling that can be discharged to the licensed capacity:
September. 1984 9
spent fuel pool assuming the present j
(End of batch discharge capability)
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1, APR 2 01978 i !
Q.C.O.S.R.
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QTP 300-532 Revision 1 flarch 1978 QUAD-CITIES REFUEll!!G 3(*
stlFORMATIOff REQUEST
~
A 1.
Unit:
1 Reload:
6 Cycle:
7
_ 82 (Shutdown E0C6) 9 2.
Scheduled date for next refueling shutdown:
12-5-82 (Startup BOC7) 3 Scheduled date for restart following refueling:
tilli refueling or resumption of operation thereaf ter require a technical 4.
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.
Schedule 3 date(s) for submitting proposed licensing action and supporting 5
August,1982 for 10CFR50.59 related changes ~ 90 days prior to Information:
shutdown.
Important licensing considerations associated with refueling, e.g., new or
. 6.
- 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 ilumber of assemblies in core:
a.
820 b.
Plumber of assemblies in spent fuel pool:
The present licensed spent fuel pool storage capacity and the size of any 8.
increase in licensed storage capacity that has been requasted or is planned in number of fuel assemblies:
140 Licensed storage capacity for spent fuel:
a.
ficne f
b.
Planned increase in licensed storage:
9 The projected date of the last refueling that con be discharged to the licensed capacity: Sep tember, 1985 spent ft 2 pool assuming the present (end of batch discharge capability)
XPPROVEC APR 2 01973
_. Q. C. c. S. R.
.-.._....__...._._._7...___.
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Vilt. GLOSSARY The following abbreviations which may have been used in the Monthly Report, are defined below:
Atmospheric Containnent Atmospheric Dilution / Containment ACAD/ CAM Atmospheric Monitoring American National Standards Institute ANSI Average Power Range Monitor APRM Anticipated Transient Without Scram ATWS Boiling Water Reactor BWR Control Rod Drive CRD Electro-Hydraulic Control System EHC Emergency Operations Facility EOF Generating Stations Emergency Plan
- GSEP High-Efficiency Particulate Filter HEPA High Pressure Coolant injection System HPCI High Radiation Sampling System HRSS Integrated Primary Containment Leak Rate Test IPCLRT Intermediate Range Moni tor IRM In-Service inspection 151 Licensee Event Report LER Local Leak Rate Test LLRT Low Pressure Coolant injection Mode of RHRS LPCI 8.ocal Power Range Monitor LPRM Maximum Average Planar Linear Heat Generation Rate MAPLHGR*
Minimum Critical Power' Ratio MCPR Maximum Permissible Concentration MPC Main Steam Isolation Valve MSIV Nation.1 Institute for Occupational Safety and Health NIOSH Primary Containment isolation PCI Preconditioning interim Operating Management Recommendations PCIOMR Reactor Scilding Closed Cooling Water System RBCCW Rod Block Monitor RBM Reactor Core isolation Cooling System RCIC Residual Heat Removal System RHRS Reactor Protection System RPS RWM Rod Worth Minimizer Standby Gas Treatment System SBGTS Standby Liquid Control SBLC SDV Shutdown Cooling Mode of RHRS SDV Scram Discharge Volume Source Range Monitor SRM Turbine Building Closed Cooling Water System TBCCW Traveling incore Probe TIP TSC Technical Support Center y
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