ML20028C987
| ML20028C987 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 01/03/1983 |
| From: | Buss R COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20028C975 | List: |
| References | |
| NUDOCS 8301140377 | |
| Download: ML20028C987 (26) | |
Text
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QUAD-CITIES NUCLEAR POWER STATION UNITS.1 AND 2.-
MONTHLY PERFORMANCE REPORT DECEMBER 1982
- COMMONWEALTH EDISON COMPANY 1
AND IOWA-ILLINOIS CAS & ELECTRIC COMPANY j NRC DOCKET NOS. 50-254 AND 50-265 LICENSE NOS. DPR-29 AND DPR-30 4
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8301140377 830103 PDR ADOCK O W fj R
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-TABLE OF. CONTENTS
'I. Introduction.
I1. Summary -of Operating Experience A. Unit One B. Unit Two III. '_ 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
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A. Operating Data Report-B. Average Daily Unit Power Level .
C. Unit Shutdowns and Power Reductions VI. Unique Reporting Requirements A. Main Stean - Relief Valve' Operations B. Control Rod Drive Scram Timing Data VII. Refueling Information VIII. Glossary
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I. INTRODUCTION Quad-Cities. Nuclear Power Station is camposed of two Boiling -
Water Reactors, each.'with a Maxistan 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 Systens are General Electric Company' Boiling Water Reactors. The Architect / Engineer was Sargent & Lundy, Incorporated, and the primary construction contractor was United Engineers & Constructors.' The condenser cooling method is a closed cycle spray canal, and the Mississippi River is the condenser cooling water source. The plant is subject to license 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 reac*or 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 canpiled by Becky Brown and Randall. Buss, telephorie ntunber 309-654-2241.,- extensions 127 and 181.
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II.
SUMMARY
'0F 'OPERATIE EXPERIEEE A. Unit One December 1-23: The unit began .the month shutdown for refueling at the end of Operating Cycle Six. On December' 21, at 0610 hours0.00706 days <br />0.169 hours <br />0.00101 weeks <br />2.32105e-4 months <br />, the unit was made critical. On December 22 the unit was placed on line at 40
.MWe. The Turbine-Generator was tripped 24 minutes later, and Reactor thermal power was maintained to cause the deterioration of a rubber shoe cover which had been lost .in the Reactor vessel. On December 23, at 1147 hours0.0133 days <br />0.319 hours <br />0.0019 weeks <br />4.364335e-4 months <br />, the unit was placed on line and began' increasing load at 50 MWe/ hour. Load was held at 160 MWe due to an inboard seal leak on the IB Recirculation pump. Control rod. drive hot scram timing was performed at this time.
December 24-31 : ' On December 24, load was increased to 400 MWe. At 1245 hours0.0144 days <br />0.346 hours <br />0.00206 weeks <br />4.737225e-4 months <br /> load was decreased to 200 MWe due to Traversing in-Core Probe trouble. On December 26, at'2311 hours0.0267 days <br />0.642 hours <br />0.00382 weeks <br />8.793355e-4 months <br />, the unit was shutdown to perform maintenance on the probes. On December 27, at 1748 hours0.0202 days <br />0.486 hours <br />0.00289 weeks <br />6.65114e-4 months <br />, following maintenance to the Traversing in-Core Probes, the Reactor was made critical and the unit was on line at 0120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> on December
- 28. On December 29 load was decreased.from 450 MWe to 250 MWe' to correct the wiring of the Local Power Range Monitors. Load began increasing to maximum load on December 30.
B. Unit Two December 1-31: The unit began'the month' maintaining'a load of approximately 805 MWe. On December 5, load was decreased to 700 MWe for weekly Turbine tests, then increased to 812 MWe. On December 11, load was decreased to 550 MWe to perform control' rod pattern adjustments.
Load then increased to 823 MWe by December 13 Load was decreased to 700 MWe for weekly Turbine tests on December 19, then increased to 820 MV 3. Load was decreased four times at the request of the Load Dispatcher due to low system demand. On December 30 load was dropped to 650 MWe due to high Condenser backpressure. On December 31, load was dropped to 600 MWe at the request of the Load Dispatcher, then increased at 50 MWe/ hour to 700 MWe and 5 MWe/ hour thereafter.
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t III, PLANT OR ' PROCEDURE CHANGES TESTS, EXPERIMENTS, ' AND ' SAFETY REIATEU MAln tz, MANGE A. Amendments to Fac!11ty License or Technical Specifications On November 2, 1982, Amendments 82 and 76 were issued to licenses DPR-29 and DPR-30, respectively. These amendments revised section 3 7.2./.4.7.2, requi rements for Primary '
Containment Integrity and Containment Leak Rate Testing.
The new requirements specify a combined leakage rate of less than or equal to.60 La for all penetrations and valves except for Main Steam Isolation Valves subject to type B and C tests, and a leakage rate of 11.5 scfh for any one Main Steam Isolation Valve. The changes also permit type A tests of less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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. This summary includes the following headings: 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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UNIY O!!E MAINTENANCE SUMMARV RESOLTS & EFFECTS ACTION TAKEN TO CAUSE ON OF PREVENT REPETITION SAFE OPERATION W.R. LER COMPONENT MALFUNCTION The penetration bolts NUMBER The penetratior; seal were tightened.
NUMBER The penetration was leaking during 82-26/03L RHR Service Water packing tightening Q21942 "A" Vault Sunp leak testing.
bolts were loose, Penetration Replaced the studs and The valve was still nuts that were missing.
Several studs and operable, 1-202-30 Recirc nuts were missing Q23005 Discharge Valve from the yoke to bonnet connection. Rebuilt operator with The operator piston Theclose, valve would not new seals and stroked 1-54013 SJAE seals were worn. the valve.
Q21027 Suction Valve The cable was replaced The valve was still with one from the 1-202-The valve control operable. 7A valve, which has been l-202-5A Recire cable was damaged.
Q20919 Discharge Valve removed.
A new cell was installed.
The cell failed 250V Battery The cell was at the during the battery end of its life. discharge test.
Q22180 Cell 88 A nov cell was installed.
The cell failed 250V Battery The cell was at the during the battery end of ,i ts li fe. discharge test.
Q22181 Cell 14 Re-seated the valve plug.
The valve was leaking The valve plug was .through.
A0-1-302-210 not seating Q23136 SDV Vent Valve properly. Re-seated the valve plug.
The valve was leaking The valve plug was through.
A0-1-302-21A not seating Q23137 SDV Vent Valve properly.
[l UNIT ONE MAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS W.R. LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q23138 A0-1-302-22 The valve plug The valve was leaking Re-seated the valve' plug.
SDV Drain Valve was not seating through.
p roperly.
Q22121 1A Recirc Pump The mechanical seal The seal cavity A new mechanical seal Seal (IA-202) was worn. pressure was spiking was installed.
up to 800 psig.
Q23055 82-26/03L A0-1-1601-55 The valve seating ~ The valve failed during The valve was cleaned; U2 Purge Valve surfaces were Local Leak Rate Testing. the LLRT was satisfactory.
dirty.
Q21512 82-26/03L I-1301-41 RCIC The check valve The valve failed during -The check valve disc &
Exhaust Check disc was bad. Local Leak Rate Testing. associated parts were Valve rep? aced.
Q21539 82-26/03L A0-1-203-2C The valve disc was The valve failed during The valve seat & disc-MSIV not seating Local Leak Rate Testing. were- lapped. The valve p rope rly, passed the LLRT.
Q21540 82-26/03L A0-1-203-IB The valve disc was The valve failed during The valve seat & disc' MSIV not seating . Local Leak Rate Testing. were lapped. The valve
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properly. passed the LLRT.
Q21541 82-26/03L A0-1-203-2B The val've disc was The valve failed during The valve seat & disc MSIV not seating Local Leak Rate Testing. were lapped. The valve-properly. ' passed the LLRT.
Q23329 1RM 16 The rod block limit The withdraw rod block The limit switch was switch was out of occurred sooner than adjusted and tested adj us tment. necessary. satisfactorily.
a UNIT OfiE MAINTENANCE
SUMMARY
CAUSE RESULTS s EFFECTS W.R. LER OF ON ACTION TAKEN TO NUMBER NUM8ER COMPONENT MALFUNCTION SAFE OPERATION -PREVENT REPETITION Q21437 82-26/03L Drywell Head The flange 0-rings The flange failed the Cleaned flange and Flange were bad. . Local Leak Rate Test. Installed new 0-rings.
Q21513 82-26/03L I-2301-45 The valve disc was !!PCI steam exhaust The check valve was HPCI Exhaust damaged. valve was found to leak replaced.,
Check Valve excessively during Local Leak Rate Testing.
Q21534 82-26/03L CV-1-220-583 The valve seating The valve failed The seating surfaces-Feedwater Check surfaces were during Local Leak Rate were cleaned and~the Valve dirty. Testing. seat 0-ring was replaced.
Q21633 1-1402-93 The valve stem was The valve will not The operator was binding. operate from the removed and the stem was Control Room. freed.
Q22442 Bus 19 The control switch Bus 19-18 tie breaker. The contacts were cleaned.
19-18 Tie contacts were would not close-in and the switch was Breaker dirty. from Control Room. tested satisfactorily.
Q22933 1-1102B SBLC The pump motor The pump was not The power cable'was Pump power cable was required during replaced.
. damaged. . refueling.
Q23189 82-26/03L I-220-62B The valve seating. -The valve failed the The valve-was cleaned Feedwater Check surfaces were Local Leak Rate Test.. .and inspected.
Valve dirty.
Q23242 1-1001-23A RHR The valve breaker The breaker tripped A new breaker was Containment Spray was worn. out while trying to installed.
Valve stroke valve.
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UNIT OilE MAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS W.R. LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q23338 1-203-1B MSIV The pilot valve Air leaks out of the The pilot valve was 0-rings were exhaust port, replaced.
worn.
Q20421 1-201 Reactor Replaced Feedwater Replace Feedwater Removed clad and Vessel Spargers. Spargers and perform installed Spargers.
clad removal.
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IV . LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for Quad-Cities Units One and Two occurring during the reporting period, pursuant to the reportable occurrence reporting requirements as set forth in sections 6.6.B.1. and 6.6.B.2. of the Technical.
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_ Specifications.
UNIT ONE Licensee Event Report Number Date Title of Occurrence 82-37/03L 13-82 Automatic Blowdown Timers out of limits 82-39/03L 12-24-82 Unit 1/2 Diesel Generator failed to start due to broken air line UNIT TWO There were no Licensee Event Reports for Unit- Two for the reporting period.
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.V. DATA TABUIATIONS The follo' wing data tabulations are presented 'in this report:
A, Operating Data Report B, Average Daily Unit Power Level C, Unit Shutdowns and Power Reductions d
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OPERATINGLDATA REPORT
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.._ ' DOCKET-NO. '50-254 s
ee ~ _. ~. , . . - . ...m_.-.._.a _ _ - m u UNIT EONE' - _ _ . . _
DATEJo'nuarv'03'i983 r_e . , ,
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' COMPLETED:BYRondoll'D Buss-a, . . _ . _ , . _ __ _ _. ._... , _.m._ _ _ m ._ ._ IELEPHONE309-654-2241xiBi- i OPERATING STATUS ' .
_.x _ _ . _a .. . _ _ _ _ _0000.120182 _ _ __. _ __ '...s _. _c._.,._ . . . _ . . _ .i .. ____ _-- 'i 1.. Reporting period 2400~123182 Gross', hour.s"in reporting period 744: -
a 2, .._Curren t lyf ou th ori zed .p..owetale velm(MWt ) : 2511_Hox. Depend copacity_ 4 - ,_ .
~(MWe-Net): 769* Design electrical rating _ (MWe-Net) 789 '
,._ . 3._ P owe r le v e 1_ t o wh ich _ test tic.t ed ( lt onyf) ( MWe--Ne t ) L N A. ._ ___._ _. _ , _ _ . . .
- 4. Reasons.for restriction (if o'ny)'s
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This Month Yr.to Date Cumulative ~
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- 5. Number of_ hours..r.eoc. tor _was_cr,itico C 239.0 _ 6072.1 _- 75171.2 _._
- 6. Reactor reserve shutdown hours O.0 '0.0 34215
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- 7. Hours generator on line $ 782 9, 5955.1 72006.6 D. Unit _ reserve._ shutdown _ hours..__ _ . _ _ ~0.0_ 0.0,_ 909.2._..
'/ . Gross thermal energy generated (MWH)' ' i54149
' i1154632~ - i46212991
- 10. Gross electrical energy generated (MWH)_- . . _ _ . 59734__ 3592948 __47121881-__..
_. 11. Het electrical ener.gy._ generated.(.HWH) _ 50185_. 3244824 _ 43828908 _,.._
- 12. Reactor service factor 32.1 69.3 80.6
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1.3 . Reoctor av011ob111ty foctor ~ ^ '32Ti ~ 69'.3 ~ BA3
. .L 4. . Unit. service foctor. _ . _
23.9._ 6 0 . 0 _, 77.3 _._,
1.5.. Un it avullability factor 23.9- 68.0 78.2
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~ 16. Unit capacity factor (Using'MDC) ~
O . 8' 48.2 61.1
- 1. 7 . Unit capacity.foctor (Usin.g Des.MWe)_ _ 8.5 .
46.9 _ 59. 5, _
-18. Unit forced outage rate 13.0 1.9 6.7
- 19. Shutdowns scheduled over next 6 nonths (Type,Date,and Duration of each):
i..ja0 . If shutdown at end. of. report perl.od estf inated dote..o.f _ star t up.
. NA 8The MC not be lower then 7M leie dering perleds of high emblent temperatore 'det f to the _thernel perfernence.ef the spref.cenel._ _. . _ _ _ __ . _ _ . _
1 SWOFFICIAL COlFAllt HUMIERS ARE USD IN THl$ REPORI L _
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-- OPERATING DATA. REPORT- -
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. DOCKETJNO. =50-265'
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_ _ _.___q.a _ u 1 -.i. UNIT ~ TWO- 1 u . .#
DATEJonvorv 03'1983 -
. . COMPLETED'BYRondall"D Buss-4 .
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5 .0PERATING STATUS: .. .
_ , . - . _ ... m.. _ _ ._ .,.,.._0.0.00.i201821 _ . _ _ _ _ , _ _ _ _ _ _ _ . _ . . . . . _ _ _ . _ . -
t.: Reporting period:2400 123182" Gross hours in repor_ ting period: '744-
, n
._ ._ .12 ', s Cu r r e n t l y. .ou t h o c i z e d ..p owe r_le v e l ;.( MW t Lin 25ii_ Max . De pe nd; c ap aci t yi - -
(MWe-Net): 769* Design electrical.roting-(MWe-Net) :789 ~.
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- 3. . P o we r le v e 1_ t oc wh ich1 e a s t r.ic_t ed ( 1 f._on y.) ( HWe rNe.t ),siH A._._.
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- 4. Reasons for~ restriction.~.(if ony):. .
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This Month Yr.to Date Cumulative
, _ - .5 .. . N u mb e r. o.f h.o u r. .s r.e ac..t. o r_ was_c r. .i .t. i c al__
744,0_ 7411.6 72263.4 .._ ,
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'6. Reactor reserve shutdown hours 0 'O
- 0 '0 2985.8'
- 7. Hours generator on line 744,0 7346.9 69588.1
,0., Unit reserve shutdown; hours.. _ ._. _ ._
0.0 - 0 . 0 ,. 702 1._...
- 9. Gross thernal energy generated (MWH) 1755870 16704411 i'44591494-
-. ~. .. . . . . .- . .. ~ .. . . - - . . ~ ... .. - . . = .. ~. - .. - . - .. _ -. -
- 10. Gross electrical energy generated (MWH) 569540 5331295 46037535
_ 11. . Net electrical _ energy. generated (MWH)._ 532969 , 5058983_ 43183567 _
- 12. Reactor service foctor 100.0 84.6' 78.2
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L3. React or avo11ob11'It y f act or 100.0 84.6 81.5-
- 14. . Unit service foctor. _ _ . _ _ , , ._
100.0 83.9 _ 75.3 _. .,
L5. Unit avullability_ factor. 100.0 83.9 76.1
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L '6 . Unit capacity factor (using'MDC) ~' 93.2 75,i 60 .U~
- 17. Unit capacity. factor (Using. Des.M.We)_. 90.8 73.2 59.3 __
L8. Unit forced-outoge rote ,
0.0 14.6 9.1
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_ 19. Shutdowns scheduled over next. . . . ~ . . .6 months. . . _ . -(Type,Date ,and Duration of each ):
, ;20. ;If shutdown,at_end.of. report _perlod, estimated dote.ofistortup NA
- 8The IWC not be lever then 7H leie dering perleds of high edient temperatore doe to the _thernel.perfernence of the spret cenel. _
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= IWWFFICIAL COlFAllt 14500$ Allt USD Ill THIS RDORT ,
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~ APPENDIX B. . .
AVERAGE- DAILY = UNIT POWER LEVEL '
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.._ufi. + --- a .- i DOCKET ~N $ '
50'-254' w -.- .- '
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UNIT ONE-
.J ATEJanuarv-D 03 I 1983' .. ,
. . COMPLETED - BYRan' doll' D ' Buss
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, _-TELEPHONE 309-654-224ixi81-- _ =
-.. - . N.HOT. H . December 1982 - -. .. , _ . ~ .
BAY AVERAGE. DAILY POWER LEVEL. DAY' AVERAGE.LbAILYPOWER'LEVEb'
_ _s. ._
-.jMWe-Net)1 _,__ _ _ _ ___ _
(MWe-Net). *
- 1. -2.5 17 . . . -0.6
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2.
-2.5 18. -9 .' O
. ._ _3 . . ~-4.0 .__ . s ._1.9i. . -10.7 ._
- 4. -5.0 20. -10.4
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'6..
-7.6 _ , . __ _2 2 . . -15.9 __
- 7. -6.2 23. -
64.8
- 8. -5.8 24.. 176.6 9.:
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-6.0 __ _ _ _ _ _._.25.. 126.4 _ _ _
- 10. -6.0 26. 115.6
- 11. -0.0 27. -12.0
._ .12 . . -7.7 - ._.28. 321.5 . . _ .
- 13. -0.1 29. -
382.8
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- 14. -0.7 ,
- 30. 490.8
_ 15. - 8 '. 3 31, 592.6
- 16. -0.4 INSTRUCTIONS ~
~ 0n this forn, list th everage dolly seit power level in HWe-Met for each der in the repeting senth Conpete to the Gerest whole negemett.
These figwes will be sted to p}et a graph tw each repeting neeth. Note thet when notinen dependable cepetite is - - -
esed for the net electrical rating of the salt thre ney be ecceslens when tk daily everage power leetl exceeds the i 100% line (or the restricted power level line) In suh cases,the average daily enit power setpet sheet shoold h feetnoted to esplein the apparent onenely L. _.. , . - ._ ._
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. " APPENDIX B ...
. AVERAGE DAILY ' UNIT POWER ? LEVEL
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' DOCKET;NO.- 265' ,
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..-w.. a -.- - .=.--.,m 2.._ .- -. .m ._.._u._.,._____._ 'DATEJonuorv 03'1983 ___._
, !COMPLETEDLBYRandall D-Buss
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1 TELEP HONE 1Q.9-654-2241 x i 8i -
._iHONTH . .
Decemb'er i982~ _ __ ____._,y., _a._,_._._
BAY AVERAGE DAILY POWER LEVEL. DAY: AVERAGE ~ DAILY' POWER ~ LEVEL
.. _ __ . ~-(MWe-Net)- _ . . _ - . . . . _ . . _ . _ _ _ . _ - -.___ _ .. ( MWe Net,>'- -.~. _*._... - _
- 1. 749.9 17. 768.3
~
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. . . , . . - - . . -755.2'
-18,
. . . .2. 1 700.1
- 3. 763.O _ . : ._ _ - 19 , _ 734.8 .__ 2 _
- 4. 761.7 20, 735.2
- 5. 746.7 '. 21. 715.6 6.' 770.5 . _ _ . _ ._ i 2 2 ., 698.7 . . _
- 7. 757.3 -23. 762.8 _.
. . - . . - . - . . . . . _ - -..... . _ . . _ . ~ . .-. .. _ _ - .u - - - - ,- -
- 8. 772.1 24. 569.i-
- 9. 775.4 . _. .25'. . 539.i- _.
- 10. 760.0 26. '623.3
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. .y 12 . 672.3 . _ 28.. 630.3 . .
13, 753.1 29, 718.0
- l. 4 . - 771.7 30. '674.5
- 15. 757.9 . _31. 611.6 .
16, 765.5 INSTRUCTIONS On this forn, list the surege daily unit power level in inle-liet for each dat in the reporting enth.Coupete to the
- neerest whole negewett.
Theu figeres will be oud to plot e graph for each reporting nenth. Note that when notinen dependeble copecite is-used for the net electrical rating of the soit there not be occasions when the delly everage power level exceeds the I illt line (or the restricted power leul line),In suh cases,the awrege dolly enit power estyst sheet sheeld be l, . . , . . festnoted to explain the apperent onently
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M M M M M M M ! M M M M M M M M y ID/SA APPENDIX D QTP 300-S13 UNIT Sl!UTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO. 050-254 August 1982 UNIT NAME Ouad-Cities Unit One COMPLETED BY R. Buss DATE Janua ry 3. 1983 REPORT HONTil DECEr:BER 1932 TELEPil0NE 309-654-2241-
~ 5 g m x e$# 3
$c $ LICENSEE P$ y$
o o @ mo ga h"m DURATION w M p EVENT $" o" No. DATE REPORT No. u (if0URS) CORRECTIVE ACTIONS / COMMENTS o
02-85 820906 S 514.9 C 4 RC FUELXX Continuation of Cycle Six Refueling Outage 82-06 021222 S 24.5 B 9 CA ZZZZZZ Turbine-Generator tripped while Reactor maintained thermal power to deteriorate a rubber shoe cover in the Reactor vessel.
32-27 821224 F 0.0 A 5 IF !!!STRU Reduced load dee to Traversing in-Core Probe problems 32-08 021226 F '26.7 0 2 ,
IF 1lSTRU Unit shutdown to perform maintenance on Traversing in-Core Probes 82-39 021229 F 0.0 I3 5 ID !!!STRU Load reduced to perform maintenance on Local Power Range ftonitor instrumenta-tion APPROVED AUG 101982 (final) ggg3g
m M P""1 M M M M M M M M M M M M M M y ID/5A APPENDIX D QTP 300-S13 UNIT SIIUTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO. 050-265 August 1982 UNIT NAME Quad-Cities Unit Two COMPLETED LY R. Buss DATE January 3, 1933 REPORT MONTil DECEt1CER 1902 TELEPil0NE 309-654-2241 N $ g w" 5 $E9 gm Wm g@ M LICENSEE m@ @
w DURATION h" "g EVENT $ o u
NO. DATE (110URS) o REPORT NO. CORRECTIVE ACTIONS / COMMENTS o
02-92 321205 5 0.0 3 5 I!A xxXX::x Reduced load to perform weekly Turbine tests 02-93 321211 S 0.0 il 5 RS C0::R03 Load reduced for Control Rod Pattern adjustment 82-94 021219 S 0.0 B 5 1: A . xxxxx:: Reduced load to perform weekt Turbine tests 82-95 321220 5 0.0 F 5 EA ZZZZZZ Load reduction requested by Load Dispatcher due to low system demand 02-96 821221 S 0.0 F 5 EA ZZZZZZ Load reduction requested by Load Dispatcher due to low system demand U2-97 821222 S 0.0 F 5 EA ZZZZZZ Load reduction requested by Load Dispatcher due to low system demand 02-90 22122h S 0.0 F 5 EA ZZZZZZ Load reduction requested by Load Dispatcher due to low system demand 82-99 821225 S 0.0 F 5 EA ZZZZZZ Load reduction requested by Load Dispatcher due to low APPROVED system demand AUG 1 G 1982 (final)
VCUSH l
M P"I m m M M M M - M M .
P?1 M P"4 W c
ID/5A APPENDIX D QTP 300-S13 UNIT SIIUTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO. 050-265 August 1982 UNIT NAME Ouad-Cities Unit Two COMPLETED BY R. Buss J nuary 3, 1983 DECE!GER 1932 DATE REPORT HONTH TELEPHONE 309-654-2241 N $ g w" 5 EE gw !?w
- $ Q S d LICENSEE w@
~
w DURATION M N EVENT $" {@
o" NO. 'nTE J REPORT NO. "
(HOURS) CORRECTIVE ACTIONS / COMMENTS ca 32-100 821220 S 0.0 F 5 EA ZZZZZZ Load reduction' requested by Load Dispatcher due to low system demand 82-101 821230 F 0.0 11 5 HC llTEXCil Reduced load due to high Condenser Backpressure 32-102 821231 S 0.0 F 5 EA ZZZZZZ Load reduction requested by Load
- Dispatcher due to low system demand APPROVED AUG 101982 (fina1) V. C. O. S. H
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% .a .' . ' :.;;- ~'- l 8 .- . J . I'~.h ?.N % 4 " , , . .
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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 Description Unit Date Actuated Actuations Conditions of Events 1 12-22-82 1-203-3A 1 Hanual Rx Press Surveillance 1-203-3B 1 Manual 945 T.S. 4.5.D.I.b l-203-3C 1 Manual 1-203-30 1 Manual 1-203-3E I Manual l
9 a-B. CONTROL ROD DRIVE SCRAM TIMING IRTA POR UNITS ONE' AND WO - 1 l
I The basis for reporting this data to the Nuclear Regulatory
- Commission are specified in the surveillance requirenents of -
Technical ^ Specifications 4.3.C.1 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 scran timing was performed with reactor pressure greater than 800 psig.
.l 1
1 i
RESULTS OF SCRAM TIMING MEASUREMENTS PERFORMED ON UNIT I & 2 CONTROL ROD ORIVES, FROM I-I TO 12-31-02 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) 12-18 177 0.22 0.46 0.95 1.69 1.91 Unit 1 A & B Sequence Cold Prior (ll-1) to Startup 12-24 177 0.27 0.65 1.43 2.52 3 02 Unit 1 A & B Sequence Hot During (J-10) Startup a
m
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VII, REFUELING INFORMATION
~
The following information about future reloads at Quad-Cities Station was requested in a January 26, 1978, licensing memorandum (78-24) from D E, O'Brien to C. Reed, et al. , titled "Dresden, .
' Quad-Cities, and Zion Station--NRC Request for Refueling Information",
dated January 18, 1978.
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l
L ~ ~
QTP 300-S32 6~ ,~ - .
Revision 1 j, QUAD-CITIES REFUELING March 1978 l6 l
l INFORMATION REQUEST .
l p' 1. Unit: 1 Reload: 6 Cycle: 7 l
- 2. Scheduled date for next refueling shutdown: Sept 12, 1982
{
3 Scheduled date for restart following refueling: Dec 4, 1902 -
c-
- 4. Will refueling or resumption of operation thereafter require a technics 1 specification change or other license amendment:
I" YES u
l 5 Scheduled date(s) for submitting proposed licensing action and support!ng C . Information:
- JULY 26, 1982 r,
- 6. Important licensing considerations associated with refueling, e.g., new or l*
' different fuel design or supplier, unreviewed design or performance analysis
,., [ methods, significant changes in fuel design, new operating procedures:
o ::
'litPLEttENTATION OF TliE ODYN TRANSIENT ANALYSIS CODE AND RESULTS (MCPR SCRAM TlHE DEPENDENCE)
T -
/
9 .
r l
7 The number of fuel assemblies.
[ a. Number of assemblies in core: 224 new/724 total 5-after the
- b. Number of assemblies in spent fuel pool: outage 1940 r.
L 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: 2920
- b. Planned increase in IIcensed storage: 4636 new/7556 total 9 The projected date 'f the last refueling that can be discharged to the spent fuel pool assuming the present IIcensed capacity:
LOSS OF FULL CORE DISCHARGE CAPABILITY - 3/04 y LOSS OF RELOAD CORE DISCHARGE CAPABILITY - 2/86 A P P R.O V E D l
APR 2.01978 W '
. Q.C.O.S.R.
[.. .,
QTP 300-S32 f~ ~ ,
Revision 1 n* -
1-QUAD-CITIES REFUELING H:rch 1978 INFORMATION REQUEST u~ 1. Unit: 2 Reload: 6 Cycle: 7 .
l n 2. Scheduled date for next refueling shutdown: Feb 27, 1983 3 Scheduled date for restart following refueling: April 23, 1983 -
I"
. 4. Will refueling or resumption of operation thereaf ter require a technical specification change or other license amendment:
' NO
- 5. Scheduled date(s) for submitting proposed licensing action and supporting
. Information:
l' NONC r'
- 6. Important licensing considerations associated with refueling, e.g., new or l ~ ,, ' different fuel design or supplier, unreviewed design or performance analysis
,, / methods, significant changes In fuel design, new operating procedures:
).* fl0NE e
j _
~.
n i'
' 7 The number of fuel assemblies. -
f a. Number of assemblies In core: 192 new/724 total L after the
- b. Number of assemblies in spent fuel pool: outage 2132 v~
,, 8. The present licensed spent fuel pool storage capacity and the size of any increase in IIcensed storage capacity that has been requested or is planned
,, In number of fuel assemblies:
1"
- a. Licensed storage capacity for spent fuel: 2920
- b. Planned increase In licensed storage: 4636 new/7556 total 9 The projected date of the last refueling that can be discharged to the spent fuel pool assumin9 the present licensed capacity:
LOSS OF FULL CORE DIStilARGE CAPADILITY - 3/0;.
i v LOSS OF RELOAD CORE DISCHARGE CAPABILITY - 2/86 APPROVED APR 2.01978 Q.C.O.S.R.
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,m VIII. - CLOSSARY .
The following abbreviations which ~ may haveI been us'ed in = the Month 1'y Report, are defined below: '
ACAD/ CAM : ~ ' Atmospheric Containment- Atmospheric Dilution /Containnent - <
Atmospheric-Monitoring . . ,
ANSI- - - American National Standards Institute APRM - Average Power Range Monitor.
ATWS- -- Anticipated Transient Without Scram' BWR -
Boiling Water Reactor CBD -
. Control Rod. Drive EHC - ' Electro-Hydraulic Cont'rol. System Bnergency OperationsLFacility EOF . -
GS EP - -
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 - Inteemediate Range Monitor ISI -
Inservice Inspection LER' - Licensee Event Report LIRT -
Local Leak Rate Test LPCI - Low Pressure Coolant Injection Mode of RHRS LPRM - Local Power Range Monitor .
MAPLHCR - Maximum Average Planar Linear Heat Generation Rate MCPR- - Minimum Critical Power Ratio MFLCPR - Maximum Fraction Limiting Critical Power Ratio MPC - Maximum Permissible Concentration MSIV -
Main Steam Isolation Valve NIOSH - National Institute for Occupational Safety and Health PCI - Primary Containment Isolation PCIOMR - Preconditioning Interim Operating Management Recommendations RBCCW - Reactor Building Closed Cooling Water System RBM - Rod Block Monitor RCIC - Reactor Core Isolation Cooling System RHRS - Residual Heat Removal System RPS - Reactor Protection System RWM -
Standby Cas Treatment System
-SBLC - Standby Liquid Control SDC. - Shutdown Cooling Mode of RHRS SUV -
Scram Dis harge Vc9ume SRM - Source Range Monitor TBCCW - Turbine Building Closed Cooling Water System TIP - Traveling Incore Probe TSC - Technical Support Center
-