ML20066C129

Monthly Operating Repts for Oct 1982

ML20066C129
Person / Time
Site:	Quad Cities
Issue date:	11/01/1982
From:	Buss R COMMONWEALTH EDISON CO.
To:
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ML20066C105	List: ML20066C104 ML20066C129
References
NUDOCS 8211090356
Download: ML20066C129 (23)

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{{#Wiki_filter:._ _. 1 l I QUAD-CITIES NUCLEAR F0WER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT OCTOBER 1982 COMMONWEALTH EDIS0N COMPANY IOWA-ILLIN0IS GAS & ELECTRIC COMPANY NRC DOCKET NOS. 50-254 AND 50-265 LICENSE NOS. OPR-29 AND DPR-30 I t 9 .,e \\ 10 hfo t, w-,----~,,ww v4 w.-w --c a m w-- --w -vnen-, v m. e m, - - - -- e- -m-

TABLE OF CONTENTS 1. Introduction ll. Summary of Operating Experience A. Unit One B. Unit Two Ill. 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 Appr, val D. Corrective Maintenance of Safety Relateo Equipment IV. Licensee Event Reports V. Dat1 Tabulations A. Operating Data Report B. Average Daily Unit Power Level C. Unit Shutdowns and Power Reductions VI. Unique Reporting Requirements A. Main Steam Relief Valve Operations B. Control Rod Drive Scram Timing Data VII. Refueling Information Vill. Glossary 6 i

4 l. INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Eoiling i 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 Nuclear Steam Supply Systems are General Electric Company Boiling Water Reactors. The Architect / Engineer was Sargent & Lundy, incorporated, and the primary constructicn 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 reactor criticalities for Units 1 and 2, respectively, were October 18, 1971, and April 26, 1972. j Commercial generation of power-began on February 18, 1973 for Unit 1 i and March 10, 1973, fo r Un i t 2. This report was compiled by Becky Brown and Randall Buss, l telephone number 309-654-2241, extensions 127 and 181. l i } l I 5 l

ll.

SUMMARY

OF OPERATING EXPERIENCE A. UNIT ONE October 1-31: The unit continued its End of Cycle Six Refueling Outage throughout the month. B. UNIT TWO October 1-7: The unit began the month derated approximately 30 MWe due to high vibration in the 2A Recirculation pump Motor Generator Set. At 1600 hours, on October 1, the unit began decreasing load at 20 MWe/ hour while repairs were made to the High Pressure Coolant Injection System. The load drop was stopped at 0130 hours on October 2. The unit started increasing load at 0615 hours, reaching 780 MWe on October 3 October 8-16: On October 8, Unit Two began dropping load in p repa ra tion for a weekend Maintenance Outage. On October 9, the unit. was manually scrammed by putting the Reactor mode switch to SHUTDOWN. The Reactor was made critical again on October 12 and the unit went on line at 0343 hours on October 13 Between 0450 hours and 1530 hours, load was held at 200 MWe for control rod scram timing, before increasing at 100 MWe/ hour to 500 MWe for a control rod pattern change. Load then increased to 792 MWe by October 16. October 17-20: On October 17, at 0050 hours, the unit began decreasing 100 MWe/ hour to 700 MWe to perform weekly Turbine tests. At 0127 hours, the Reactor scrammed on an Average Power Range Monitor High-High signal due to a Condensate Demineralizer valve problem. The unit was critical at 0840 hours and on line at 1048 hours. Load increased to 550 MWe, where it was held for seven hours before increasing to 791 MWe by October 19 Load was then increased by withdrawing additional control rods without changing speed on the Recirculation pumps to 811 MWe; thus, eliminating the Recirc MG Set high vibration derating. October 21-31: On October 21, at 0950 hours, Unit Two began decreasing load 100 MWe/ hour in preparation for unit shutdown and was manually scrammed at 1702 hours by putting the Reactor mode switch to SHUTDOWN. Unit shutdown was required to repair the 33A Drywell to Suppression Chamber vacuum breaker, which stuck open during surveillance testing. The outage was extended to perform repairs to the M0-2-202-5A Recirculation Pump Discharge Valve stem. On October 25, the Reactor was made critical at 0438 hours and the unit was on line by 1145 hours. Af ter testing the repaired valves, load was increased to approximately 810 MWe by October 29 At 2300 hours, on October 30, load was reduced i to 700 MWe to perform weekly Turbine tests. In the interim, the Nuclear Engineer decided to reduce load further, to 655 MWe, to change control rod pattern. During the load increase to full power, the "B" Recirculation Motor Generator Set tripped on low oil pressure causing load to drop to 313 MWe. At the end of the reporting period, the unit was increasing load 5 MWe/ hour from 460 MWe, after restarting the 2B Recirculation Pump.

Ill. PLANT OR PROCEDURE CHANGES, TESTS, EXPERlHENTS, AND SAFETY RELATED MAINTENANCE A. Amendments to Facility License or Technical Specifications On August 6, 1982, Amendments 81 and 75 were issued to licenses DPR-29 and DPR-30, respectively. These a,nendments replaced the Appendix B (non-radiological reporting require-ments) with reporting requirements based on the Station National Pollutant Discharge Elimination (NPDES) Permit. B. Facility or Procedure Changes Requiring NRC Approval There were no Facility or Procedure changes requiring NRC approval for the reporting period. C. Tests and Experiments Requiring NRC Approval There were no Tests or Experiments requiring NRC approval for the reporting period. D. Corrective Maintenance of Safety Related Equipment The following represents a tabular summary of the safety related maintenance performed on Unit One and Unit Two during the reporting period. The headings indicated in this summary include: Work Request Numbers, LER Numbers, Components, Cause of Malfunctions, Results and Effects on Safe Operation, and Action Taken to Prevent Repetition. i. ,m g e ._.m.. .y 7 y.g .-.-+7 .,-p s,.m,. wT e

~ m u UNIT ONE MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R. LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q21763 82-20/03L RCIC Area High The temperature The isolation logic A neve temperature Tempe ra t u re switch was defective. was still operable. switch was installed Switch 1-1360-16C and tested satisfactorily. Q17253 HFA Relay 100C The relay spool was The relay would still The coil spool was 590-1000 c racked. operate properly. replaced. Q22073 82-26/03L RHR Service The 1 ink seal was Leakage was found on The seal was tightened Water Vault found loose in the both seals during leak and the leak rate test. Penetration penetration rate testing. was performed satisfactorily. 4 I l lj i

W UNIT TWO MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R. LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q21253 82-17/03L HPCI Steam Line The flange gasket HPCI steam line flange Removed the old gasket; Flange 2-2305-was leaking. was leaking steam. cleaned surface and re-10' B installed new gasket. Q22000 82-18/03L Diesel Generator The heat exchanger Unit 2 Diesel The heat exchanger was (Unit 2) (6600) was fouled. Generator trips on replaced and the Diesel high temperature. was tested satisfactorily. Q20354 Reci rc Inboard The solenoid The isolation valve The pilot valve was Sample Valve operated pilot was still operable. replaced. A0-2-220-44 va l ve was wo rn. Q21724 82-25/0,'L "A" SBGT Heater The temperature The heater would trip Reset temperature trip Tempe ra tu re switch setting and require manual to 425 F. Trip 1/2-7541-was incorrect. resetting. IIA Q22009 82-17/03L HPCI Steam Line The flange gasket Steam was leaking Disassembled flange, Flange 2-2305-was leaking. f rom the flange. took out two flexi-10~ B tallic gaskets and re-assembled. Q22159 A0-2-1601-20B The solenoid valve The valve will not The solenoid valve was Reactor Bldg to was wo rn. close from the replaced. Torus Vacuum Con t rol Room. B reake r 1 i

UtilT TWO MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R. LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION-Q22222 82-17/03L llPCI Steam Supply The flange gasket Steam was leaking A new gasket was installed. Line Flange 2-2305-was found to be through the gasket. 10~ B leaking. Q21905 82-19/03L 2A Recirculation Found studs Valve will not close Repaired the stem Pump Discharge missing on valve f rom Cont rol Room. and tested. Val ve 110-2-202-5A and valve shaft was bent. Q22221 fiain Steam Line A steam leak was The low pressure The switches were Low Pressure found at a fitting isolation would not calibrated and the leaky isolation Channel on the "B" RPS clear until reactor fittings were tightened. 'B'595-103D channel pressure pressure was 920 psig. switch. Q22360 82-22/03L Suppression The shaft was The vacuum breaker was The shaft was lubricated Chamber to Drywell binding against stuck open. The unit and the valve was cycled. Vacuum Breaker the valve bushing was shutdown to repai r 2-1601-33A and packing. it. I 4

'l 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. l. and 6.6.B.2. of the Technical Specifications. UNIT ONE Licensee Event Report Number Date Title of Occurrence 82-31/03L 9-5-82 RCIC Inoperable 82-32/03L 9-5-82 HPCI Inoperable 82-33/03L 10-5-82 lAl 24 V Battery Failed Discharge Test 82-34/03L 10-13-82 Fuel Pool Monitor 82-35/03L 10-21-82 0.G. Isolation Less Than 15 Minute Failure UNIT TWO 82-17/03L 10-1-82 HPCI Inoperable to Repai r Steam Leak 82-18/03L 10-6-82

1. 2 Diesel Generator Trip - High Temperature 82-19/03L 10-21-82 SA Reci rc Pump Discharge Valve Failure 82-20/03L 10-13-82 Torus Level out of Specifications 82-21/03L 10-13-82 Turbine Pressure Gegulator Bypassed 82-22/03L 10-21-82 1601-33A Vacuum Breaker Suppression Chamber to Drywell Failed Open

i j t i t b i V. DATA TABULATIONS i l The following data tabulations are presented in this' report: 1 A. Operating Data Report i-l B. Average Daily Unit-Power Level l C. Unit Shutdowns and Power Reductions J 1 i ,i t I I 4-i i i 4 4 ) i i t i i 1 4 i i i i a 1 l'. _. _. _.- _.._. __. _ _ _ _.. _ _ _. _.....-._, _.

j OPERATING DATA REPORT -.. - ~ __._ UNIT ONE _ _j l DATENovember Oi 1982 ~ ~ ~ ~ - ~ -~~' "~~ ~ ~ - ~ ~ ~ ~ COMPLETED BYRondoll"D Buss ~~~~~ ~ ~ I ~~ ~ J TELEPHONE 309-654-2241xi8i OPERATING STATUS i. __-~ 0000 100182._ _ _ _, __j .._ Reporting period 2400 103182 Gross hours in reporting period: 1745 q __2. Currently autho.rized p_ower;. level..(MWt): 2511;M.ox. Depend copocl_t.y_ __ (MWe-Net): 769* Design electrical rating (MWe-Net): 789 ._ _3. Power level to which restricted (if ony)(HWe-Net): .,N A 4. Reasons-for restriction (if any): ~ imuloVive ~ Triis Month 'Yr.to Date ~ C 5_ Number.of. hours reactor _was_, critical _ 0.0 1. 5833.1_ 74932.2 _

6. Reactor reserve shutdown hours 0. 0.

0.0 3421.9'

7. Hours generator on l'ine'

~ ~ ~ ~ ~ ~ '"0.0 ' 5'777 T ~ 719 8.T ~ ~ i

8.. Unit _ reserve shutdown hours.__

0.0

0. 0_

909.2 9. Gross thernal energy generated (MWH) 0 11000483- ~i46058842 ' i0. Gross electrical energy generated (MWH)' ~ ~~ 3533514 ~ 47062147^ ~ ~ ~ 4 ii. He't electrical energy. generated (MWH)_ -11188 3196684 43780768 _

12. Reactor service factor 0. 0.

77.9 81.6

13. Reactor av011ob111ty foctor 0.0 79.9

~ 851~ ~ _14. Unit service _ factor 0.0 79.2, 78.3 _ f

15. Unit avc11ob111ty factor 0.0 79.2 79.3 L6. Unit capacity factor (Using HDC)

-2.0 57.0 62.0 -i 9 55.5 60.4 L7. _.Un i t capacitu factor.(Using Des.MWe).

10. Unit forced outage rate 0.0 1.5-6.7-

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

of report _ period, estimated date of startup _I2g__82

20. If shutdown at end

$1he llDC ney be lower then 769 Itie dering perleds of high onblent tenperatore doe to the thernal per(ornence of the sprey canel. $ UNOFFICIAL COMPANY MUMIGS ARE USED Ill THIS KPORT

OPERATING DATA REPORT ._.: UNIT TWO~ DATENovember 01 1982 ~ _TELEP. HONE 309-654-2241x181 __ OPERATING STATUS 0000 100182 ~ ~ ~ ~ ~ ~ i.' ~ ~ Reporting period 2400 103182 Gross hours 1i reporting perlodi" 745 ~ ~ ~ ~ ~ ~ _.2. Currently.-authorized power level.(M_Wt)i.2511 Mox. Depend _ capacity _ (MWe-Net): 769* Design electrical rating (MWe-Net): 789

3. Power _ level _to which restrict _ed(if,gny).(MWe-Net): NA __

4. Reasons for restriction (if any):

5. Number of hours _ reactor was critical 564.8 5955.8 70807.6 6, Reactor reserve shutdown hours 0.0 0.0 2985.8-

7. Hours generator on line

~ ~ '~~ 544.5 ~~ '5894.4" 68135.5 O_.. Unit. reserve shutdown hours._ 0.0_ 0.0_ 702.9 _

9. Gross-thermal energy generated (MWH) 1160398 13359154 141246237

~42491'05 ~ ~ ~44955345 ~ ~'

10. Gross electrical' energy generated (MWH)~ ~ '

368912 ~ ~ ~ ____11. Het electrical eneroy generated (MWH) 350438 4046220 42170804

12. Reactor service factor 75.8 81.~ 6 77.9

13. Reactor avo11obL11ty factor 75.8 81.6 ~

~ ^81'."2 ~ 14_. Unit service factor 73.1 80.8 75.0

15. Unit uvaliability factor 73.1-80.8 75.7

16. Unit capacity factor (Using MDC) 61.2 72.1 60.3

17. Un t capacity f. actor (Using Des.MWe) 59.6 70.3 58.8

10. Unit forced outage rate 15.5 17.5 9.3

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

20. If shutdown at end of report period, estimated date o f s t ar t u p ____ NA_______

• The HDC not be lever then 769 MWe dering perleds of high enblent tenperatere det to the thernal perfernece of the spray canel.

8tM0FFICIAL COMPANY NUMBERS ARE USED IN THIS REPORT

'7 APPENDIX B AVERAJE DAILY UNIT POWER LEVEL . ~. DOCKET NO. 50-254 ._..2 DATENovember 01 1982._ COMPLETED BYRandall-'D Buss ~~ ~~ TELEPHONE 309-654-224fxi81~~~ ~ MONTH October 1982 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1. -20.0 17. -16.3 '3. -18.5 19. -22.5 4. -18.2 20. -18.5 22. -10.0 _ 6._ -18.2 7. -18.3 23. -11.2 9 _. -ii.i 25. - -13.0 10. -3.9 26. -18.7-a

1. 2.

-4.3 28. -16.9 13. -11.0 29. -16.3 14. -16.0~ 30. ~~ ~ -18'.0 ~ ~ " ~ ~ ~ ~ _15. -17.0

31.,

-18.7 16. -15.7 ~ ~ INSTRUCTIONS On this forn, list the overage daily snit power level in llle-tiet for each day in the reporting nenth.Conpete to the nearest wlule negouett. These figeres will be sted te plot a gresh for each reportina nenth, liete that when notinen dependable copocitt is syd for the net electrical rating of the onlt there nay be 6ccostecs when the daily overage power level exceeds the inst line (or the restricted power level line),In sech cases,the overage dolly enit power evtput sheet sheeld be festnoted to esplein the apperent enently

APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET No. 50-265 DATENovember 01 1982 _ _ COMPLETED BYRundoll'D Buss _.__ MONTH October 1992 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL . MWe-Net) __ ( ( MWe-Ne t ) i. 702.0

4. 7.

295.5 -.- -. ~... 3.. 731.3 .i 19. _ 744.1 4. 727.8 20. 767.1 .6.__ 725.6 22._ -- 10. 8 ' 7. 730.1 23. -10.5 .9. -9.5 _ 25.. 1,83.8 10. -7.2 26, 512.7 m.m~ ,m 712.4 . 12. -B.3 _28.. 13, 214.0 29. 761.6

.._..15..

749.0 31. 518.2 i 16, 652.5 On this fern, list the euroge daily enit pwer laul in IWe-Het for each det in the reporting nenth.Cenpete to the nearest uhele negouett. . These figeres will be esed to plot a grap'h for each reporting nenth. Note that uhen noninen dependable copocite is. .~. 190% line (or the restricted power leetl line),tnere ney be occasions when the delly eseroge power level ex esed for the net electrical rating of the snit .In sech cases,the enrege dolly enit power evtpet sheet shaald be festnoted to esplein the apparent onenely a S -,. -.. ~

C ID/5A APPENDIX D QTP 300-S13 UNIT SIIUTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO. 050-254 August 1982 a. UNIT NAME Quad-Citics Unit 1 COMPLETED BY R. Buss ext. 181 N w er 1, 1982 REPORT HONTil OCTOBER 1982 TELEPil0NE 309-654-2241 DATE N H w "' 8 SEE g@m $w y S M LICENSEE e EVENT w DURATION M Eg o o V NO. DATE (IlOURS) REPORT NO. CORRECTIVE ACTIONS / COMMENTS o c2 82-85 820906 s 745:00 C 4 RC FUELXX Continuation of Cycle Six Refueling Outage l l APPROVED AUG 161982 , (final) ygg3g

m n m n m M EMS m EE"5 m M M M M n N f ID/5A APPENDIX D QTP 300-S13 UNIT SilUTDOWNS AND POWER REDUCTIONS Revision 6 DGCKET NO. _0_50-265 August 1982 UNIT NAME Quad-Citles Unit 2 COMPLETED BY ~~ ext. 131 R. Buss DATE November 1, 1982 REPORT MONTil OCTOBER 1982 TELEPIIONE 309-654-2241 N eb g m n x $w w o o g w@ "c LICENSEE m DURATION EVENT u w o O NO. DATE (110URS) REPORT NO. CORRECTIVE ACTIONS / COMMENTS o 82-73 821001 F 0.0 B 5 82-17/03L SF PIPEXX Reduced load while performing repairs to liigh Pressure Coolant Injection System 82-74 821008 S 0.0 B 5 ZZ ZZZZZZ Load reduction prior i.o weekend Maintenance Outage 82-75 821008 S 100.3 B 2 ZZ ZZZZZZ, Scheduled shutdowr. for weekend Maintenance . Outage and Battery Discharge Test 82-76 821017 S 0.0 B 5 !!A XXXXXX Reduced load to perform weekly Turbine test 82-77 821017 F 9.3 A 3 IIG DEMl!!X Reactor scram on Average Power Range Monitor !!igh-liigh signal due to increase in Feedwater flow caused by Condensate Demineralizer valve problems 82-78 821021 S 0.0 B 5 82-22/03L SD VALVEX Load reduction prior to short Maintenance Outage to repair 33A vacuum breaker i APPROVED AUG 1 G 1982 1 (final) yCUbH

i M O O O M M M M M M M M O O O M s ID/SA APPENDIX D QTP 300-S13 UNIT SIIUTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO. 050-265 August. 1982 UNIT NAME Quad-Cities Unit 2 COMPLETED BY R. Buss ext. 181 DATE flovember 1, 1902 REPORT HONTH OCTOBER 1932 TELEP110NE-309-654-2241 w EU l N m x x w o o <c gw Q M LICENSEE us O O w DURATION M EVENT g" ou NO. DATE (110URS) REPORT NO. CORRECTIVE ACTIONS / COMMENTS o 82-79 321021 F 19.2 B 1 82-22/03L SD VALVEX Scheduled Maintenance Outage to repair 33A Vacuum Breaker 82-80 321022 F 71.7 B 4 82-19/03L CB VALVEX Outage continued to repair SA Recirculation Pump Discharge Valve stem G2-Ul 821030 S 0.0 B 5 llA XXXXXX Reduced load to perform weekly Turb'ine test 82-82 821031 F 0.0 A 5 CB lilSTRU Load reduction due to "B" Recirculation MG Set trip on low oil pressure APPROVED AUG 1 G 1982 - (final) y g,9 g g

VI. UNIQUE REPORTING REQUIREMENTS The following items are included in this re-ort 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 2 10-9-82 2-203-3A I Manual Rx Press Surveillance 2-203-3B 1 Manual 820 T.S. 4.5.D.l.b 2-203-3C 1 Manual 2-203-30 1 Manual 2 10-12-82 2-203-3E 1 Manual Rx Press Post-215 Maintenance (replace valve) 2 10-12-82 2-203-3A 1 Manual Rx Press Surveillance 2-203-3B 1 Manual 215 T.S. 4.5.C.2 2-203-3C 1 Manual (HPCI Out of 2-203-3D 1 Manual Service) B. Control Rod Drive Scram Timing Data for Units One and Two The basis for reporting this data to the Nuclear Regulatory Commission are specified in the surveillance requirements of Technical Specifications 4.3.C.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 scram timing was performed with reactor pressure greater than 800 psig. ~

j] RESULTS OF SCRAM TIMING MEASUREMENTS PERFORMED ON UNIT 1 s 2 CONTROL ROD DRIVES, FROM l-1 TO 12-31-82 AVERAGE TIME IN SECONDS AT % Max. Time INSERTED FROM FULLY WITHDRAWN For 90% insertion DESCRIPTION NUMBER S 20 53 90 Technical Specification 3.3.C.1 s 3.3.C.2 (Average Scram insertion Time) j DATE OF RODS 0.375 0.900 2.00 3.5 7 sec. 10-13 89 0.29 0.67 1.44 2.55 2.95 Unit 2 Hot Scram Time (D-10) "B" Sequence

Vll. REFUELING INFORMATION The following information about -future reloads at quad-Cities Station j was requested in a January 26, 1978, licensing memorandum (78-24) from D. E. O'Brien to C. Reed, et al., ti tled "Dresden, Quad-Ci ties, and Zion Station -- NRC Request for Refueling Information", dated January 18, 1978. l 1 5 s---ep ---mww w- --rm e-- -~-y-y u ,,..,,,~- -- - - - - -- p- +- w- ~ vv

QTP 300-S32 n R2 vision 1 QUAD-CITIES REFUELING March 1978 i I (*/ INFORMATION REQUEST ~ 1. Unit: 1 Reload: 6 Cycle: 7 m 2. Scheduled date for next refueling shutdown: Sept 12, 1982 3 Scheduled date for restart following refueling: Dec 4, 1982 r-{ 4. Will refueling or resumption of operation thereaf ter require a technical specification change or other license amendment: YES a 5 Scheduled date(s) for submitting proposed IIcensing action and supporting F Information: JULY 26, 1982 6. Important IIcensing considerations associated with refueling, e.g., new or ' different fuel design or supplier, unreviewed design or performance analysis ./ methods, significant changes in fuel design, new operating procedures:

• i.

" ~ litPLEf1ENTATION OF THE ODYN TRANSIENT ANALYSIS CODE AND RESULTS (tiCPR SCRAM TIME DEPENDE!1CE) a' 3~ m 7 The number of fuel assemblies. e a. Number of assemblies in core: 224 new/724 total after the b. Number of assemblies in spent fuel pool: outage 1940 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 6 9 The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: LOSS OF FULL CORE DISCHARGE CAPAB'LITY - 3/04 ' g p> p Ft () T/ E E) / LOSS OF RELOAD CORE DISCHARGE CAPABILITY - 2/86 APR 2.01978 C2.c:.C).55.Ft.

i, QTP 300-S32 fT QUAD-CITIES REFUELING March 1978 R2 vision 1 ] -, Q INFORMATION REQUEST l l ~ 1. Unit: 2 Reload: 6 Cycle: 7 i u n 2. Scheduled date for next refueling shutdown: Feb 27, 1983 3 Scheduled date for restart following refueling: April 23, 1983 r-{ 4. Will refueling or resumption of operation thereaf ter require a technical specification change or other IIcense amendment: g l' NO 5. Scheduled date(s) for submitting proposed IIcensing. action and supporting { Information: 3-N0llE r-6. Important licensing considerations associated with refueling, e.g., new or ' different fuel design or supplier, unreviewed design or performance analysis ~ methods, significant changes in fuel design, new operating procedures: / R l- ~ NONE YV U-n ~ l L. i, 7 The number of fuel assemblies. l l' a. Number of assemblies in core: 192 new/724 total after the u b. Number of assemblies in spent fuel pool: outage 2132 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 of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: LOSS OF FULL CORE DISCHARGE CAPABILITY - 3/84 WPPROVED v LOSS OF RELOAD CORE DISCHARGE CAPABILITY - 2/86 APR 2.01978 m. Q.C.O.S.R. m i

i .O-VIII. CLOSSARY The following abbreviations which may have been used in the Monthly Report, are defined below: ACAD/ CAM Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring ANSI American National Standards Institute APRM Average Power Range Monitor ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor CRD Control Rod Drive EHC Electro-Hydraulic Control System EOF Emergency Operations Facility GS EP Generating Stations Emergency Plan HEPA High-Ef ficiency Particulate Filter HPCI High Pressure Coolant Injection System HRSS - - High Radiation Sampling System IPCIAT Integrated Primary Containment Leak Rate Test IRM Intermediate Range Monitor ISI Inservice Inspection LER Licensee Event Report LLRT Local Leak Rate Test' LPCI Low Pressure Coolant Injection Mode of RHRS Local Power Range Monitor LPRM MAPLHCR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MFLCPR Maximum Fraction LLmiting 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 Rod Worth Minimizer SBGIS Standby Cas Treatment System SBLC Standby Liquid Control SDC 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 4 i i 4 --r}}