ML20059F469

Monthly Operating Repts for Aug 1990 for Quad-Cities Nuclear Station Units 1 & 2

ML20059F469
Person / Time
Site:	Quad Cities
Issue date:	08/31/1990
From:	Hamilton L, Robey R COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RAR-90-66, NUDOCS 9009110218
Download: ML20059F469 (22)

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+ Commonwealth Edison; k *:*i; "" / ^, \T ' ound CitXs Nuclear Power Station

-* -- d ) 22710 206 Avenue North r . k.@ - Cordove, Illinois 61242 9740

. Telephone 309/654 2241 -

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RAR-90-66. -

September'4, 1990

, U. S.-Nuclear Regulatory Commission

• . ' ATTN: Document Control Desk Washington, D. C.- 20555 .

SUBJECT:

Quad Cities Nuclear Station Units 1 and 2 Monthly.PerformancezReport

'NRC Docket Nos. 50-254 and 50-265 1 ,

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% Enclosed for your information is the Monthly Performance Report covering' '

.the, operation of Quad-Cities Nuclear Power Station, Units One and Two,- -,

o :during:the* month of August, 1990.

L _ Re's'pectfully, v ..

. COMMONWEALTH EDISON COMPANY

>% QiJAD-CITIES NUCLEAR' POWER STATION-

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'N7 l:f;.hlos y" (.. R A. Robey- .

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-Technical. Superintendent

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4; cc:; A.B. Davis, Regional Administrator.

T. Taylor, Senior Resident Inspector

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i TABLE'0F CONTENTS

i I .' Introduction f . ,. ' ~II: 1 Summary of Operating Experience

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. ".A.. Unit-One

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B. Unit Two S , . ;III._ Plant or Procedure Changes, Tests, Experiments, and1 Safety -

Related Maintenance Amendments to Facility License or Technical Specif'ications-A.

. ' So B. . Facility or Procedure Changes Requiring NRC Approvall C. Tests"and Experiments Requiring NRC Approval

}g D. Corrective Maintenance of Safety Related Equipment IV. . Licensee Event Reports an' .'

- 1 V. Data Tabulations n

A. Operating Data Report;  ;

o B. Average Daily Unit = Power' Level .!

C. Unit Shutdowns;and Power Reductions g-VI. Unique Reporting Requirements <

u . i E A. Main Steam Relief Valve 0perations B. Control Rod Drive Scram Timing Data

.f' VII. Refueling.Information

< ' VIII. Glossary .

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H W I. INTRODUCTION 8

Quad-Cities Nuclear Power Station is composed off two Botling Water Reacto'rs, each with a Maximum 0ependable Capacity of 769 MWe Net, located:in w

' Cordova, Illinois. - The Station is jointly owned by Commonwealth Edison -

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JCompany and Iowa-Illinois Gas & Electric Company. The Nuclear Steam Supply' -

3 Systems are General Electric Company Bolling Water Retctors. The j a  ;

Architect / Engineer'was Sargent & Lundy, Incorporated,:and the primary . I

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construction contractor'was United Engineers & Constructors. The Mississippi-! l q

River is the-condenser co'o ling water source. The plant is subject to ' license,. ,' .-j numbers DPR-29 and DPR-30, issued October '1,1971, and= March 21,1972,-

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1 respectively; pursuant-to Docket Numbers 50-254 and 50-265. The date of- )

initial Reactor criticalities for Units One'and Two, respectively were , ,

. . i E-October 18, 1971, and_ April 26, 1972. Commercial generation of' power began,on: H February 18.-1973 for Unit One and March 10, 1973 for Unit Two.  !

.e This-report was compiled by Lynne Hamilton and Verna Koselka, telephone i

number 309-654-2241, extensions 2185 and 2240. ,'i y

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, 'II.

SUMMARY

OF OPERATING EXPERIENCE c'

A.. -Unit One E

Unit One began the month of August operating at 250 MWe at the request of the Chicago Load Dispatcher. At 1150 hours0.0133 days <br />0.319 hours <br />0.0019 weeks <br />4.37575e-4 months <br />, the unit was taken to full power levels per the Load Dispatcher. On August 2, at 2206 hours0.0255 days <br />0.613 hours <br />0.00365 weeks <br />8.39383e-4 months <br />, the' unit was placed in Economic Generation Control'(EGC). On August 3, at 0745 hours0.00862 days <br />0.207 hours <br />0.00123 weeks <br />2.834725e-4 months <br />, the unit was taken out of EGC and load was increased to

-810 MWe. .From August 4 to August 9, the unit remained in EGC or operated L near full power with power levels being adjusted according to the demands ,

of the load dispatcher. Routine surveillances were performed.

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• k On August-9, at 0820 hours0.00949 days <br />0.228 hours <br />0.00136 weeks <br />3.1201e-4 months <br />, a power reduction to 250 MWe was taken for' Instrument Maintenance to test and repair the #1 Turbine Control Valve ,

after noting valve oscillations. At 1155 hours0.0134 days <br />0.321 hours <br />0.00191 weeks <br />4.394775e-4 months <br />, during troubleshooting .

of the #1 Turbine Control Valve, an unexpected RPS A 1/2 Scram was received- l f 1 while the.#1 control valve was being held closed with the test switch.- ~

The cause of the 1/2 Scram was unexpected relay chatter of the' Turbine Control Valve fast closure scram bypass. The 1/2 Scram was immediately-

. reset and ,an additional group of rods were inserted to reduce load and prevent relay chatter. Instrument Maintenance then tightened a loose t- wire on the local junction box terminal strip for the #1 control valve.

'A t 1808 hours0.0209 days <br />0.502 hours <br />0.00299 weeks <br />6.87944e-4 months <br />, the unit was taken to 797 MWe.

On August 11..at 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />, while testing electromatic relief valves g for a-scheduled HPCI inoperability the 1-203-3C electromatic relief valve .

P failed to open the first and second attempt. The valve opened and functioned  ;

normally on the third attempt. At 0735 hours0.00851 days <br />0.204 hours <br />0.00122 weeks <br />2.796675e-4 months <br />, tailpipe temperature on the 3C relief valve was greater than 330*F, indicating that the valveL '

may not have fully closed. It was-determined by Station Management to drop load..de-inert the:drywell,'and inspect the valve. It was discovered ~

that a' spring on the-valve actuator mechanism had come loose. .This spring ,

and another spring on the opposite side of the actuator armature keeps  ;

iL the armature in alignment while it is stroked. At 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br />, it was decided to bring the unit to cold shutdown to repair the valve. 'At l 2151 hours0.0249 days <br />0.598 hours <br />0.00356 weeks <br />8.184555e-4 months <br /> Unit One was shutdown. The unit remained' shutdown through August 17 while repairs were made to the 3C electromatic' relief valve. <

Preventive Maintenance was performed on various other systems as well. ,

On August 17, unit startup commenced,.and at 0842 hours0.00975 days <br />0.234 hours <br />0.00139 weeks <br />3.20381e-4 months <br />, the reactor became critical. . At 2301 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.755305e-4 months <br />, the' generator was synchronized to the grid.

An increase.to full power level was begun using control rods and recircu-lation pumps. On August 18, at 2315 hours0.0268 days <br />0.643 hours <br />0.00383 weeks <br />8.808575e-4 months <br />, full power was achieved.

For the remainder of the month, the unit remained in EGC or operated near full power with power levels being adjusted accordingly. Routine surveillances were performed. ,

B.. Unit Two Unit Two began the month of August operating at full power levels. Normal operational activities occurred for the month with power levels being adjusted according to the demands of the Chicago Load Dispatcher. The unit remained in Economic Generation Control or operated near full power levels with minor interruptions to perform routine surveillances.

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'lII. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS,-

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. AND SAFETY RELATED MAINTENANCE t> .i

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Amendments to Facility License-or Technical Specifications, G .

.,There1were no_ Amendments to the Facility License'or Technical Specifications -

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B.:, Facility or Procedure Changes Requiring NRC Approval' '

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[7!# -There.were no1 Facility or: Procedure changes requiring NRC approval for '

1 7 the; reporting 1 period.

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JE C. Tests and Experiments Requiring NRC Approval

• o Thereiwere.no Tests or Experiments requiring NRC approval for the j

.. reporting period. 4

-[D . Corrective Maintenance of Safety Related Equipment i

The following represents-a tabular summary of the major safety related; )'

maintenance. performed on Units One and-Two during the reporting period..

This summary includes the' following: Work Request Numbers, Licensee ,

LEvent Report Numbers, Components, Cause of Malfunctions. Results,and

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' Effects on Safe Operation,=and Action Taken to Prevent' Repetition.-

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SUMMARY

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idRK: REQUEST ' SYSTEM. EID DESf;RIPTION -NORK PERFORMED

- Q80846: . 0293 -Valve. 1 pare Electromatic2 Relief- 'As:found: ? Seat on piston (disc)-ruined by ultrasonic

- S/N BXO1293 ' cleaning. As.left: Relapped main seat; install 2dInew:

, disc,.. rings, guide;.assembledivalve;~ tested;on nitrogen.,

Q81937 '0293 - Valve, SparelElectromatic: Relief _, , Seal. welded disc retainers Lon electromat-ic relief: valvel S/N-BX00293- BX293.- * ~

- As found: Inspected:and:found'no problemsJon EPN's.:-

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Q86574 0203: Valve Electromatic Pilot that need to be: addressed at this time. .As leftF.'EPN' ,

solenoids will be rebuilt-under Q84018.

Q85893 1001 Valve RHR.lA Heat Exchanger Service As-found: Nut loose,.on valve stem to transition bl'ock..

Water Discharge Control' As left: Locticed _: and' tightened nut..

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IV. LICENSEE EVENT REPORTS

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, ,  ; The - following isL a tabular su"n;ry 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 1:. 6.6.B.l. and 6.6.B.2. of the Technical Specifications.

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- Licensee' Event 7, . Report Number , Date Title of Occurrence-

,  ; 90-16' 08/09/90 Failure of 1/2A Fire Pump to Start on Low Fire Header Pressure.

Also Failed to Start, Manually 17 08/11/90- Failure of HPCI Steam' Exhaust Check Valve 1-2301-45 ,

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' 18 08/29/90 'Both Units Potentially

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Outside' Design Basis for Divisional Separation UNIT'2 ,

Thety were no licensec e-< reports for Unit 2 for this reporting period.

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V. DATA TABULATIONS f ~.+ .;3, - .

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a. The following; data tabulations are presented in this: report: , ,

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• -B.' Average. Daily. Unit' Power Level- '

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f C. Unit Shutdowns and Power Reductions 1 g

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08EUTM E&ia FEF0Et h:ket Ns. 50254 tmit Or4 Date 5eftet:er 5, 199il C;t:lettd Iy Linct hatilttt Tele:hore Ki-154- ;4!

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1. Fe;otiegFerled't22,.,(8diR 8

Gress h:grs in Re:cet Pe icdt '44

2. terrently Auth:rlied Fcner Levt! Pith Gil Man. Ie:e*.d. Ca:a:ity (Meedeth ll]

~ hstgn Ele:trital Rating tht4eth lij

3. fewer level te ktich Kestricted (li Any) (Se-Net): MA 4 Frascis For Festriction (!f anyh THIE P MTH YE 10 (ME C' MAT!VE

5. herter ef Fws batter kas Critica! 617.2 55:!.E 124744.6 6.' Reacter Feserve Snddenn R ars 0.0 0.0 34:1.4

7. Fws 59neraier Cm Line 592.6 f546.9 1:5t37.5 0, !Jnit Faserve Shutican H urs 0.0 0.0 999.2 '

t. _ Grcss Trattal Energy Etneratt! Ph) 1345691.0 1 i!0EE0 h!O3200.0

20. Erces ElettHeal Energy Generated mh)

. 431111.0 4211766.0 57052325.0

!!. ht Electrical Ene gy 6enerates Ph) 416423.0 40467!2.0 915s9!!!.0

12. Reatter Service Facter E2.4 95.8'- B0.5

13. Les:ter evallatility Fatter ' $2.4 S!.E 82.6

14. Usit Service fact:- E 0.5 91.1 M.F

.15. LYit h allatility F6:ter E0.5 95.1 70.5

16. tmit Capacit,y Fatter f0 sic; GC) 72.9 E3 16.1

17. L' tit Ca;a: ty Factor (Unleg tesige nel 70.9 EE.0 64.4 .

-IB.UnitFo*cedOctageRate 4.4 1.9 5.3

'19. Ent: cans S:tetalef Over hot i Montti (Type, hte, art bratten of Ca:hh E

20. H ihut h un at End :f Re:crt 'e led, Esticated late cf Sta-ty : _ _ _ _ _ , , _ , , .

?!. thits in lest itates if ri;r to heter:ial C;traticch rere:Ast A:ble.te Initial Criticality- _____ _ _ _ _ _ _

IritialEle:tricity ..

Cco e":tal 0:eratico ______ _______.

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Date fe: tenter % 1990 Coe;1etedty' tytre Faeil e, Isle: tere 309 154-:241

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'! .. De;0 ftir? f eriod EM.fil9L__,, 6tcis ti:a's it se;0rt Periods !al .,

2. Curreatly Authorizef fewee Level %th Ril Mar. !a:e d. Capacity the Meth Id resigt tjectrical' Ratirg Pae-4t): Hi

!. fcerer lhel to ui:h Festrictec W Any) Pit-Net): h /,A,

4. Fea6Lis for f estriction (!f 4 y):

THIS O TH YR 70 I4TE CUM]LATIV!

5. N x er of Ecurs Featter Was Crftical 744.0 !51.2 12:535.6

4. Rea:t:r Feserve Sw td:ss F:.rs 0.0 0.0 2?i5.5 J. H:urs 6snerater Dr Line 744.5 3!03.4  !!9601.9

' E. Unit Reserve Statdann Hours 0. 0 = 0.0 702.9

3. Bress therni Energy Baneratic PiiSi 1 " 5116.4 76:441E.0 257051(35.0

10. Grecs Ele:trical Erergy Generated (PWal u4192.0 24i0240.0. !2429924.0

!!. het Electrital Energy tenerated 045) 54!053.0 2360734.0 77E60353.0-

12. Reactor Service Factor 100.0 10.4 - 77.0

. 13. Rea:tc* Availab:lity Fa:ter - 1C 0. 0 60.9 76.i '

14. Voit fervice Tatter 100.0 60.1 74.9

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5 W , l i 6 W I . m - O a 9 C H 0 / e n 3 S U I C n I y T L C A _ T S E . V . D E I E N T _ T O C E H E L P R 2_ P 2 I I M I O - O E C C T . e n o N . S E G I T C 0 go" - U 9 B 9 $3" e 1 M . R t E s ) DW u O c l XP u uS" a n I A D fmE . i f N ( A . - NSI A1 0 t m " ON 'T " NT 1 D l o 'ER F U N M T IVO L EPE S F R D T P . I E " I I R U o* I oh 5$E I) I CSt o Tts I , w T A0 t 0 t 9 N I (N + i 9 D n 1 U , s 5 " 5 n., e i r uh 5 t e 6 i b 2 C m - e E .- 0 d t T . 5 a p A u e D Q S O N M A . T N O - E . I T E N . C I T ^ O I S A ~ D I D I V >-q -} t ,ip" [{%v-VI .'- UNIQUE REPORTING REQUIREMENTS , :s The following' items'are included in>this report based on prior commitments E :to=the commission:: O F- . A.. Main Steam Relief' Valve Operations  ; . . -1 Relief. valve operations during the reporting period are sumarized in the -following table. The table includes information as to which relief valve l was actuated. how:it was actuated and the circumstances resulting in its - actuation.z j Unitt One; Date: August II, 1990  : Valves Actuate _d,d ' No. & Type of Actuation j 1-203-3A 1 Manual . .i 1-203-3B- l'Hanual  ! 1-203-3C 1 Manual 1-203-3D 2 Manual. + 1-203-3E 1 Manual ' i , Plant Cdnditions: Reactor Pressure - 989 PSIG , Description of Events: Routine surveillance Semi-annual, Manual Operationt of Electromatic Relief Valves (QOS 201-1), l,  ; , i-. -s Tech Spec: Ref. 3.5/4.5.D.l.a. u Unitt One ~ Dates August 17,-1990 Valves Actuated- No. 6 Type of Actuation [ t - a 1-203-3C 1 Manual- + l i 1-203-3D 1 Manual- , Plant; Conditions: Reactor: Pressure - 990 PSIG [ Description of-Events: . Operability Test (QOS 201-1),- l Tech'Spect Ref. 3.5/4.5.D.I.a . B'..= Control Rod Drive Scram Timing Data for Units One and Two > f

F P- Th'e' basis for reporting this data to the Nuclear Regulatory Commission-are specified in the surveillance requirements of Technical Specifications-U 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. i h 5 )

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.~ . _ , = . - _ . _ ~ ~ _m 0 _ "'.7" -~~ . . . ESULTS OF SCRARI TIMING EASLEEENTS , PEM0ftED ON tmiT 1&2 CONTROL R0D DR1VES, FR0tl 1-1-90 TO 12-31-90 - AVERAGE TIE IN SE0000S AT % ht4X. T RE IllSERTED FROM FULLY WIT N FOR 90'; INSERTION DESCRIPTION NLABER S- 20 50 90 Technical Specification 3.3.C.1 & DATE OF RODS 0.375 0.900 2.00 3.5 7 sec. 3.3.C.2 (Average Scram insertion Time) 1-22-90 1 0.27 0.65 1.42 2.52 N-10 Unit 1 Hot Scram Timing for test due (2.52) to 127 Valve Diaphragm Replacement . 1-31-90 1 0.29 0.64 1.34 2.36 R-7 Unit 1, Hot Scram Timing for test due (2.36) to 127 Valve Diaphragm Replacement 5-09-90 177 0. 10 0.68 1.45 2.54 H-11 Unit 2, Hot Scram Timing during Start _Up . (3.29) Sequence A&B, Cycle 11 6-21-90 88 0.29 0.66 1.4d 2.46 M-11 Unit 1, Hot Scram Timing for Sequence 8 (2.70) Cycle 11, 1st Sequence of Cycle 8-11-90 1 0.30 0.65 1.38 2.39 P-10 Unit 2. Hot Scram Timing test for (2.39) Accumulator Replacement l 0027H/0061Z _ ~ . . _ _ _ _ _ , _ . _a. - . , _ - ~ . . , - . . _ . _ . _ _ _ _ _ _ _ _ _ _ . .__ ___._._...m -. p, e ._ -- _= . - _ . .f..-. - , es

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1. 10'Brien to C Reed, et al., titled "Dresden,-Quad-Cities, and; Zion  !

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.a a. 4 . 1 i s .a m 3.- . . . ,a nt . >tr * ,' i rd g,e _ -) < Ns0 , J -J e QTP 300-532 Revision 2-QUAD CITIES REFUELING October 1989 INFORMATION REQUEST

1. Unit: 01 Reload: 10 Cycle: 11

2. Scheduled date for next refueling shutdown: 10-27-90

3. Scheduled date for restart following refueling: 1-4 .4.

Will refueling or resumption of operation thereafter require a Technical Specification change or other license amendment: Yes, a proposed change to Technical Specification will be made to relax the Minimum Critical Power Ratio (MCPR) scfety limit. This proposal is based on the Unit One Reload 11 Cycle 12 fuel loading.

5. Scheduled date(s) for submitting proposed licensing action and supporting information:

AUGUST 31, 1990

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:

i NONE AT PRESENT TIME. -l

7. The number of fuel assemblies. i

a.  !

Number of assemblies in core: 724 I

b. Number'ci' assemblies in spent fuel pool: 1537

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 ,

I planned in number of fuel assemblies:

a. Licensed storage capacity for spent fuel: 3657

b. Planned increase in t hensed storage: 0 9.- The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2008

• Ninety-six new fuel bundles have arrived on site and will be stored in the Unit One new fuel storage vault.

APPROVED i , 14/0395t  ! O.O.O.S.R. J .+ *- QTP 300-532 < Revision 2 i QUAD CITIES REFUELING October 1989 INFORMATION REQUE5T

1. Unit: 02 Reload: 10 Cycle: 11 j

2. Scheduled date for next refueling shutdown: i' 9-7-91

3. Scheduled date for restart following refueling:

12-9-91

4. Will refueling or resumption of operation thereafter require a Technical Specification change or other license amendment:

NOT AS YET DETERMINED. ' i p '5. Scheduled date(s) for submitting proposed licensing action and 1 supporting information: 1 NOT AS YET DETERMINED. 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: .s NONE AT PRESENT TIME. i

7. The number of fuel assemblies. '

a. Number of assemblies in core: 724

b. Number of assemblies in spent fuel pool: 2011 S. The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is p planned in number of fuel assemblies:

a. Licensed storage capacity for spent fuel: 3897 +

i

b. Planned increase in licensed storage: 0

9. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2008 h  !

APPROVED i (final) f 14/0395t W30M Q.C.O.S.R. s a: e-VIII. GLOSSARY I The following abbreviations which may have been used in th Monthly Report, l are defined below: ACAD/ CAM - Atmospheric Containment Atmospheric Dilution / Containment ' Atmospheric Monitoring ANSI - American National Standards Institute APRM - Average Power Range Monitor ATHS - Anticipated Transient Without Scram BWR - Bolling Water Reactor > 4 . CRD - Control Rod Drive EHC - Electro-Hydraulic Control System Em - Emergency Operations facility GSEP - Generating Stations. Emergency Plan l l- HEPA - High-Efficiency Particulate Filter ' HICI - High Pressure Coolant Injection System . HRSS - High Radiation Sampling System IPCLRT - Integrated Primary Containment Leak Rate Test > IRM. .- Intermediate Range Monitor ISI - Inservice Inspection

• lER ,- Licensee Event Report i.LRT -

Local Leak Rate Test I 'LPCI - Low Pressure Coolant Injection Mode of RHRS LPRM - Local' Power Range Monitor , l MAPLHGR - Maximum Average Planar L'near Heat Caneration Rate ' HCPR - 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 i 'RBCCW - Reactor Building Closed Cooling Water System , -RBM - Rod Block Monitor RCIC -- Reactor Core Isolation Cooling System RHRS - Residal Heat Removal System RPS - Reactor Drotection System RWM - Rod Worth Minim 12er a SBGTS - Standby Gas Treatment System i SBLC - Standby liculd Control $ SDC - ' Shutdown Ccoling Mode of RHRS SDV - Scram Disctarge Volume  ! SRM - Source Range Monitor TBCCW - Turbine Building Closed Cooling Water System TIP - Traversing Incore Probe TSC - Technical $1pport Center 0027H/0061Z