Text

Monthly Operating Rept for Aug 1979
	ML19208A821
Person / Time
Site:	Quad Cities
Issue date:	09/10/1979
From:	Hannum D; COMMONWEALTH EDISON CO.
To:	;
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ML19208A818	List: ML19208A817; ML19208A821;
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	NUDOCS 7909170561
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.

QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT AUGUST 1979 COMMONWEALTH EDISON COMPANY AND 10WA-ILLIN0ls GAS & ELECTRIC COMPANY NRC DOCKET N05. 50-254 and 50-265 LICENSE NOS. DPR-29 and DPR-30 Eb[

7909170 1

TABLE OF CONTENTS

1. Introduction ll. Summary of Operating Experience A. 'Jni t One B. Unit Two 111. Plant of Procedure Changes, Tests, Experiments, and Safety Related Maintenance A. Ammendments to Facility License or Technical Specifications B. Facility or Procedure Changes Requiring NRC Approval C. Tests and Experiments Requiring NRC Approval D. Other Changes, Tests and Experiments 1, Facility Hodifications

2. Special Tests E. Corrective Maintenance of Safety-Related Equipment IV. License Event Reports V. Data Tabulations VI. Unique Reporting Requirements A. Main Steam Relief Valve Operations B. Control Rod Drive Scram Timing Data Vll. Glossary 3a7087

l. INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Boiling Water Reactors, each with a Maximum Dependable Capacity of 769 MWe net, located in Cordova, Illinois. The Station is jointly owned by Commonwelath Edison Company and towa-lllinois Gas & Electric Company. The Nuclear Steam Supply Systems are General Electric Company Boiling Water Reactors. The Architect / Engineer was Sargent & Lundy, Inc. 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 plait 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 1 and March 10, 1973 for Unit 2.

This report was compiled by David Hannum, telephone number 309-654-2241, ext. 179 Siii'OS8

II.

SUMMARY

OF OPERATj 0 EXPERIENCE A. Unit One August 1: Uni t One began the reporting period operating at 791 MWe.

August 2-11: Unit One held an average load of 775 MWe. Load was reduced to 650 MWe on August 4 for turbine testing and condensate demineral-Izer backwashing.

August 12: Load was reduced to 500 MWe for main condenser flow reversal.

August 13: Uni t One held an average load of 766 MWe.

August 14-15: At 0418, the Unit One Reactor scrammed from a loss of main condenser vacuum. Prior to the scram, a gradual loss of vacuum caused load to be reduced to 270 MWe. A failure of air ejector 1B .

was suspected, and in-leakage was investigated over the subsequent weeks. At 1637 the reactor was made critical, with air ejector 1A in service. No vacuum problems were encountered de '99 unit s ta r tup. On August 15 at 0309 Uni t One was placed on-line and load was subsequently increased at the rate of 56 MWe/hr.

August 16-31: Unit One held an average load of 800 MWe. Load was reduced to 700 MWe for weekly turbine testing on August 19 and August 25 3:i"<'083

B. Unit Two August 1: Unit Two began the reporting period operating at 543 MWe.

August 2-31: Unit Two held an average gross load of 514 MWe. Load decreased throughout the month due to end of fuel cycle coastdown. On August 16 load was reduced to 330 MWe due to low system demand and reactor feedwater pump flow indication problems. On August 23, load was also reduced to 390 MWe for main condenser flow reversal.

357030

ll1. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS AND SAFETY RELATED MAINTENANCE A. Amendments to Facility License or Technical Specification The following amendments were added to the Technical Specifications during the reporting period:

Amendment 52 to DPR-29, and Amendment 49 to DPR-30.

These changes are as follows:

1. Paragraph 3.E Recirculation Loop inoperable is ueleted.

2. Paragraph 3.F Security Plan is renumbered 3.E.

3 Paragraph 3.F is added as follows:

3.F The licensee may proceed with and is required to complete the modifications indentified in Paragraphs 3.1.1 through 3.1.13 of the NRC's Fire Protection Safety Evaiuation (SE), dated July 27, 1979 for the facility. These modifications will be completed in accordance with the schedule in Table 3.1 of the SE and supplements thereto.

In addition, the licensee shall submit the additional information identified in Table 3.2 of this SE in accordance wi th the schedule contained therein. In the event these dates for submittal cannot be met, the licensee shall submit a report, explaining the circum-stances, together with a revised schedule.

The licensee is required to implement the administrative controls identified in Section 6 of the SE. The administrative controls shall be in effect immediately, except for those modifications indicated in Section 3.1 of the SE, which shall become effective on the dates indicated in Table 3.1 of the SE.

35'7031

Amendment 53 to DPR-29 and Amendment 50 to DPR-30.

These changes are as follows:

1. Technical Specification 4.B.2 is revised to indicate that the in-vessel sample program shall conform to ASTM E 185-66 and 10CFR50 Appendix H.

2. Table 4.6.2 is added to indicate the sample wi thdrawal schedule.

B. Facility or Procedure Changes Requiring NRC Approval There were no facility or procedure changes requiring NRC approval during the reporting period.

C. Tests or Experiments Requiring NRC Approval There were no tests or experiments performed during the reporting period requiring NRC approval.

D. Other Changes , Tests, and Experimen ts

1. Facility Modifications.

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M-4-2-75-73 RHR Service Water Vault Pumps Description of Modification This modification was to Install receiver / pump units in each of the RHR service water vaults. These units were designed to collect the seal water leakage from the RHR service water pumps and pump it into the service water discharge header. The intent of this modification was to reduce the drainage and humidity problems which had been experienced in the service water vaults.

Summary of Safety Evaluation The new receiver / pumps will reduce leakage into the vaults and prevent standing water from accumulatng thus improving system reliability. The new vault penetrations have been tested to assure leak tightness. The margin of safety as defined in the Technical Specifications is not reduced since the new system will improve vault drainage capabilities.

So/033

M-4-1(2)-77-8 RHR Service Water Vault Bulkhead Doors Pressure Test Taps Description of Modification These modifications involved installation of pressure test taps between the double gasket seals on the RHR service water vault bulkhead doors.

These test taps will enable the volume between the two gaskets to be pres-surized so that the seal area may be checked to verify the service water vaul t flood protection system in tegri ty.

Summary of Safety Evaluation The test tap is a passive component and does not change the door's structural arrangement or reduce the structural integrity in any way. The margin of safety as defined in the basis for the Technical Specifications is not reduced since the modification allows for a more efficient means of testing which meets the requirements of the Technical Specifications.

2. Special Tests.

Special Test 2-19 Unit Two Suppression Chamber -

Drywell Vacuum Breakers

Purpose:

The purpose of this test was to allow an evaluation for the drywell-torus vacuum breakers. This test documents an experimental change to the limit switch mounting bracket and valve operating arm in an effort to improve valve performance during surveillance testing. This test affects only valves A0-2-1601-32D and A0-2-1601-33A.

Summary fo Safety Evaluation:

This test was an effort to improve the performance of the vacuum '

breakers during monthly exercising. The operation of the vacuum breakers was not changed from that as given by the FSAR. The indicating and test circuitry remained unchanged as did the force necessary to open the valves. The limit switches and disc were the same.

E. Corrective Maintenance of Safety Related Eauipment 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.

367033

UNIT ONE ttA!NTENANCE SUMitARY .

CAUSE ltLSULTS 6 LFFEC15 OF ON ACTION TAKEN TO U.R. LER PREVENT REPETITION NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION SJAE Suction The solenoid valve The valve would not open. The o-rings were replaced Q00511 Valve o-rings were wo r. Mech. Vac. Pump was and the valve cycled 3 (A0 1-5401B) isolated. times.

SJAE Suction The solenoid valve The valve would not open. The o-rings were replaced Q00510 and the valve was cycled Valve o-rings were worn. Mech. Vac. Pump was (A0 1-5401A) isolated. 3 times.

Q00552 RHR HX Valve The torque switch The valve would not re- The thermals were re-(1-1001-4A) needed adjustn ent. open electrically. Re- placed and the torque dundant RHR paths were switch was readjusted.

available and the valve The valve was cycled 3 would open manually. times.

RilR HX Valve The torque switch The thermals tripped The torque switch was Q00551 (1-1001-186A) needed adjusting. when trying to operate re-adjusted and the valve valve. RHR was operable. cycled 3 times.

3067-79 U-l IRM #17 1RM #17 read low Redundant IRM's on that Repaired H.V. power sup-1-755 on range 10. RPS channel would have ply and adjusted pre-amp tripped had high flux overlap.

occurred.

itPCI Gland Seal The cammn field The pump tripped when The common wiring was 3800-79 Cooling Water wiring in the MCC trying to start. The repaired and the pump Pump (1-2301-57) was defective. HPCI turbine was still was tested.

operable. Pump used only during testing.

tiPCI Test Viv The tarque spring The thermals tr!pped when A grease line was in-3254-79 stalled to relieve hy-(1-2301-10) was l acki ng. trying to operate the valve. The primary node draulic lock on the of the HPCI system was torque spring. The valve operable at all times. was test operated 3 times.

The pressure The pressure transmitter The transmitter was (k

C 3806-79 ilPCI Press.

Transmitter transmitter was out was sending a low signal. re-calibrated.

J (1-2359) of calibration. IIPCI was operable.

'[ New bolts were installed.

hh 3B71-79 SBGT Inlet Screen Hold down bolts The screen was loose.

(1-5741) were missing from SBGT was still operable.

Ei the inlet screen.

UtilT ONE ttAINTEllNICE

SUMMARY

CAU5E RLSULlS E ElFECIS OF ON ACTION TAKEN TO U.R. LER PREVEhl REPETITICN IJUMBER NUltBER C0!tP0llENT liALFUNCT1011 SAFE OPERATION -.

3846-[9 itPCI Valve The valve operator llPCI was operable. The gasket was replaced (1-2301-49) gasket was de- and the valve tested.

fective.

3935-79 LPRM 48-41A The card was The LPRM read downscale The LPRM card was (1-756) defective. while bypassed. replaced.

Q00509 CRD 22-23 The valve was The valve was hot and The valve was replaced Scram Solenaid defective. making noise. and tested.

Valve (1-305-118) h103-79 Snubber on RCK The snubber was The snubber was loose The snubber was tightened.

Viv. (1-1301-60) loose. but still capable of its designed function.

Core Spray Test An aux contact was The valve would not open. The aux contact was 3620-79 Valve defective. Core Spray was operable. replaced.

(1-1402-4A) Normal position is closed.

3550-79 Of f Gas liigh Pressure switch The operator received The switch was calibrated Pressure Switch needed calibration. the off gas high pressure and tested.

(PIS-1-5441-16A alarm. The off gas dnd B) system was operating properly.

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UNIT TWO MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R. LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION 3551-79 79-13/03L RHR Suction Viv The thermals The valve would not open. The thermals were reset.

(2-1001-7C) tripped. The 2A, 2B and 2D pumps Amperage checks were made were operable and avail- and the valve test oper-able if needed. LPCI ated 3 times.

Mode was operable.

3598-79 RHR HX Bypass The thermals The bypass valve failed The thermals were reset.

Valve tripped. closed. RHR was still Amperage checks were made (2-1001-16B) operable with flow and the valve was cycled through the heat ex- 3 times.

changer.

3717-79 APRM #6 The slide wire was The chart was driving The slide wire was (2-750-100) shorted to ground. downscale. repositioned.

3722-79 79-14/03L RCIC Steam Supply The torque switch The valve would not The torque switch was Valve was defective. close. RCIC was still replaced. The aux con-(2-1301-16) operable, as was HPCI. tacts in the MCC were Isolation valve also replaced.

M0-2-1301-17 was operable.

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IV. LICENSEE EVENT REPORTS The following is a tabular summary of all license 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 Specifications.

UNIT ONE Licensee Event Date of Title of Report Number Occurrence Occurrence 79-25/03L 8-13-79 HPCi Area High Temperature Switch Drift UNIT TWO Licensee Event Date of Title of Report Number Occurrence Occurrence 79-16/03L 8-15-79 HPCI Area High Temperature Switch Drift 79-17/03L 8-22-79 Torus to Drywell Vacuum Breaker Division #1 Alarm Failure O

V. DATA TABULATIONS The following data tabulations are presented in this report.

A. Operating Data Report B. Average Daily Unit Power Level C. Unit Shutdowns and Power Reductions

c. . . ; e4 x Y'Y d .WJ Y

I

, OP EU.TI::G DnTA r.EP0xT CCCKET NO. 050-254 _

UlilT One lk COMPLETED BY D. Hannum

~

TEL EPliONE - (309) 654-2241, Ext. 179 OPEP ATit:3 STATUS 0000 080179

, J. Reporting period: 2400 083179 Gross hours in reporting period: 744

2. Currently authorized po.ier level (MWt) : 2511 Max. depend. capacity (MVe-::c t) : 769: Design electrical rating (n'.!e-t e t) : 789 -

3 Power level to which restricted (if any) (MUe- Me t ) : NA .

4. Reasons for restriction (if ny):

Y Cumulacive This Month Yr. to Date 5 Nuaber of hours reactor was critical 731.1 4754.5 52087.4 1

y 6. Reactor reserve shutdown hcurs 0.0 0.0 3329.6 7 Hours generator on line 721.2 4615.1 49558.0 _

J' 8. Unit reserve shutdown hours. 0.0 19.8 _999.2

' 1711778 10056740 98864616 9., Gross thermal energy generated (MUH)

-( 10. Gross cicetrical engergy generated (MWH) 549306 3191580 31764075 a

11. Net electrical Energy Generated 523758 3013483 29646056 y

.. 12. Reactor service factor 98.3 81.5 81.3

,_ 13 Reactor availabili f. factor 98.3 81.5 86.5

,(t ,

14. Unit service factor 96.9 79.1 77.4 3

15 Unit availability facter 96.9 79 5 78.8 r

Unit capactly factar (Us ing MDC)

16. 91.5 67.2 60.2 1 17 Unit capacity factor (Using Des. MWe) _

89.2 65.5 58.7 ,

18. Unit forced outage rate 3.1 3.6 _7. 8

, 19 Shutdowns sch:duled over n *xt 6 mcnths (Type, date, and duration of each):

20. I f shutdcwn at end of report period, estimated date cf startup;. NA

Th2 110C nay be lower then 763 MUc during periccis o' high ambiant temperature d. a ',

to the ther. mal performance of ti.a spray cc:,cl. , ,

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I 0F Ei.U !hG DATA T,EPChi COCKET f:0, 050-265 _

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l l DATE 9-5-79 COMPLETED BY D. Hannum TELEPHONE -(309) 654-2241, Ext. 179 OPERATING STATUS N00 080179 J. R:;:orting period:2400 083179 Gross hours in reporting period: 744

2. Cu rren tly au tho'-i .ed pc.ie r l evel (Mut) : 2511 Max. depend. capacity

, (MWe-Ne t) : 7600 Design electrical rating (U.'..'e- fic t ) : 789 .

e 3 Power level to which restricted (if any) (MWe-Mat) 7 NA .

4. Reasons fcr restriction (if any):

This Month Yr. to Date Cumulaiive t

5 Nunber of hours reactor was critical 744 5728.9 51105.3 J

6. Reactor reserve shutdown hours 0.0 0.0 2985.8 7 Fours generater on line 744 5682.7 48785.6 ,_

I 8. Unit reserve shutdown hours. 0.0 0.0 702.9 1270794 11462766 100204890 9., Gross thermal energy generated (MUH)

10. Gross electrical engergy generated (MWH) 382097 3550759 32039136 Net electr.ical Energy Generated 342893 3272260 30033720

.- 11.

. . _ . 12. Reactor service factor ,,,1,00. 0 98.2 80.9 13 Reactor availability factor 100.0 98.2 85.7

?(., 14. Unit service factor 100.0 o7.5 _77.3 L

15 Unit availability fccter 100.0 97.5 78.4 1

16. Unit capactly f actor (Using itDC) 59 9 73 0 61.9

17. Unit capacity factor (using Des. MWe) 58.4 71.1 60.3 -

i

-! 18. Unit forced outage rate 0.0 0.6 _9. 5

, 19 Shutdowns scheduled over nax: 6 months (Type, date, and duration of er.ch):

20. If shutdcwn at end of report period, estimated date ef startup:

NA

The !!DC may be lower thcn 769 MUc during periods of high a- biant temperature d.. ,

to the thermal parform.'rce of tha spray car.cl. , , , , , < !

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Docket No. ,050-254 Unit one Date 9-5-79

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l Completed by D. Hannurn Telephone (309) 654-2241, Ext. 179 Il0NTil August 1979 DAY AVERAGE CAlLY PO'.,'ER LEVEL' DAY AVERAGE DAILY POUER LEVEL (mle-fle t) (MWe-Net)

1. 731 17 754

2. 753 18, 755

3. 745 19 713

4. 651 20. 764

}5 _,_6 S 9 _

21 752

6. 723 22. 758

7. 752 23 760

8. 753 24. 763 l

t 9. 759 25 738

, 10, 745 26. 757 l

II. 757 , 27 759

12. 618 28. 764 13 748 29 753 14- 117 30. 758

15. 354 31. 754 a e e i< u v E D

16. 619

. JUN 2 01976 INSTRUCTIONS

) On this form, list the averag dii!y unit power !cve! in MWe Net for cath da, in the r: porting mch., rhkuh:OitIEie nessest w!w!c mecawatt.

, Rese Ggure.t will t.e thed to plut a paph for cuh reporting month. Note th at when nuxarnum dependsb!c cap; city is u:,ed 10: th, net c!ectrical r: ling of the unit, there may be ace:sions wiien the d.nly ascrage power level exce:ds the 1001 line (or ih. restri, red power !csel hne). In susn c.:ses, the averar: Jaily umt power cuipt.t sheet ;; hot.!d be Icuinoted to c.xplain the .ip;uren anoma!y.

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Coc!<e t flo. 050-265 Unit Two Date 9-5-79

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1 Telephone (309) 654-2241, Ext. 179 Il0fiTH August 1979 DAY AVERAGE CAlLY PO'.,'ER LEVEL DAY AVERAGE DAILY POWER LEVEL

(!tue-tie t ) (MWe-tie t )

1. 474 17 461

2. 487 18. 459

3. 478 19 456

4. 480 20, 456

} 5. 477 21 449

6. 467 22. 450

7. 500 23 448

8. 504 24. 449

9 485 25. 437

, 10. 468 26. 450

11. 473 , 27 439

12. 464 28. 437

13. 472 29 418

14. 473 30. 435 i 15. 470 31. 427

( aresuVED

16. 442 JUN 2 01976 INSTRUCTIONS

} On this forrn. !!st the average dsi!y unit power !cvel in .'. Wc Net for eachi da/ n the reporting nOthkuh:23ic'3 e nesicst who!c rnecawatt.

T.ese Ggurcs ' vill be used to plot a paph for cuch reporting mor:th. Note t!ut when nuximum dependable capacityis I

used for the net electoc.il r: ling ot' the unit there nuy be occasions when .he d.nly ascrage power loci exceeds the 1004 line (or the iestri. rett power esel 1:ne). In susn c.ises. the aver. ige J.n:y unit power outst.t sheet should !;c I footnoted to exj'l;in t!ic .spparent anonialy.

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APPENDIX D QTP 300-513 UNIT SHUTDOWNS AND POWER REDUCTIONS Revision 5 .

DOCKET No.

050-254 March 1978 U:llT NAttE Quad cities one COMPLETED BY D. Hannum DATE 9-5-79 REPORT HONTH August TELEPil0NE (309) 654-2241 Ext. 179 m 5 s m = eb z 5 E c: 8 g $ LICENSEE yM gg .

po' DURATION $ g{3 e EVENT *8 @: 8 tio . DATE (liOURS) C y5g REPORT NO, a g CORRECTIVE ACTIONS / COMMENTS R

17 790804 F --

H NA NA NA NA Load was reduced for turbine testing and condensate demineralizer backwashing.

18 790812 F --

H HA Nk NA NA Load was reduced to 500 MWe for main condenser flow reversal.

19 790814 F 22.8 A 3 NA NA NA Unit One scrarrened due to a loss of main condenser vacuum. Air ejectors were changed over, and no vacuum problems persisted during subsequent unit startup.

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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 There were no main steam relief valve actuations during the reporting period.

B. Control Rod Drive Scram Timing Data for Units One and Two There were no control rod drive scram timing exercises performed during the reporting period.

U d 8 ' .1. 'J f

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

CR0 -

Control Rod Drive System SBLC -

Standby Liquid Control System MSIV -

Main Steam isolation Valve RHRS -

Residual Heat Removal System RCIC -

Reactor Core Isolation Cooling System HPCI -

High Pressure Coolant Injection System SRM -

Source Range Monitor IRM -

Intermediate Range Monitor LPRM -

Local Power Range Monitor APRM -