ML19261D440

Monthly Operating Rept for May 1979

ML19261D440
Person / Time
Site:	Quad Cities
Issue date:	06/05/1979
From:	Hannum D COMMONWEALTH EDISON CO.
To:
References
NUDOCS 7906190246
Download: ML19261D440 (25)

Text

-

QUAD-CITIES NUCLEAR POWER STATION UNITS I AND 2 MONTHLY PERFORMANCE REPORT MAY 1979 COMMONWEALTH EDlSON COMPANY AND 10WA-ILLIN0lS GAS r, ELECTRIC COMPANY NRC DOCKET N05. 50-254 and 50-265 LICENSE NOS. DPR-29 and DPR-30 2307 160 7 9 0 61 goy G

TABLE OF CONTENTS

1. Introduction ll. Summary of Operating Experience A. Unit One B. Uni t Two Ill. Plant or 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 Modifications

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. Refueling Information Vill. Glossary 2307 161

1. IflTRODUCTION Quad-Ci ties fluclear Power Station is composed of two Bolling Water Reactors, each with a Maximum Dependable Capacity of 769 MWe net, located in Cordova, Illinois. The Station is jointly owned by Common-wealth Edison Company and Iowa-l11inois Gas & Electric Company. The fluclear 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 plant is subject to license numbers DPR-29 and DPR-30, issued October 1, 1971 and March 21, 1972 respectively, pursuant to Docket flumbers 50-254 and 50-265 The date of initial reactor criticalities for Units I 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 Uni t 2.

This report was compiled by David Hannum. Telephone number 309-654-2241, extension 179 2307 162

11.

SUMMARY

OF OPERATING EXPERIENCE A. Unit One May 1: Unit One began the reporting period operating at an electri-cal load of 775 MWe.

May 2-4: Unit One held an average electrical load of 750 MWe. At 2000 load was reduced to 400 MWe for main condenser flow reversal.

May 5-9: Load was steadily increased to approximately 775 MWe and held at an average load of 750 MWe.

May 10-14: At 1300 Unit One load was reduced in preparation for a weekend maintenance outage. On May 11 at 0301 the generator was taken off-line and at 0822 all control rods were fully inserted.

The weekend outage work was accomplished in a satisfactory manner.

On May 13 at 1402 the reactor was made critical and the unit was placed on-line on May 14 at 0825 Load was subsequently increased at the rate of 100 MWe/hr.

May 15: Load was reduced to 35 MWe due to a decrease in condenser vacuum. After making some valve adjustments and investigation corrective actions, vacuum was returned to normal. At 1500 load was increased at a rate of 100 MWe/hr.

May 16: Load was increased to 708 MWe and was then reduced to 500 MWe for control rod pattern change.

May 17-19: The load was steadily increased to 775 MWe and then held steady.

May 20: Load was reduced to 175 MWe due to a reduction of condenser vacuum. Load was increased after corrective actions were taken.

May 21-23: Load fluctuated between 800 MWe and 716 MWe May 24-27: Load was reduced at the rate of 100 MWe/hr in preparation for a maintenance outage to repair an apparent condenser tube leak.

On May 24 at 2313 the unit was taken off-line. The condenser hotwell was flooded to determine the location of the postulated tube leak. Four condenser tubes were plugged during this outage.

On May 26 at 1555 the reactor was made critical and the unit remained in HOT STANDBY due to low system demand. On May 27 at 1144 during the turbine roll, the field breaker would not close-in and a manual turbine trip was initiated. At 1811 on May 27 the generator was put on line.

May 28-31: Load was steadily increased to 791 MWe.

2L307 1e3

B. Unit Two May 1: Unit Two began the ' reporting period operating at an electri-cal load of 439 MWe.

May 2-5: Load was steadily increased on a preconditioning ramp of 3 MWe/hr. Load was then held steady at 716 MWe.

May 6-11: Unit Two held an average load of 650 MWe.

May 12-13: Load was reduced to 400 MWe for weekly turbine testing.

May 14-18: Unit Two held an average load of 675 MWe.

May 19-20: Load was reduced to 591 MWe for main condenser flow aversal.

May 21-23: Unit Two held an average electrical load of 666 MWe.

May 24: At 1259, a reactor scram occurred due to a false low reactor water level signal induced by accidental vibration of an Instrument rack. At 1725 the reactor was made critical and at 2055 the main generator was put on-line.

May 25-28: Load was steadily increased on a preconditioning ramp of 3 MWe/hr to 725 MWe.

May 29-30: Unit Two held an average load of 708 MWe.

May 31: Load was reduced to 450 MWe in order to satisfy minimum system load requirements.

Ill. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS AND SAFETY RELATED MAINTENANCE A. Amendments to Facility License or Technical Specification.

There were no amendments to the Facility License to Technical Specifications during the reporting period.

B. Facility or Procedure Changes Requiring NRC Approval.

There were no facility or procedure changes requiring NRC approval during the reporting period.

C. Tests and Experiments Requiring NRC Approval.

There were no tests or experiments performed during the reporting period requiring NRC approval.

2307 164

D. Other Changes, Tests, and Experiments

1. Facility Modifications There were no modifications completed during the reporting period requiring NRC notification.

2. Special Tests

'Jnit Two Drywell Cooling Air Damper Supply and Control isolation The purpose of this test was to check the feasibility of fully opening the Unit Two drywell cooler dampers, and disabling the control and supply air lines to the dampers.

Safety Evaluation:

The function of the drywell cooling system as described in the FSAR did not change. This test drew more cooling air in the lower regions of the drywell where more safety equipment exits.

Also, this change decreased the demand for nitrogen during unit operation. There was no loss of nitrogen due to control damper tubing leaks. This test opened the dampers and permitted the system to function properly.

Unit Two Pressure Setpoint Change Test The purpose of this test was to demonstrate the ability to withdraw intermediate control rods on a preconditioning ramp farther than would otherwise be possible without a pressure setpoint change.

Safety Evaluation:

The probability of an occurrence or malfunction previously evaluated was not increased because at the lower reactor pressure there is even more margin to accident analysis peak pressures for analyzed transients. Also, all operation was within normal operational limits - and fell within Technical Specification limitations.

E. 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.

07 ies

UNIT One MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS ON ACTION TAKEN TO LER OF W.R. SAFE OPERATION PREVENT REPETITION NUMBER COMPONENT MALFUNCTION NUMBER Reactor Head Vent The operator dia- The diaphram was leaking. The diaphram was replaced.

2625-79 Valve (1-220-47) phram was defectivelalve operability not affected.

2611-79 Main Steam Line The limit and The valve was giving dual The switches were recalibrate Drain Viv torque switches i ndication. Valve closure and the valve was tested 3 (M0-1-220-2) needed adjusting. and isolation function not times.

affected.

RC1C drain valve The solenoid coil rhe valve would not open. The air solenoid coil was 2592-79 (1-1301-13) was defective. ICIC was operable at all replaced.

imes.

RHR H. Ex. Bypass The breaker was 'he breaker would trip whenThe breaker was replaced. Th 2407-79 valve was tested.

Viv. (1 -1001 - 16B) defective. rying to close valve.

falve is normally open.

The limit switch was replaced 1235-79 MSIV (1-203-1C) The limit switch Dual indication was not was defective. eceived during MSIV bi-ucekly surveillance. All LPS relays actuated satis-actorily.

2573-79 Refuel Bridge A cable controlling"he grcppie had no vertica The cable was replaced with a the vertical travel ravel. No refuel activi- spare. The grapple was teste (1-833) of the grapple was 1les were being performed.

broken.

3x. Blda. Vent A solenoid coil was~he vent system isolated The solenoid valve was 2528-79 t,ithout a control room replaced. Proper operatier Systemi1-5700) shorted.

()$ clarm. Isolation capa- the vent system was verified.

CD 'ility or performance not sa affected. -

RHR H. Ex. The expansion joint The expansion joint was The gasket was replaced.

~7' 1006-79 R0 79-17/03L C7\ (1A-1003) casket was defectivolenkinn. RHR Hv was nearn

UNIT One MAINTENANCE

SUMMARY

CAUSE RESULTS 6 EFFECTS OF CN ACTICN TAKEN TO W.R. LER PREVENT REPETITION COMPONENT MALFUNCTION SAFE OPERATION NUMBER NUMBER The valve was The rod would not with- The valve was replaced and th 2109-79 CRD withdraw viv (1-305-122) defective. iraw. Scram capability rod tested.

Tot affected.

2468-79 Pressure Suppres- The versa solenoid The valve was leaking air. The versa valve was rebuilt.

sion Valve valve was defective /alve Isolation capability The valve was cycled 3 tines.

(1-1601-63) tas not affected.

)perator was leaking a i r.- The air operator was cleaned.

2464-79 SBGT (AO-1-7510A) The air operator was defective, iBGT was still operable. The valve was tested.

SBGT (AO '.-7510B) The air operator Operator was leaking air. The air operator was cleaned 2470-79 SBGTS was still operable. The valve was tested.

was defective.

RHR Vlv The motor was The breaker tripped when The motor was rebuilt and th 2402-79 trying to open the valve. torque switch was replaced.

(1 -1001 -186A) defective.

RHRS was fully operable. The valve was test operated.

Connectors werc ~he LPRM's were reading up The connectors were replac:d 1586-1589-79 tpgg s (1-756) defective, scale. APRM function not compromised.

he valve would not open. The coil was replaced. The 2776-79 RX Bldg. Vent The dc coil was defective. Isolation function not valve was test operated.

isol. Valve (A0 1-5742B) affected.

RHR Service Watc' The torque switch he thermals tripped durin :The torque switch was adjust 2673-79 valve operation. RHRS was The valve was cycled 3 times rNJ Valve needed adjusting.

l>4 (M0-1-1001-186A) fully operable.

CD N

N o

UNIT Two MAINTENANCE

SUMMARY

1 -

CAUSE RESULTS & EFFECTS OF ON ACTION TAKEN TO W.R. LER PREVENT REPETITION NUttBER COMPONENT MALFUNCTION SAFE OPERATION NUMBER Pressure Switch A wire was found The valve was leaking. The wire was removed and the 464-79 Switch was operable. seat was cleaned.

Isol. Valve around the stem.

(PS-2-1622A) The seat was dirty.

Shutdown Cooling The aux. contacts Valve WOULD NOT close The contacts were cleaned.

724-79 79-04/03L Viv. (2-1001-50) were dirty. from control room. Mo The valve was cycled 3 'imes..

1001-47 was operable &

closed.

A Hi input was giving a The connector was tightened.

LPRM 6C-24-09 The detector con-2297-79 Hi-Hi alarm. Rx was in .

(2-756) nector was loose.

shutdown mode.

CRD 06-31 valves The valves were The scram operation func- Replaced insert directional 2270-79 tion of the drive was control valves.

(2-302-121 c 123) sticking closed.

unaffected.

inboard MSIV The limit switch The switch was not giving Replaced limit switch.

2321-79 an open indication. Scram (2-203-1A) was defective.

function was unaffected.

The limit switch The switch was not giving Replaced limit switch.

2320-79 Inboard MSIV (2-203-1B) was defective, an open signal. Scram function was unaffected.

Outboard MSIV Limit switch was Valve showed dual indica- Adjusted limit switch and 2295-79 tion. Scram function was test 'perated, (2-203-2A) out of adjustment. .

unaffected.

Inboard MSIV Limit switch was Limit switch was slow in installed new limit switch an 2272-79 out of adjustment. actuating relay. Scram tested scram circuit.

(2-203-1B) function was operable.

Excessive Seal The drive had excessive Replaced the drive mechanism 2203-79 CRD (42-39)

S/N 1339 Leakage. stall & run flows. The and fully tested.

rs) scram function was still l'd operable.

CD N -

6 T

[

IV. LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for Quad-Cities Units 1 and 2 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 0NE Licensee Event Date of Report Number Occurrence Title of Deviation 79-15/03L 5-24-79 Drywell to Torus Vacuum Breaker 1-1601-32D Dual Indication 79-20/03L 5-25-79 MSIV l-203-2C Excess Closure Time UN IT TW0 5-14-79 Fuel Pool Monitor Fa led Upscale t

79-10/03L 79-ll/03L 5-21-79 ESS Cable Routing Error V. DATj 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 2307 i69

February 1977 I

OPERATitlG DATA f;CPORT DUCKET !!O. 050-254 D

' UtilT One DATE 6-5-79 COMPLETED BY D. Hannum

~

TELEPH0!!E (303) 654-2241 ext. 179

~ OPERATitiG STATUS 0001 050179

3. Reporting period:2400 053079 Gross hours in reporting period: 744

2. Currently autho'-i :ed power level (MWt) : 2511 Max. depend. capacity (MWe-Ne t) : 769n Design electrical rating (MWe-flet) : 789 . .

3 Power level to which restricted (if any) (MUe-tiet): tJA

4. Reasons for restriction (if any):

This Month Yr. to Date Cumulative 654.8 2558.8 49891 7 5 tiumber of hours reactor was critical 0.0 0.0 3329.6

6. Reactor reserve shutdown hours 599.6 2429 9 47372.8 7 Hours generator on line _

19.8 19.8' 909.2

8. Unit reserve shutdown hours.

4882138 93690014 9., Gross thermal energy generated (MWH) 1273164 Gross electrical engergy generated (MWH) 401000 1534558 3')107053 10.

379998 1433093 28065666

. 11. Net electrical Energy Generated

.. 12. Reactor service factor 88.0 70.6 80.7 13 Reactor availabilify factor 88.0 70.6 86.0

14. Unit service factor 80.6 67.1 76.6 g

15. Unit availability factor 83 3 67.6 78.1 ._

16. Uni t capactly f actor (Using MDC) 66.4 51.4 59.0 17 Unit capaci ty factor (using Des. MWe) 64.7 50.1 57.5

18. Unit forced outage rate 17.2 5.8 8.1 19 Shutdowns scheduled over next 6 months (Tr a date, and duration of each):

20. I f shutdown at end of report pariod, estima ced date of startup':.. .

]V A

.w ,

• The f tDC nay be lo. er than 769 MUe during periods of high ambiant temperature due -

to the thermal performance of the spray canal. -

"/ ~ ~. i 2 3 07 l (f.inal) 17 0Q.7 ".c. v.- r.

u . . c

, .. u , , . ,

.. !; P '4 C !:,i! : *. UNil-P0...R

. . EL June 1976 Docket No. ,50-254 0

Unit One Date 6-1-79 hh .

s

\,S u

b, Completed by D. Hannum Telephone (309)654-2241 Ext. 179

.HONTH May, 1979

. DAY AVERAGE DAILY POWER LEVEL- DAY AVERAGE DAILY POWER LEVEL

(!tWe-Net) (HWe-Net)

1. 731 17 533

2. 696 18, 667 3 713 19 732

4. 683 po, 591 691

}5 '

708 21 716

6. 551 22.

7. 710 23 743

8. 24. 572 677

9. 679 25 -11 -

10. 612 26. -11

,11. 8 27 22

12. -9 -

28. 512  ;

13 -13 29 695

14. 189 30. 726 15 351 31. 741 AeeHuVED

16. 625

. - JUN 2 01976 INSTRUCTIONS

) .

On this farm, list the average daily unit power Icvel in !.iWc Net for cath day in the reporting mo9thkinhOttfhe nearest who!c inegawatt.

These figure.s will t<e used to plot a paph for each reporting month. Note that when nuximum dependabic capacityis used for the s:et c!cetric.d rating of the unit. there may be occasions when the daily average power I vel exceeds the 100'4 hoe (or the aestnered power lesel hne). In such cases. the averar.c daily unit power culpt.t sheet should be footnoted to exp! in the apparent anono y.

1 (final) 2307 171

~

~)-.~.. . .._ ..-. . .

OPERATI:lG DATA liCPOR:

050-265 _-

DOCI'ET NO.

Two .

UNIT 6-5-79 DATE _

L c .w .

D. Hannum -_

COMPLETED BY _

(309) 654-2241 ext. 179

~

TELEPH0!!E _

~ OPERATING STATUS 0001 050179 Gross hours in reporting period:__744

3. Reporting per iod:2400 053079 2511 _ Max. depend. ccpaci ty (MWt) : 789  : _.

2. Currently authorized power level 769A Design electrical rating ',

(MWe-Ne '

(MWe-Me t) :

): NA Power level to which restricted (if any) (MWe-Met .

3

~~

4. Reasons for restriction (if any): Cumulative This Month Yr. to Date 3529.4 48905.8

_739.6 5 Number of hours reactor was critical 0.0 2985.8

~. 0.0 _

6. Reactor reserve shutdown hours 46595 0 736.1 _ _

3492.1 7

Hours generator on line 0.0 702.9 0.0 '

Uni t reserve shutdown hours. 96111134

8. 1528104 7369010 Gross thermal energy generated (MWH) 2306890 30795267 9., 468898 10.

Gross electrical engergy generated (MWH)_ 2146975 28908435 __

428220 _

80 3

11. Net electrical Energy Generated 97 4 _

.- 99.4 Reactor. service factor 97.4 85.2 _

. 12. 99.4 _ _

13 Reactor availabilify factor 96.4 76.5

,. 98.9 '

14. Unit service factor 96.4 77.6

i. 98.9 _

15 Unit availability factor _

77 1 61.7 74.8 _

60.1

16. Unit capactly f actor (Using MDC) _

75 1 __

72.9 __

17 Unit capacity factor (using Des. nWe) _

0.5 '99 _

1.1 E 18. Unit forced outage rate f each):

scheduled over next 6 conths (Type, date, and duration o 19 Shutdowns . NA 20.

I f shutdown at end of report pariod, estimated date of startupj rature due

• The tt0C may be lo.ter than 769 MUe during periods of;t.high g,i a nbian f f. /

to the thermal perforuance of ti.e spray canal. . Q . w~. ra. u. , .

1 (final) 2307 172

n'. nt,c r ".:!: - UN il-P:^. % ~I.EVEL June 1976 Docket No. 050-265

) 0 Unit Two

. Date 6-1-79 s Completed by D. Hannum Telephone (309)654-2241 Ext. 1,79 Il0 NTH May, 1979 DAY AVERAGE DAILY POWER LEVEL

. DAY AVERAGE CALLY POWER LEVEL-(!iWe-Het) (HWe-fle t) 17 622

1. 412

18. 610 2

2. 488 19 605 3 560

20. 577

4. 627 21 615 663

}56. ~

592 22. 605 613 23 597 7

8. 6?7 24. 326 616 25 428-3, 582 26. 515 10.

27 586

11. 582

28. 644 .

12. 585 29 656 13 469

30. 618

14. 634

31. 538 614 15 aeeauVED

16. 623 JUN 2 01976 INSTRUCTIONS On this form, list the average dai!y unit power icvel in !.lWe Net for each day in the reporting mo

} . neas est whole inegawatt.

The.es finore.s will t.e used to plot a ruph for each reporting month. Note thit when nuximum dependable capac used f or the net electrical rating of the unit. there may be occasions when the daily average power level exceeds the 100% line (or the nestnered power lesel line). In sucii cases, the averar,c d.iity unit power culput sheet should be foutnoted to explain the apparent anomaly, i (anai) 2307 173 . . .

n-. . - . -

.. ....-, - .w. . . . .

( t U .,

APPENDIX D .; b - QTP 300-S13 1 -

Revision 5

' UNIT SHUTCOWNS AND POWER REDUCTIONS March 1973 050-254 __ ,,

DOCKET NO-D. Hannum u es One .

COMPLETED BY UNIT NAME -

May, 1979 TELEPHONE (309) 654-2241 6-5-79 REPORT MONTfl DATE m s -

= o !! t; z 5 m

LICENSEE pg gy .

y e.:

po DURATION $* 8{$

%,* EVENT *8 m

g: 8 o CORRECTIVE ACTIONS / COMMENTS NO. DATE

' (HOURS) y,y7 REPORT NO.

R 4 NA NA NA Load was reduced for condenser flow reversal.

F H 9 790505 NA NA Unit One was shutdown for a maintenance outage.

8 NA 10 790511 s 77.4 ,

1 4 NA NA NA Load was reduced due to a reduction of condenser 11 790515 F. H vacuum.

4 NA NA NA Load was reduced due to a reduction of condenser F H 12 790520 vacuum. ,

NA' NA Unit.One was shutdown to plug condenser tubes.

13 790524 F 67.0 B '1 NA N

u co N

N 4

(finof)

APPENDIX D QTP 300-S13 Revicien 5 UNIT SHUTDOWNS AND POWER REDUCTIONS 050-265 March 19'/8 g,.7,lT NO.

D. Hannum Quad Cities Two ,

COMPLETED BY UNIT NAM.r - - -

79 May, 1979 . TELEPHONE (309) 654-2241 REPORT HONTH ,

DATE w 5 b

.. m = Eb =

gs 0 - .

w e g g-$ LICENSEE 58 . .

>- o DURATION < :c $ e EVENT *8 R: 8

' y REPORT NO.

$ g -

CORRECTIVE ACTIONS / COMMENTS NO. DATE (HOURS) ~

R NA NA NA .oa d was reduced for weekly turbine testing.

S H NA 8 790513 NA NA NA Jnit'lwo reactor scram occurred due to a false low 9 790524 F 79 G 3.

reactor water level signal induced by accidental

/ibration of an instrument rack.

o N .

V '

CD .

M t

t (final) ..

. . . ... .. . * ** 4

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 operations during the reporting period.

B. Control Rod Drive Scram Timing Data There were no control rod drive timing tests performed during the reporting period.

Vll. 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.

2307 i76

'qTP ';.;0-532 . .

Revision 1 QUAD-CITIES REFUELit!s March 1978

] $ .

INFORl!ATION REQUEST c

1. Un i.t : 1 Reload: 5 cycle: 6

2. Scheduled date for next refueling shutde'.-;n: September 1, 1980(Shutdr

EOC ~

.3 Scheduled date for restart following refueling: Dec 15, 1980(Startup Bac_

4. L'ill refueling or resumption of operation thereaf ter require a technical specification change or other license amendment:

~ '

5. 5heduled Infornation: date(s)

The QCl for [ubmihting R5 licensing submittal proposed is scheduled licensing for Sept action1980 and suf,por 6...._Important licensing _ considerations associated with refueling, e.g., new or

' dif ferent fuel design or supplier, unrevie. cd cesign or performance analysis

ic t ho::s , significant changes in fuel desig.., r.cw operating proced.:res:

fl0NE

~

7 The number of fuct assemblies.

a. Ilumber ~of assemblies in core: 724

___b....!! umber of assemblies .in spent fuel pool: . 343 8.

The present licensed spent fuel pool . storage capacity and the s.ize 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: -

b. Planned increase in licensed s torage: -

None 9- ~he projected date of tiie last refueling that can be discharged to the spent

.. fueI pool assuming the present 1icensed capacity:

WPPROVED

. 2307 177 . APR 2.01973 O. C. O. S. R.

' QTP 300-533' ,

RELOAD LICENSING PACKAGE

• RcVisica i PREPARATION. SCHEDULE March 1973 .-

UNIT Q1 RELOAD li CYCLE 5 ,

' ~

.- ACTIVITY .-

DATE RESPONSIBILITY CENTER .

NFS receives draf t Licensing Submittal f rom GE 5/78 GE NFS Transmit copy of draf t to Station f.or Comments .

I

! NFS Transmit NFS and Site comments / questions to GE '

29/78 NFS D6 o i n Tach _ Sac.c _cbann/es.. Saf.c't .y. Eia.Inat.tca. and. co.va.r_.Lc.t.tcr (E UFS receives final Licensing Submittal and answers to CECO questions f rom GE

/30/78 NFS - Con 2 Died e fi n al M FS re.yj ew o f 1 I c.n ns i n g_ Wi t tal _ f n el-w..m.re _ en _ rf rh queuj.c

/1/73 NFS Transmit complete occkage for on/off site review ' = . _ _

i/3/73 Station On-site review comnleted 1/6 //3 PSA Off-e,tre_ review ccm.intnd n AN

\

Compi e t cd 1 I cene,1 net parl10n enenhe H y 4Rf~

1/11/78 NLA ^ .

. :. ' " e) I U

90 day

, Anticipated. unit shutdown W 12/79 -

. 28 days . '.6. -

' . 5?.e) y=b

, w.  :

.$ y

' Receipt of operating License

/9/79 .

b p

.- Jf

. Me ys

..a . . .

day outage y,

Anticipated. Unit Startup - Assumes -

56

/9/79 , . . .

,,_8 weeks

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CD. -

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. NFS/DWR

, .- "., Prepared by- -

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Reduire Changes i -

Item page _

ceneralize trording and -

Scccm ReactivI'ty 4 reference thh submi t._ ~ .

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ene. Adequate .pressur..e..

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RI.

Satpoin ts "

'Changd to '{.65w+XX) Es' r.c ., id ..;

LCO 3 2/4.2-14 change operabi1 i ty t6 xx.t;

)

3.2/4.2-7 Change Referet:ce; I to N 03 . ?C Sases -

3 2/4.2-8'

- g ^v n've'- -

. 5 -. .

.L c

. i.

- *- t. .-

. . o

-l -; 2  ; J [.

j A.uto Flow Control 3.3 /4. 3-5 Cone . sjability' anal'isits not l LCO , , -

- l ic.i t i ng .'

I. E-' '(

f .

t. .=- t u t t ' . -

t o -[,

' ~s - .}

'l 5 Sases .

3.3/4.3-11  ! !!one. .

l- ._

a 2.. .;

i u

g i

j - -

- s m .-'t,.

- - . i.

F .

g < _p .

[

l{.*,P LH CR

~

Fig. 3,5,1

• Revise curves to refleet-LCO

- '(shts, I to 3) -

oc: cncly --

3 .

s

i. .

i ,

• Change reverences .to eqrlect Deses , ,

, .3.5/4 5-14 nca analyses o f MED3-2 30 45^. -

. 1 -

_ _t .. .,

, y 4

-g

p. 3 7/ji . 3 , r,

r. c , s . ,

ne. ,,a 3 n . . >._

....}.....

am (.f x f y .,

.v. t. ,, ,. ., o r. g

% . r. v v. . (u y. o, t.r- J .

General.iuc description- csf P.ases 3 5/4.5-14 l i;ai t i ng . t rans ient (s) .

I r

- ~~- - _

2307 179

. "r.

GI. c.ir m r ..

rc !. n i t:.; Isa.1d i c ' utid - r s.eiorate co J ' P 'r. e :'d ; i a. .c i C . X Y. CP?. P taity fm nmi Lcediaj Error Ax.ident

QTP. 300-532

• , c. . .. .- P.evision 1 -

j: -

llarch 1978 ,

s QUAD-CITIES REFUELittG -

INFORMATIOi! REQUEST _

~

~~;- Rsl oa d :'- -4~

- : _ Cycle: g; (next outage) . . . . . ...

. OniS-E " -'t _

septen5sr 30,1979 (shutde. .

EOC4)

' Schei[ure~d date for Wext'iefeiling shutdownr .

2.

~.

January' 20.1980 (Startu;i

.... ~ '

BoCS)

Scn'eddled.. ..dat.e fof.r5stait} ;fhll'owidg-

.- . -2  ; , refueling: .-.

3 . -. .

hi l Uille refueling or resumption of operation t: thereaf Simlar Tech. te'r 'require Spec. changes a tec-n ca 4.

. specification change or other license aman aan d '

~

,._,, , , _ ,s,. _ < < . .

~ to Pil oa d ;3.. Cy cl e 4. - _ , . . ... - ,, . p . - 7,. ,

d su; porting ~

. e. -

t.  : ..; :.--

i Scheduled date(s) for submitting proposed licensing act on 5 " ,.

in forma tion: ._. . .. ....

? . .

to shutdown. . :t - .; - . . . . . . . . .

' r m:.t . .. ;.:i ; :.- _ _

6. !apahtant licensing censidsrations associatei' ui thf ref uelingl, sis e.g.,-ne

' dif ferent . fuel design. or supplier, unreviewed design .

t ing or procedures:per ormance ana y r:ethods,, signi ficant changes in fuel design, ne.i opera ; ~

Retrofit 8 x 8 fuel (app rofiWtdly'196) .

- -- a . . _ . . .

Hev: fuel designs: .

O fGQ ..

• . ~ .

w f&n,issf.

. . ~ - _ . _ .

~

g 5 g. _ , . , ,

3 Di: .Ts u.%L*-:: ci

- 1- u . c. .

'w  ;:

• C: E i. .:  ! ! .- ;__ . . ~ . . .

. 7.- .

7. The number. of. fuel L..: assemblies.

• _. . ~ ~ ~ '

.224

a.r. !!umbeg .of as.semblips _in core: ,,. .

- - ~745 ~-

b... ..Mumber of assemblies ih spent fuel pool:-

8. The p. resent increase in licensed spent fuel pool storage capacity and ' '

ill number...of fu21 assemblics: ~ .

60 C- -Liceased stcrage capaci ty for. spcnt fuel: -

_ !!on e .

5 ;. :.i . . ..s  : .. . ,, ' .

b. Planned increase in" licensed storage:

refueling that can be discharged to the Last refueling date '.-ri S. The' pro.iected spant fue1 pool assuoing the date of the Iastpresent 1icer. sed capaci ' t'/:

Present capacity: September. 81 '

23.07'l80 RELOAD LICENSING. PACKAGE OTP 300-533 U"!T OC 2 ' .

PREPARATION SCliEDULE - '

PcVisio1 1 '

RELCAD 3 CYCLE 4 , P.' arch 1973

'c DATE RESPONSIBILITY CENTER '

ACTIVITY 10/6/77 GE NFS receives draft Licensing Sdbmittal from CE Transmit copy of draft to Station N1' S for corrnento. .

NFS 10/2C/77 dFS Transmit NFS and Site comments / questions to.GE Begin Tech. Spec. changes, Safety Evaluation and Cover Letter.

• G P.

11/3/77 NFS UFS receives final Licensing Submittal and answers to CECO questions from GE.

Comp 1nte finn 1 Urs vnvien of 'Aenm in d uhriet,1 1, ,, ,1 n n nt 7e rs en CICo nuerinn, 11/3/77 NFS Transmit complete pachace for on/off site review '

11/16/77 Station on-nite reviou compiercJ 11/18/77 PSA Off/ nite revicu connleted' 12/1/77 NLA Completed Licennine n.,cknee rnenivert' hy Nnc

, 90 dayc 1/16/78 -

Anticipated unit shutdotm

• N .

g 28 days CD , '

N l .

3/5/78 _ :-

Receipt :of .operatin'g License oo ,

l , Day outage i 3/15/78 -

Anticipated Un1t Startup - Assumes -

1. 8 Wecks

. I

~? .

... 1 i i} Prepared by JAS . NFS/BWR

, , 06tc .  ?/3'l!70

e"

1. J

-..~- nce

' I ' ' " ~ '" ~ : c "~ " T 5 PREL!;illipKY CI!ECKLISI IOP EELC.. -

~

hh Ol Q[Il 4r

'h REL 3 ,

4 j l

0 K'g CY C L.E:

]

' - Require Changes 1 ten .

s Pecc  : .

~

l Generaliz wording and 4

X Scraa Reactivity - reference the subrait, ' ~

. I:E00-24063. ~ ~~ ~ -~ ~-~ ~ --~ -

-- '! Safety Valva Setpoints ' i:ane. Adequate pressure . . .

1. 2/2, . 2,-1 marg:n.

.a. !L555 .

C'larify and add boundi_ng

'l - 1.2/2.2-2,3 -

X i - '

peak pressure.  :

. l, F,as es-g

i. -

i Change to (.65w+42) ? .'-

i RS,9 Setpoints - .

p-3.'2/4.2-14 I Lco l Char.ge operabil i ty to.3Od.

Y 3 2/4.2-7 .

l " ,

, 3 2/4.2-3 'l Ch2nse Reference 1 to- ' -

,: !I hs'es - f .

lNEDG-24053 I

l

.f. i  ! .

i Auto Flow Control

' ticae. stability caalysis

A I LCO

.3 3/4.3-5 not liraiting. . ..

' *b . .

~

None- .

3 3/4.3-11 1

.  ! sases . - ,

J. y 1

1 - . ,g -

I 3 s

y '

Revise curves to reflect _.

MAPLl!GR  :

Fig. 3.5.1 new analyses.

Y. LCG^

(shts. 'I to 3) 1 3 - '

• Change referencis~ to ref tuct-X Sases

' 3.5'/4.5-14 ne.. analyses of' liECO-24G46.

.: :- ' , u .- . .

s..

... _i

._g .

!!s.: values: Idi .33 (7 E~7)

- ' 3.5/4.5-10 l l;CPP.

1.35-(8 x-8} -

Y LCO '

, ceneralize descriptio.. of Bases 3 5/4.5-lh ~

~ '

limiting transient (s) .

2307 182 t

.1 r r b i n.n, :,ar.dled unde r c.rpara te cova r.

F."' i. H C R c i t :i n ~., .. c.. -

r P ' .;30/1l E y [ U Eue} LCOd.InC b.fIOI 3#C30UnE

• * * - - "Y

Vill GLOSSARY

/

q I The following abbreviation which may have been used in the Monthly Report, are defined below:

CRD - Control Rod Drive System SBLC -

Standby Liquid Control System MSIV -

Main -Steam isolation Valve RHRS - Residual Heat Removal Systen' ,

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 -

Average Power Range Monitor _

Traveling incore Probe i TIP -

RBCCU - Reactor B'uilding Closed Cooling Water System TBCCW - Turbine Building Closed Cooling Water System RWM -

Rod Worth Minimizer SBGTS -

Standby-Gas Treatmen t System HEPA -

High-Efficientry Particulate Filter RPS - Reactor Protection System IPCLRT - Integra ted Primary Containment Leak Rate Test LPCI - Low Pressure Coolant injection Mode of RHRS RBM -

Rod Block Monitor BWR - Boiling Water Reactor 151 - In-Service inspection MPC -

Maximum Permissable Concentration 2307 183

PCI - Primary Containment isolation SDC - Shutdown Cooling Mode of RHRS LLRT - Local Leak Rate Testing MAPLHGR - Maximum Average Pianar Linear Heat Generation Rate R.O. - Reportable Occurrence DW -

Drywell RX Reac to r EHC

- Electro-Hydraulic Control System MCPR -

Minimum Critical Power Ratio PCIOMR - Preconditioning Interim Operating Management Recommendations LER - Licensee Event Report ANSI -

American flational Standards institute NIOSH - fla t'iona l Institute for Occupational Safety and Health ACAD/ CAM - Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Moni toring 2307 184