ML19274D384

Monthly Operating Rept for Dec 1978

ML19274D384
Person / Time
Site:	Quad Cities
Issue date:	12/31/1978
From:	Hannum D COMMONWEALTH EDISON CO.
To:
References
NUDOCS 7901230282
Download: ML19274D384 (25)

Text

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r QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFOICIANCE REPORT Deccaber 1979 COFDIONWEALTil EDISON COMPANY AND IOWA-ILLINOIS GAS & ELECTRIC COMPANY NRC DOCKET NOS. 50-254 and 50-265

~ ~~~~~~~ ~~ ~ ~ ~ ~ ~ ~~

- ---~

LICENSE NOS . DPR-29 and DPR-30 ' ~~~~ -~ ~ ~

O 790123 oMfM

TABLE OF CONTENTS .

I. Introduction II. Summary of Operating Experience A. Unit One B. Unit Two III. 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 Mcdifications

2. Special Tests E. Corrective Maintenance of Safety-Related Equipment IV. License Event Reports V. Data Tabulations VI. Unique Reporting Requirements A. Mhin Steam Relief Valve Operations B. Control Rod Drive Scram Timing Data VII. Refueling Information VIII. Glossary O

I. INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Boiling Water Reactors, each with a IIaximum Dependabic Capacity of 76911We net, located in Cordova, Illinois. The Station is jointly owned by Commonwealth Edison Company and-Iowa-Illinois 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 111ssissippi 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. Commercial gener-ation of power began on February 18, 1973 for Unit 1 and 11 arch 10,1973 for Unit 2.

This report was compiled by David llannum. Telephone number 309-654-2241 ext. 252.

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II. SUh%tRY OF OPERATING EXPERIENCE UNIT ONE December 1: Unit One began the reporting period operating at an electrical load of 543 MWe.

December 2-12: Unit One held an average electrical load of 525 MWe.

December 13: At 2043 load was reduced to 430 MWe due to 1A reactor "

feed pump trip.

December 14: Electrical load was increased to 500 MWe.

December 15: Load was reduced to 400 MWe for condenser flow reversal.

December 16-27: Unit One held an average electrical load of 483 MWe.

December 28-31: Load was reduced to 385 HWe to remove from service the IB reactor feed pump for repairs.

B. UNIT TWO December 1: Unit Two began the reporting period operating at an electrical load of 777 MWe.

December 2-9: Unit Two held an average electrical load of 766 MWe.

December 10: Load was reduced to 400 MWe for HsIV surveillance tests.

December 11-29: Unit Two held an average electrical load of . -. - ---

733 MWe. .

December 30-31: Load was reduced to 200 MWe for the control rod sequence change.

G e

- III. PLANT or PROCEDURE CHANGES, TESTS, EXPERDENTS, and SAFETY RELATED MAINTENANCE A. Amendments to Facility license or Technical Specification Am'endments to Technical Specification during the reporting period include. revision Amendment No. 48 to DPR-29 and revision Amendment No. 47 to DPR-30. These amendments revise the Technical Specifications to provide operating temperature and pressure limits in accordance with Appendix G,10 CFR Part 50.

B. Facility or Procedure Changes requiring NRC approval M-4-2-76-4 Modification Summary:

This modification involved changing the power sources for Unit 2 Main Steam Line Radiation Monitors Band D. The previous source of power for the monitors had been Reactor Protection Systen Bus B. The new power source for the monitors is the 120V Essential Service Bus. The intent of this change is to prevent a Group I isolation upon any loss of power causing the loss of both Reactor Protection System buses.

Safety Evaluation:

The possibility for an accident or malfunction of a different typc than any previously evaluated in the FSAR has not been created because no change in the integrity of the logic circuits or scram functions has been made. A higher standard of reactor protection and safeguard has been attained as a result of this modification.

C. Test and Experiments requiring NRC approval. There were no tests or experiments performed during the reporting period requiring NRC approval.

D. Other Changes, Tests, and Experiments On December 22, 1978 Special Test 2-21, to install a Dynamation carbon monoxide monitor in place of one of the three existing MSA carbon monoxide monitors, was completed. The test consisted of replacing one C0 monitor of the service air system for six months. Operation of the monitor proved satisfactory

~

and the other monitors will also be replaced.

Special Test 1-40, to increase the turbine pressure setpoint on Unit One,was completed on December 20, 1978. The purpose of the test was to provide information on maintaining higher reactor power during unit coastdown by raising the reactor pressure. A small increase in power was realized when the reactor pressure nas increased by approximately 10 psig.

All pressures were returned to normal following completion of the test.

O

E. Corrective Maintenance of Safety-Related Eq.uipment.

The following represents a tabular sur.: mary of the safety-related maintenance performed on units 1 and 2 during the reporting period. The headings indicated in this summary include Nature of Maintenance and Work Requests number, LER numbers, components, cause of malfunctions, results and effects on safe operation, and action taken to prevent reptition.

... = . , * * - .- .. m

UNIT I MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS ~

ON ACTICN TAKEN TO LER OF W.R. SAFE OPERATION PREVENT REPETITION NUMBER COMPONENT MALFUNCTION NUMBER The Limit The Interlock System The Limit Switches Corrective Refuel Limit Switch llolding were Replaced Bridge - Switches Needed were Adjusted 4769-78 Plate was moved (833/834), Recalibration i requiring adjustment of the Switches.

'The Pressure The Breaker was A Different Set of Corrective 15 Diesel The D. G. Contacts on the Pressure Generator Switch contacts Tripping Out.

5543-78 was still Operable Switch were connected.

(h-6601) were Defective The compressor _was successfully test operate Corrective RO 78-33/03L-0 Rx. Bldg Vent The Solenoid valve The valve would not close The Solenoid valve was replaced. The valve was was defective on Isolation signal 5749-78 Valve cycled three times.

(1-5741-A)

Corrective RO 78-33/03L-0 Rx. Bldg. Vent The Solenoid valve The valve would not close The Solenoid valve was replaced.

was defective on Isolation signal.

5750-78 Valve (1-5741-B) i e

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UNIT 2 MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS

• ON ACTION TAKEN TO OF V.R. LER MALFUNCTION SAFE OPERATION PREVENT REPETITION NUMBER NUMBER COMPONENT Tile VALVE WOULD NOT TIIE VALVE WAS STROKED n.'O Drywell Vent NO PROBLEM Corrective OPEN WIT 110UT llELP TIMES AND WORKED PROPERLY.

VALVE COULD BE FOUND 5338-78 (2-1601-23)

Refuel Bridge The lloist loading The lloist was not working The relay was replaced an Corrective ccIl relay was properly, the associated interlock 5458-78 (2-834) .

checks were made.

burnt out.

The Solenoid Valve The Valve would not open The Solenoid Valve was Corrective RO 78-38/03L-0 Drywell Vent replaced. The air Valve was defective 5382-78 throttic valve was (2-1601-23) adjusted. The valve was cycled three times.

The Solenoid Valve The Valve would not open The Solenoid Valve was Rx. Bldg. Vent replaced.

Corrective was defective. on isolation signal 5751-78 Valve.

(2-5741-B) 4 I

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IV. LICENSEE EVEilT REPORTS The following is a tabular summary of all license event reports for Quad-Cities Units one and two during the reporting period which were submitted to the commission pursuant to the reportable occurrence reporting requirements as set forth in sections 6.6.B.I. and 6.6.B.2. of the Technical Specifications.

UNIT ONE Licensee Event Date of ._

Report Number Occurrence Title of Deviation 12-18-78 HPCI High R0-78-32/03L Drywell Pressure

- Initiation Switch 12-19-78 Reactor Building R0-78-33/03L Vent Dampers Failed UNIT TWO Licensee Event Date of Report Number Occurrence Title of Deviation 12-18-78 HPCI High Drywell R0-78-39/03L Pressure Initiation Switches R0-78-40/03L 12-21-78 Main Steam Line Pressure Switch

._. Drift - - . - - - - - - - - - - _ _ . . - _ . . . _ _

PS-2-261-30C O

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- , = . . . , -

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

A. Operating Data Report B ~. Average Daily Unit Poaer Level C. Unit Shutdoans and Power Reductions e

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F:::.=rua[y 1;;'y I .

OT ER E.T!;f 0.'.: A R"PCRT .

DOCI'.ET ID. ._050-254 U!!!T ONE DATE y .g_7o h

C0;iPLETED DY n . Ih n n,-

TELEPHONE ' (309)654-2241 0?EP.ATit;G STATUS 0001-781201 P.cporting pcrio3400-781231 Gross hours in reporting pccicd: 744 J.

Currently author-ized pc.ier level (in! ) : 2511 liax. depend. capacity 2.

(MWe-i?c t) :

76c,* Design electrical rating (irle-fic t) : 789 i .

NA Power level to which restricted (if any) (iiMe-!'et):_

3 .

4. Reasons for restriction (if any): .

This ponth Yr. t.o Date Cuculative 744 8461.0 47332.9 5 Number of hours reactor was critical 0 0 3329:6

6. Reactor reserve shutdo.en hours 744 8317.1 aao R o 7 Hours gendrator on line 0 _45.2 8R4.4-

8. Unit reserve shutdown hours.

1195827 ._, 16530801 8880]&7,3_ _ _ _

9., Gross ther:aal energy generated (ItVM) 363526__ 5148779 28572495

10. Gross electrical engergy generated (10S!)

321464 4721055 26632573 .

Net electr.ical Energy Generated 11.

100 96.6 80.4

12. Reactor service factor

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13 Reactor avai lebi l i fy fa.c. tor, inn 96.5 _33,9 .

14. Unit service factor inn oa_o _ 7A 3 Unit 5vailability factor inn _ 95.5 _ __77.8 _

15

16. Unit capactly f actor (using MDC) s6__2 __7n_i Sn n

17. Unit ce.pacity factor (Using Des. IP.!c) 54.8 ._.6 8_. 3 _

57.i

10. L'ni t forctd outcge rate 0,g_ - ,_.3,L_ __

_ R _7 __

19 Shutdo..m schedule J cver ne>.t 6 r.enths (Typc, date, and duratiori c.i cach):

20. If shutde..n at end of repoct p.riod, estimated darc of startop,;., .

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• 7:0 14 tv:y t lu.<2 r t h ,n 709 f r.4 du r i r: ; pe r i od s o f hi gh 0+ i m t tcm.nratt:re dde t, th th:rc : .. rio. r: ,cc of t:.. ,':ni c::n= ! . ..' , :' * .

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OTEr./,T8:,.i D.'.TI. P.CF CRT .

DOCi'.ET 1;3. 050-265 Ui!!T Th'o

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[g \k@h DATE l-9-79 '

D 6 9 \ D. Hannum d C0!!PLETED DY TELEPH0!!E (309) 654-2241 0?EFATiliG STATUS 0001-781201 2400-781231 Gross hours in reportin.3 pc.-iod: 744 J. Reporting period:

Currently authorized power level (entt) : 2511 __ Max. depend. capacity

2. 769 (MWe-Ne t) :

769n Design electrical rating (n'..'e-fict): , .

3 Power level to which restricted (if any) (ilUe-tiet): NA

4. Reasons for restriction (if any): .

.This Month Yr. to Date Cumulative

~ .

744 7171.7 45376.4 5 :iunber of hours reector was critical 0 113.1 2985.-8

6. Recctor reserve shutdown hours 744 7025.1 43102.9 7 Hours gencirator on line 0 128.2 702.9

8. Unit reserve shutdown hours. '

1687219 14877931 88742124 c.,, . , Gross thermal energy generated (MWH) 533878 4656862 28488377

10. Gross electrical engergy generated (MWH)_

508610 4426512 26761460 Net electrical Energy Generated

11. _

100.0 81.9 78.3

12. Reactor service factor .

R1 9 gal 13 P.cector avai labi l i fy fac. tor. inn n

'100.0 80.2 _ 74.3

14. Unit service factor _

Unit availability factor 100,_Q _ 81.7 _75.6 _

15 88.9 65.7 60.0

16. Unit capactly factor (Using MDC)

__64.0 58.5

17. Unit capacity factor (Using Des. M'.!c) _

86.6__ __

_ q,q,,_ (,l_,,,,_, ,,}o,7

10. Unit forced outage rate

19. Shutdo.-ras sch :dulr d over ne>.: f. xnths (Typc,date, and cioration c.f each):

20. I f shutdown at er d of report period , es tima ted dat e of s tartop,* .

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' T h I.?'. f.v.y b, l owe r t ha ri ' f.) !nic <h:r i ng : c r i oP (. T h i g h c'h i rn t tem, m rature dua th:rcal r. r fc rr c r,.:e o f t h,- arra/ canal. , ,

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M!. itu; f !!.'t !: *, U.l* l i - P '. . . i.EV E L June 1976 Docket Ho. ,

050-254 s

Unit ONE

% i 1-4-79

. Date Yi f.

Completed by D. Hannum 9(s Telephone- (309)654-2241 f!ONTH DECEMBER DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL-(liWe-Het)

(!!We-Met) 477 17 434

},

2, 470 jg, 455 470 3 9,. 422 3,

476 20, 428 4.

497 21 428

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22. A9a

6. 479 .

467 23, 420 7

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8. 453 . 24. 417

-.- 25.

472 '419 9 - . ,. -.. _ .

444 26. 422 10.

440 27 409 -

11.

439 28. 365

12. -

433 29 344 13 439 30. 350 14.

438 33, 353

15. aeeauVED 411

16. JUN 2 01976 INSTRUCTIONS

')

On this forni, list the average dai!y unit power Icyd in ?.iWe Net for each day in the reporting nearest veho!c inesswatt.

These figurc.s will te tsed to plot a paph fcir cadi seponing month. Note slot v. hen nuxiinu[n depend taed tur the nel ricetrical rsting of the unit. it.:re nuy L: occasions when. the daily average power I vel c).ceed'. t 1002 lioe (or the iestraciett puwcr !ad I;:ic). In sud enes, the aver.q,e daily tant power output sheet shou!J be featnoted tu err!.:in th. ; ;.areni ani.:io'y.

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APPii!O ly ,; -

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.E'J E L Ju

e 1976

' 7. MC E !' li!: Y U.N il-P Docket !!o. 050-265 Unit TWO Date 1-4-79

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s D Completed by D. Hannum 0%DT Telephoac (309)654-2241 110 NTH DECEMBER DAY AVERAGE DAILY POWER LEVEL.

DAY AVERAGE DAILY POWER LEVEL-(th-Net) (HWe-llct) 17 704

1. 738 717 18, 698 2.

736 19.. 680 3

742 20. 692 4,

61 691 21 5 - . - - - .

750 22. 688

6. .

738 684 7, 23 720

24. 680

8. -

.746 23 681 ..

9 600 26. 684 10.

674 -

~

.I1. 702 27 712 28. 674

12. .

716 29 672 13

14. 706 30. 364

15. 709 31, 432

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16. 701 JUN 2 01976 INSTRUCTIONS On this form, hst the average dii!y unit power I:v.l in .'.ihh*ct for cath day in the repo: ting moh

~) flC21est WIlO!C la pJW.Ti!.

These fit . ore.s will t.c tr<cd to ptut a paph for c::ch sepo:iing manth. Note that when nuxicaum dep:ndsb!c caiucity tued tus the net eteettial ratiny, of the unit there i:ny la creasians whesi the d.tily ascrar.e power fmt excee&, the IMt hoe (or the nextri.ted ;.swer lesel h: c). la v:d> c.'s:2. the aver.ir.c d.iity unit pa.ser output sheet shot.id I; IWtnoted t-) crrhia ti:. .ipp.iren: :: ion'.dy.

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APPENDlX D Q.TP 300-513 UNIT SHUTDOWNS AND POWER REDUCTIONS Revision 5 050-265 DPR-30 March 1978 DOCKET NO. .

COMPLETED BY D_. Hannum UN!T NAME Quad-cities Two _ _ _ ,

REPORT MONTH TELEPHONE (309)654-2241 DATE 1-4-79 December ,

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.- m = zo LICENSEE gg gg ,'

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'p DURATION $" p { ,5 u$ ' EVENT gg S:g 3 CORRECTIVE ACTIONS /C0tF.EllTS

!!O . DATE '-

(HOURS) g5g REPORT NO.

R 30 12-10-78 S load was reduced to 400 IMe for MSIV surveillance tests.

31 12-30-78 F load was reduced to 200 MWe for control rod sequence change

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Vil REFU ELil1G ' I!;FORfMTION The following information about future reloads at quad Cities Station was i 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.

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QTP 300-S32

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Revision 1 Harch .1978 ,

QUAO-CITIES REFUEiIf:G -

INFORiRTION REQUEST i.-

Reload: 4 _ Cycle: 5

1. Unit: 1 January 12. 1979 (Shutde.<n

2. Scheduled date for next refueling shutdown: EOC4)

Ap r'i l 12. 1979 (Start'up 80:5 3

Scheduled 'date for restart following refueling:

4. Will refueling or resumption of' operation thereafter require Yes; See attached a technical checklist specificatica change or other license amendment: -

.for Tech. Spec. and' License Amendment. ,

scheduled date(s) for subm1tting proposed licensing action and supporting 5

informa tion:- The QC1 R4 licensing. submittal is scheduled for' Nov. 11, 1978.

- ~, e.g., new or 6.

Importan't licensing considerations associated with refueling,

• different fuel design or supplier, unreviewed design or performance anaiysis methods, significant changes in fuel design,,new operating . procedures:

N_ew fuel designs: .

Retrofit 8 x 8 fuel (192) a) nat. U at bund _le top and bottom. -

b) two larger water rods,

' ' - c) new enr,ichment distribution. ,

Last Test Assemblies (14 s s ,

for GE PCI-resistant design development program. _

..____.j ___ __

7 The number of fuel assemblies. . ...

72 4

a. Number of assemblies in core: -

ISI

b. Number of assemblies in spent, f'oel pool: _

8. The present l'icensed spent fue'l pool . storage capacity and the size of any increase. in licensed storaga capacity that has been requested or is planned in number of fuel assemblies:

1460 a ." Licensed storage capacity for spent fuel: _

None

b. Planned increase In licensed storage: ,

refucting that can be discharged to the 9 The projected date of th.: last spent fuel jioo! asso.. tin; U: .: present licensed ccpacity: Last refueling (cad of batch discharge capability) date with present capacity: March, 85 I

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1

, 1 1

- . - ,, QTP 300-533 cEE@

RELOAD LICENSING PACKAGE * '

Revision 1 g -

PREPARAT10N SCHEDULE Narch 1973 t UNIT 0.1 * @

4 RELOAD CYCLE ~5 , &=3 m

.- ACTIVITY -

RESPONSIBILITY CENTER .

DATE "

M C9)

CE NFS roccives draft Licensing Submittal from GE R '

/78 NFS Transmit copy of draf t to Station for Comments - .

(BL Transmit NFS and Site comments / questions to GE NFS B 6 a i n Te ch . S of:.c r.banges_, S a f e t y F vn l ii o i nn ad Ce va,r.- Lc.i t ""-

9/78 NFS NFS receives final Licensing Submittal and answers to CECO questions and ensucre en rFrh ni,xf t.jo rom vGE GE Complete final NFS review of 11cnne,ing Lihmitent 30/78 NFS ^

NFS Transmit complete package for on/off site review 1/) 3 Station On-site review completed

/3/78 PSA Off-site review comninted

/6 //8 Completed licensino nnekngn enen t end by Mn ' .

I

/11/78 NLA rp ,

• 90 days 1

12/79

- .' Anticipated.unic shutdown .

28 days  ? -

l s . .

- ' Receipt of operating License .

j d/ +

/9/79 s/

Antic.1patcid'. Unit Startu'p - Assumes 56 day outage . .

/9/79 ..

8 weeks

. _, lI nc NFS/ BUR -

. .. . Prepared by - .

Datc* 1g/23/77' .-

j PRELIMINARY Cl!ECKLIST FOR RELCAD LICEMSE Anli!I

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hNIT. Ouad-cities 1 i ' i RELCAO: 4' i ,

j CYCLE: 5' ) . . .......... .......

Require Changes I tem pagg .

Generalize wording and Scram Reactivity 4 .

X reference the submit. ,

' .- FlE DO-XX'/YX

^

Hone. Adequate. pressure

• Safety Valve Setpoints raa rgin . -

1.2/2.2-1 KA LSSS

!!one, if the' peak vessei ,

Bases 1.2/2.2-2.3 ' pressure is 1325 psig. during .

tlA -

S.V. sizing trans.

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RBM Setpoints ' ' Change to (.65w+XX) as regid. .

X' LCO 3 2/4.2-14 Change operability to Xxt

- - 3 2/4.2-7 Change Reference I to llEDO-X ' Bases 3 2/4.2-8 XXXXX. ,

t A'uto Flow Control hone. Stability analysis not NA LCO .

3 3/4.3-5 limiting.

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None. . ~

NA Bases 3 3/4.3-11 -

s

~. HAPLHCP.

Fig. 3 5.1

• Revise curves to reflect new analyses.

X LCO

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' (sh t s. I to 3) .

• Change references to reflect X Bases :3.5/4.5-14

.new analyses of NE00-24046.

' tiew values: **1.XX (7 x 7)

MCPR '3 5/4.5-10 .

1.XX (8 x 8) .

X LCO Generalize description of B Bases 3 5/4.5-14 1imiting tren s ien t (s) .

$ l T~ ?LHGP. cha ups are being handled under separate cover.CPP. p.:n

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"  ! Nlt' der, a:'di t i ena ! C. XX

!i.ensinj ;asition). D""D U X ititeri.

o. o aSb $ 1a a

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. r QTP 300-532 l' '

Revision 1 -

llarch 1978 '

QUAD-CITIES REFUELillG JNFORMATION REQUEST fi Cycle: _ c; (next outage) 2 Reload: 4 ,_.

1. Unit:

Scotember 30.1979 (shutdown 2.

Scheduled date for next refueling shutdown: EOC4) i .

January' 20,1980 (Startup' 3

Sc'heduled date for restart following refueling: BOC5) 4.

Vill.' refueling or resumption of operation thereafSimlar ter require Tech. Spec. a technical changes -

specification change or other license amendment: -

- to Reload 3. Cycle 4. .

Scheduled date(s) for submitting proposed licensing action and supporting -

5 Reload Submittal to be provided approximately 90 days prior ,

. informa tion: '

to shutdown. .

e.g., new or

4. 6.

• different Important licensing considerations associated with refueling, fuel design or supplier, unreviewed design or performance analysis nethods,,significant changes in fuel design, new operating procedures:

Hew fuel designs: Retrofit 8 x 8 fuel (approximately 196).

7 -

s. -

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4 . .

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7. . The number of fuel assemblies.

. 724

a. Ilumber of assemblics in core:

745 -

. b. Number of assemblics in spent fuel pool:

- 8. The present licensed spent fuel pool storage capacity and the size of any incrqase in licensed storage capacity that has been requested or is planned * '

in number of fuel assemblies: ~

Licensed stcrage capacity for spent fuel: .1460-

a. . .

None.

b. Planned increase in licensed storage: '

refueling that cca be discharged to the S. The projected date of the last liter. sed capacity: Lost refueling date with spent fuel pool asr.ca:ng the present -

Present capacity: Septernher. 85 D

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+ D ]& u

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M QTP 300-533 N

RELOAD LICENSING PACKAGE' r

Revision i b UNIT OC 2

- PREPARATION SCHEDULE March 1973 g 3

RELOAD 4

b CYCLE  !

ACTIVITY g

RESPONSIBILITY CENTER OATE i

cag to Station @

10/6/77 GE NFS roccives draf t Licensing Submittal from GE Transmit copy of draf t @ .

UFS for comments. .

i g

Transmit NFS and Site comments / questions to.GE Begin Tech. Spec. changes, Safety

• NFS 10/20/77 NFS Evaluation and Cover Letter.

CC NFS receives final Licensing Submittal and answers ,na to CECO answers questions from CE.

tn CFCo ottos* ion -

NFS Complete final NFS review of T.4 consinc 4tthmfeen1 11/3/77 NFS Transmit complete package for on/off site review 11/8/77 i on-site review completed 11/16/77 Station PSA Off/ nite review completed 11/18/77 NLA Completed Licensine packace reco4vnd by NRP 90 days 12/1/77 -

I 1/16/78 - Anticipated unit shutdown 28 days .

3/5/78 - Receipt of operating License 9 ..

I Day outage i Anticipated Unit Startup - Assumes , 8 Weeks 3/15/78 -

~

. Prepared by JAS . NFS/BWR Datc 2/21/78 Q

. PREL!MiliARY Cl!ECKLIST FOR 'P.ELOAD L ICE ll$E IJtE!!DP '

U;;lT: 0.uad-Cities 2 RELCAO: 3' l ,

CYCLE: 4'l ..................

I Require Changes Iten page t ,

i -

Generalize vording and Scram Reactivity 4.

..X - reference the submit,

[1E00-24063 -

Safety Valve Setpoints  !!one. Adequate pressure .

1.2/2.2-1

!!A LSSS

\. margin. ,

1.2/2.2-2,3 .

C'larify and add bounding '

(X Bases '

peak' pressure. <

) .

~ . .

RBH Setpoints Change to (.65w+42) ..

X LCO 3.2/4.2-14 Change operability to-30%.

3.2/4.2-7 Change Reference 1 to X tiases 3.2/4.2-8 .

IJEDO-24063

- t .

Auto Flow Control t!one. Stability analysis IIA LCO .3 3/4.' 5 not limiting.

3 3/4.3-11 tione. ,

tA Dases -

- . . 9 i

1 -

PAPLHGR Fig. 3 5.1  :

• Revise curves to reflect -

X LC0 new analyses.

(shts. I to 3).

• Change references to reflect X Bases . 3 5/4.5-14 new analyses of flE00-24046.

~

W

.. 'a. . . . .

73.5/4.5-10 tiew values: **1.33 (7 x 7)

. IlCPR ~

1.35 (8 x d)

.X LCO -

ceneralize description of Bases 3 5/4.5-14 ~

limiting transient (s) .

FA?lliOR changes are being handled under separate cover.CPR penalty fo ~

• ' includes addi tion 2! 0.XX 0 'Dl h bo'C in:crin licca,ing por.ition). 0

• J

. . .,,..yy ,, Wu u

~

Vill CLOSSARy ,

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

CRD - Control Rod Drive System SBLC

- Standby Liquid Control System MSlY Main Steam isolation Valve .

RHRS - Residual Heat Removal System' RCIC - Reactor Core Isolation Cooling System HPCI - High Pressure Coolant Injection System .

SRM - Source Range Moni tor i

IRH - Intermediate Range Monitor LPRM - Local Power Range Monitor APRM - Average Power Range Monitor TIP - Traveling incore Probe RBCCW

- Reactor Bbilding Closed Cooling Water System

.TBCCW

- Turbine Building Closed Cooling Water System RVM - Rod Worth Minimizer

~ - - - - - - - - -

SBGTS

- Standby Gas Treatment System ~ - ~ ~ - - - - - -

HEPA

- High-Efficientry Particulate Filter RPS

- Reactor Protection System IPCLRT - Integrated Primary Containment Leak Rate Test LPCI - Low Pressure Coolant injection Mode of RHRS R8H Rod Block Monitor BVR

- Boiling Water Reactor ISI - In-Service Inspection llPC -

Maxinum Permi ssabic Concentra tion

PCI

- Primary Containment isolation

- Shutdo.vn Cooling Mode of RHRS SDC LLRT

- Local Leak Rate Testing MPLHGP.

- Maximum Average Planar Linear Heat Generation Rate ,

R.O.

Reportable Occurrence DW Drywell PJ(

Reactor EHC - Electro-Hydraulic Control System MCPR -

Minimum Critical Power Ratio FOMR - Preconditioning Interim Operating Management Recommendations ,

LER

- Licensee Event Report ANSI American National Standards Institute NIOSH

- Nacional Institute for Occupational Safety and Health ACAD/ CAM

- Atmospheric Containt, .it Atmo pheric Dilution / Containment Atmospheric Monitoring e+e e e- ha e

- -.-. -- e m.=,-><-

_ , _ , e ya. -