ML19274D384
| 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
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
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LICENSE NOS. DPR-29 and DPR-30 ' ~~~~ -~ ~ ~
~~~~~~~
~~ ~ ~ ~ ~ ~ ~~
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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 A small increase in power was realized when the pressure.
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.
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UNIT I
MAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS W.R.
LER OF ON ACTICN TAKEN TO
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NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Refuel The Limit The Interlock System The Limit Switches Switches Needed Limit Switch llolding were Replaced Corrective 4769-78 Bridge Recalibration Plate was moved were Adjusted (833/834),
requiring adjustment of i
the Switches.
15 Diesel
'The Pressure The Breaker was A Different Set of 5543-78 Generator Switch contacts Tripping Out.
The D. G.
Contacts on the Pressure Corrective (h-6601) were Defective was still Operable Switch were connected.
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 5749-78 Valve was defective on Isolation signal replaced. The valve was 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 5750-78 Valve was defective on Isolation signal.
replaced.
(1-5741-B) i e
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UNIT 2
MAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS V.R.
LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Corrective Drywell Vent NO PROBLEM Tile VALVE WOULD NOT TIIE VALVE WAS STROKED n.'O 5338-78 VALVE COULD BE FOUND OPEN WIT 110UT llELP TIMES AND WORKED PROPERLY.
(2-1601-23)
Corrective Refuel Bridge The lloist loading The lloist was not working The relay was replaced an 5458-78 (2-834) ccIl relay was
- properly, the associated interlock checks were made.
burnt out.
Corrective RO 78-38/03L-0 Drywell Vent The Solenoid Valve The Valve would not open The Solenoid Valve was replaced. The air 5382-78 Valve was defective throttic valve was (2-1601-23) adjusted. The valve was cycled three times.
Corrective Rx. Bldg. Vent The Solenoid Valve The Valve would not open The Solenoid Valve was 5751-78 Valve.
was defective.
on isolation signal replaced.
(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 R0-78-32/03L 12-18-78 HPCI High Drywell Pressure Initiation Switch R0-78-33/03L 12-19-78 Reactor Building Vent Dampers Failed UNIT TWO Licensee Event Date of Report Number Occurrence Title of Deviation R0-78-39/03L 12-18-78 HPCI High Drywell Pressure Initiation Switches R0-78-40/03L 12-21-78 Main Steam Line Pressure Switch
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PS-2-261-30C O
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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 h
DATE y.g_7o C0;iPLETED DY n. Ih n n,-
TELEPHONE
' (309)654-2241 0?EP.ATit;G STATUS 0001-781201 J.
P.cporting pcrio3400-781231 Gross hours in reporting pccicd:
744 2.
Currently author-ized pc.ier level (in! ) :
2511 liax. depend. capacity (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) 10.
Gross electrical engergy generated (10S!)
363526__
5148779 28572495 321464 4721055 26632573 11.
Net electr.ical Energy Generated 100 96.6 80.4 12.
Reactor service factor 13 Reactor avai lebi l i fy fa.c. tor, inn 96.5
_33,9
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14.
Unit service factor inn oa_o _
7A 3 15 Unit 5vailability factor inn _
95.5
__77.8 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,;.,
- 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 I
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OTEr./,T8:,.i D.'.TI. P.CF CRT DOCi'.ET 1;3.
050-265
\\k@h Ui!!T Th'o
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D. Hannum d
C0!!PLETED DY TELEPH0!!E (309) 654-2241 0?EFATiliG STATUS 0001-781201 2400-781231 J.
Reporting period:
Gross hours in reportin.3 pc.-iod:
744 2.
Currently authorized power level (entt) :
2511
__ Max. depend. capacity (MWe-Ne t) :
769n Design electrical rating (n'..'e-fict):
769 Power level to which restricted (if any) (ilUe-tiet):
NA 3
4.
Reasons for restriction (if any):
.This Month Yr. to Date Cumulative
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744 7171.7 45376.4
- iunber of hours reector was critical 5
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 11.
Net electrical Energy Generated 100.0 81.9 78.3 12.
Reactor service factor 13 P.cector avai labi l i fy fac. tor.
inn n R1 9 gal
'100.0 80.2 _
74.3 14.
Unit service factor 15 Unit availability factor 100,_Q _
81.7
_75.6 88.9 65.7 60.0 16.
Unit capactly factor (Using MDC) 17.
Unit capacity factor (Using Des. M'.!c) 86.6__
__64.0 58.5 10.
Unit forced outage rate
_ q,q,,_
(,l_,,,,_,
,,}o,7 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,*
I tem, rature dua
' f.) !nic <h:r i ng : c r i oP (. T h i g h c'h i rn t m
' T h I.?'. f.v.y b, l owe r t ha ri to d th:rcal r. r fc rr c r,.:e o f t h,- arra/ canal.
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June 1976 M!. itu; f !!.'t !: *, U.l l i - P '... i.EV E L Docket Ho.
050-254 s
Unit ONE i
Date 1-4-79 Yi f.
D. Hannum 9(s Completed by Telephone- (309)654-2241 f!ONTH DECEMBER DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL-(!!We-Met)
(liWe-Het) 477 17 434
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470 jg, 455 2,
422 470 3 9,.
3, 4.
476 20, 428 497 21 428
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479 7
467 23, 420
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8.
453 24.
417 9
-.- 25.
'419 472 444 26.
422 10.
11.
27 409 440 439 28.
365 12.
13 433 29 344 14.
439 30.
350 438 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.
I s, f,i na l )
I:c'/ i r. i on 4 APPii!O ly,; -
Ju: e 1976
' 7. MC E ! li!: Y U.N il-P
- .E'J E L Docket !!o.
050-265 Unit TWO
. h Date 1-4-79 D
Completed by D. Hannum 0%DT s
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.
3 736 19..
680 742 20.
692 4,
691 61 21 5
6.
750 22.
688 738 684 7,
23 680 720 24.
8.
9
.746 23 681 10.
600 26.
684
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702 27 674 12.
712 28.
674 716 29 672 13 14.
706 30.
364 15.
709 31, 432
.a e e a o V E D 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
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flC21est WIlO!C la pJW.Ti!.
These fi. 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 t
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 _ _ _
TELEPHONE (309)654-2241 DATE 1-4-79 REPORT MONTH December s
e:
M seR m
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m LICENSEE gg gg Ee 8
8 { u$ '
'p DURATION $
p
,5 EVENT gg S:g 3
CORRECTIVE ACTIONS /C0tF.EllTS g5g REPORT NO.
(HOURS)
!!O.
DATE R
load was reduced to 400 IMe for MSIV 30 12-10-78 S
surveillance tests.
load was reduced to 200 MWe for control 31 12-30-78 F
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.-
1.
Unit:
1 Reload:
4
_ Cycle:
5 January 12. 1979 (Shutde.<n Scheduled date for next refueling shutdown:
EOC4) 2.
Scheduled 'date for restart following refueling:
Ap r'i l 12. 1979 (Start'up 80:5 3
Will refueling or resumption of' operation thereafter require a technical Yes; See attached checklist 4.
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.
~,
Importan't licensing considerations associated with refueling, e.g., new or
- different fuel design or supplier, unreviewed design or performance anaiysis 6.
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 Number of assemblies in core:
a.
ISI b.
Number of assemblies in spent, f'oel pool:
The present l'icensed spent fue'l pool. storage capacity and the size of any 8.
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 date with present capacity:
March, 85 (cad of batch discharge capability)
. h I
1 1
1 QTP 300-533 cEE@
RELOAD LICENSING PACKAGE Revision 1 g
PREPARAT10N SCHEDULE Narch 1973 t
UNIT 0.1 4
&=3 RELOAD ~ 5 CYCLE m
.- ACTIVITY RESPONSIBILITY CENTER DATE M
C9)
NFS roccives draft Licensing Submittal from GE R
- /78 CE Transmit copy of draf t to Station for Comments (BL NFS Transmit NFS and Site comments / questions to GE 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 ""-
NFS NFS receives final Licensing Submittal and answers to CECO questions f rom GE 9/78 NFS rFrh ni,x t.jo v en and ensucre GE Complete final NFS review of 11cnne,ing Lihmitent 30/78 NFS
^
review Transmit complete package for on/off site 1/) 3 NFS On-site review completed
/3/78 Station
/6 //8 PSA Off-site review comninted Completed licensino nnekngn enen t end by Mn '
I rp, *
/11/78 NLA 90 days 1
.' Anticipated.unic shutdown 28 days
?
12/79 l
s
' Receipt of operating License d/
+
j
/9/79 s/
Antic.1patcid'. Unit Startu'p - Assumes 56 day outage.
/9/79 8
weeks l
I NFS/ BUR
. Prepared by -
nc Datc*
1g/23/77'.-
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PRELIMINARY Cl!ECKLIST FOR RELCAD LICEMSE Anli I
hNIT.
Ouad-cities 1 i i
RELCAO:
4' i j
CYCLE:
5' )
Require Changes I tem pagg Generalize wording and 4
X Scram Reactivity reference the submit.
FlE DO-XX'/YX Hone.
Adequate. pressure
^
Safety Valve Setpoints 1.2/2.2-1 raa rgin.
!!one, if the' peak vessei tlA Bases
' pressure is 1325 psig. during S.V. sizing trans.
RBM Setpoints 3 2/4.2-14
' Change to (.65w+XX) as regid.
~
Change operability to Xxt X'
LCO 3 2/4.2-7 Change Reference I to llEDO-3 2/4.2-8 X
' Bases XXXXX.
t A'uto Flow Control hone.
Stability analysis not 3 3/4.3-5 limiting.
NA LCO
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~
None..
3 3/4.3-11 NA Bases s
~.
HAPLHCP.
- Revise curves to reflect Fig. 3 5.1 X
LCO
~
' (sh t s. I to 3) new analyses.
- Change references to reflect
- 3.5/4.5-14
.new analyses of NE00-24046.
X Bases tiew values:
- 1.XX (7 x 7)
' '3 5/4.5-10 1.XX (8 x 8)
MCPR X
LCO Generalize description of 3 5/4.5-14 B
Bases 1imiting tren s ien t (s).
l T~ ?LHGP. cha ups are being handled under separate cover.CPP. p.:n
! Nlt' der, a:'di t i ena ! C. XX D""D
~DlD K f
!i.ensinj ;asition).
aSb $ 1a U X ititeri.
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r QTP 300-532 Revision 1 l'
llarch 1978 QUAD-CITIES REFUELillG JNFORMATION REQUEST Reload:
4 Cycle:
_ c; (next outage) fi 1.
Unit:
2 Scotember 30.1979 (shutdown Scheduled date for next refueling shutdown:
EOC4) 2.
i.
January' 20,1980 (Startup' Sc'heduled date for restart following refueling:
BOC5) 3 4.
Vill.' refueling or resumption of operation thereaf ter require a technical Simlar Tech. Spec. changes -
specification change or other license amendment:
to Reload 3. Cycle 4.
Scheduled date(s) for submitting proposed licensing action and supporting Reload Submittal to be provided approximately 90 days prior 5
informa tion:
to shutdown.
Important licensing considerations associated with refueling, e.g., new or
- different fuel design or supplier, unreviewed design or performance analysis
- 4. 6.
nethods,,significant changes in fuel design, new operating procedures:
Retrofit 8 x 8 fuel (approximately 196).
Hew fuel designs:
7 s.
I." '
4 f.'
g v
- 7.. The number of fuel assemblies.
Ilumber of assemblics in core:
724 a.
745 -
Number of assemblics in spent fuel pool:
b.
licensed spent fuel pool storage capacity and the size of any 8.
incrqase in licensed storage capacity that has been requested or is planned The present in number of fuel assemblies: ~
.1460-Licensed stcrage capacity for spent fuel:
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
[sAkkf DNK D ]&
D u
+
f M
N QTP 300-533 b
RELOAD LICENSING PACKAGE' r
Revision i
- PREPARATION SCHEDULE March 1973 g
UNIT OC 2 RELOAD 3
b CYCLE 4
g ACTIVITY OATE RESPONSIBILITY CENTER cag i
to Station @
NFS roccives draf t Licensing Submittal from GE Transmit copy of draf t 10/6/77 GE i
g UFS for comments.
Transmit NFS and Site comments / questions to.GE Begin Tech. Spec. changes, Safety NFS Evaluation and Cover Letter.
NFS receives final Licensing Submittal and answers to CECO questions from CE.
10/20/77 NFS
,na answers tn CFCo ottos* ion -
CC 4tthmfeen1 Complete final NFS review of T.4 consinc 11/3/77 NFS Transmit complete package for on/off site review 11/8/77 NFS i
on-site review completed 11/16/77 Station 11/18/77 PSA Off/ nite review completed Completed Licensine packace reco4vnd by NRP 12/1/77 NLA 90 days I
Anticipated unit shutdown 1/16/78 28 days Receipt of operating License 3/5/78 9..
I Day outage i
Anticipated Unit Startup - Assumes 8
Weeks 3/15/78
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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 4.
Generalize vording and
..X Scram Reactivity
- reference the submit,
[1E00-24063 Safety Valve Setpoints 1.2/2.2-1
!!one.
Adequate pressure
!!A LSSS
\\.
margin.
1.2/2.2-2,3 C'larify and add bounding
( X Bases peak' pressure.
)
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RBH Setpoints 3.2/4.2-14 Change to (.65w+42) 3.2/4.2-7 Change operability to-30%.
X LCO 3.2/4.2-8 Change Reference 1 to X
tiases IJEDO-24063
- t Auto Flow Control t!one. Stability analysis
.3 3/4.' 5 IIA LCO not limiting.
3 3/4.3-11 tione.
tA Dases 9
i 1
- Revise curves to reflect PAPLHGR Fig. 3 5.1 X
LC0 (shts. I to 3).
new analyses.
3 5/4.5-14
- Change references to reflect new analyses of flE00-24046.
X Bases
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. W 'a....
tiew values: **1.33 (7 x 7)
IlCPR 73.5/4.5-10 1.35 (8 x d)
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.X LCO 3 5/4.5-14 ceneralize description of Bases
~
limiting transient (s).
FA?lliOR changes are being handled under separate cover.CPR penalty fo
~
- ' includes addi tion 2! 0.XX bo'C in:crin licca,ing por.ition).
0 0
'Dl h
J u
Wu
.,,..yy
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Vill CLOSSARy The following abbreviation which may have been used in the llontNy Report, are defined below:
Control Rod Drive System CRD Standby Liquid Control System SBLC Main Steam isolation Valve MSlY Residual Heat Removal System' RHRS Reactor Core Isolation Cooling System RCIC High Pressure Coolant Injection System HPCI Source Range Moni tor SRM i
Intermediate Range Monitor IRH Local Power Range Monitor LPRM APRM Average Power Range Monitor Traveling incore Probe TIP RBCCW Reactor Bbilding Closed Cooling Water System Turbine Building Closed Cooling Water System
.TBCCW Rod Worth Minimizer RVM Standby Gas Treatment System ~ - ~ ~ - - - - - -
~
SBGTS HEPA High-Efficientry Particulate Filter Reactor Protection System RPS IPCLRT Integrated Primary Containment Leak Rate Test Low Pressure Coolant injection Mode of RHRS LPCI R8H Rod Block Monitor BVR Boiling Water Reactor ISI In-Service Inspection llPC Maxinum Permi ssabic Concentra tion
PCI Primary Containment isolation SDC Shutdo.vn Cooling Mode of RHRS LLRT Local Leak Rate Testing Maximum Average Planar Linear Heat Generation Rate MPLHGP.
R.O.
Reportable Occurrence Drywell DW PJ(
Reactor Electro-Hydraulic Control System EHC Minimum Critical Power Ratio MCPR Preconditioning Interim Operating Management FOMR Recommendations Licensee Event Report LER ANSI American National Standards Institute Nacional Institute for Occupational Safety and NIOSH Health ACAD/ CAM Atmospheric Containt,.it Atmo pheric Dilution / Containment Atmospheric Monitoring e+e e
e-ha e
e m.=,-><-
e ya.
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