ML20079E274
| ML20079E274 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 12/31/1983 |
| From: | Kalivianakis N, Misak A COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| NJK-84-1, NUDOCS 8401170210 | |
| Download: ML20079E274 (20) | |
Text
,_
QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT DECDfBER 1983 C(NM0NWEALTH EDIS9N COMPANY AND IOWA-ILLINOIS GAS & ELECTRIC COMPANY NRC DOCKET NOS, 50-254 AND 50-265 LICENSE NOS, DPR-29 AND DPR-30 i
8401170210 831231 PDR ADOCK 05000254 R
PM s
TABLE OF CONTENTS I,
Introduction II, Summary of Operating Experience E
A.
Unit One E,
Unit Two III, Plant or Procedure Changes, Tests, Experiments, and Safety Related 'faintenance A,
Amendments to Facility License or Technical Specifications B.
Facility or Procedure Changes Requiring NRC Approval C.
Tests and Experiments Requiring NRC Approval D,
Corrective Maintenance of Safety Related Equipment IV, Licensee Event Reports l
V, Data Tabulations l
A, Operating Data Report l
B, Average Daily Unit Power Level L'
C, Unit Shutdowns and Power Reductions VI, Unique Reporting Requirements A, -Main Steam Relief Valve Operations B,
Control Rod Drive Scram Timing Data VII, Refueling Information VIII, Clossary 1
r.
e 1.
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 Commonweaith Edison Company and lowc-Illinois Gas & Electric Comoanye The Nuclear Steam Supply Systems are General Electric Company Boiling Water Reactors.
The Architect / Engineer was 5 argent & Lundy, Incorporated, and the primary construction contractor was United Engineers & Constructors. The Mississippi River is the condenser cooling water source. The plant is sub!ect 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 One and Two, respectively, were October 18, 1971, and April 26, 1972.
Commercial generation of power began on February 18, 1973, for Unit One and March 10, 1973, for Unit Two.
This report was compiled by Becky Brown and Alex Misak, telephone number 309-654-2241, extensions 127 and 194.
r-II,
SUMMARY
OF OPERATING EXPERIENCE A.
Unit One December 1-10: Unit One began the month operating at full power and maintained this level until 0030 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> on December 4 when the unit dropped load to 700 MWe to perform weekly Turbine tests.
At 0045 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />, the tests were completed and the unit began a normal load increase to full power. At 0115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br />, on December 10, the unit dropped load to 550 MWe for Control Rod Pattern adjustments. At 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> the adjustments were completed and the unit began a normal load increase.
December 11-20: At 0020 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, on December 17, the unit dropped load to 700 MWe for weekly Turbine tests. At 0140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> the tests were completed and a normal' load increase to full power was initiated.
At 1315 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />5.003575e-4 months <br />, on December 17, the unit dropped load to 690 MWe to switch Condensate pumps due to excessive Icakage trom the inboard seal on the ID Condensate pump.
At 1340 hours0.0155 days <br />0.372 hours <br />0.00222 weeks <br />5.0987e-4 months <br />, the switch was completed and the unit began increasing load normally.
December 21-31: At 0030 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, on December 25, the unit dropped load to 700 MWe for weekly Turbine tests anc following their completion, at 0320 hours0.0037 days <br />0.0889 hours <br />5.291005e-4 weeks <br />1.2176e-4 months <br />, the unit began increasing load normally. The unit maintained full power until 2215 hours0.0256 days <br />0.615 hours <br />0.00366 weeks <br />8.428075e-4 months <br /> on December 31 when the unit i
. began dropping load at 200 MWe/ hour in preparation for a Dryweli entry to investigate the cause of increased leakage nonitoring values.
I B.
Unit Two Unit Two remained shutdown throughout the month for End of Cycle Six Refueling and Maintenance.
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Ill. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A.
Amendnents to Facility License or Technical Specifications There were no Amendments to the Facility License or Technical Specifications for the reporting period.
B.
Facility or Procedure Changes Requiring NRC Approval There were no Facility or Procedure changes for the reporting
- period, C.
Tests and Experiments Requiring NRC Approval There were no Tests or Experiments requiring NRC approva! for the reporting period.
D.
Corrective Maintenance of Safety Related Equipment The following represents a tabular summary of the major safety related maintenance performed on Unit One and Unit Two for the reporting' period. This summary includes the following headings: Work Request Numbers, LER Numbers, Components, Cause of Malfunctions, Results and Effects on Safe Operation, and Action Taken to Prevent Repetition.
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+ - -
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-m.
- y v
W UNIT ONE MAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS
.. 'W. R.
LER OF ON ACTION TAKEN TO NUMBER NUiiBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q30385 83-47/03L I-737-1 There was excess The possibility of an The valve was replaced TIP Ball friction about the uncontrolled release and stroked three times.
Valve valve stem.
was limited by giving the Unit 1 NSO the key to the switch for the in-line shear valve and sending an Operator to close the in-line renual valve.
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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 Q24837 RPS Relay This was a mis-The insulation on the The breaker was rep,Isced Power Breaker application of the wi re was deteriorated.
with one of the correct 590-3068 breaker.
size.
Q28216 83-14/03L 2-203-2A HSIV The seat was worn.
The valve failed the The seat was ground and Outboard LLRT. The in-line 2-the valve was 203-1A valve would successfully LLRT'd.
have provided isolation.
Q28235 PCI-Atm>s.
The seal-tight was The fuse blew twice.
The seal-tight conductors Control Out-broken.
and connectors were board Fuse repai red & replaced.
Q28445 83-16/03L HPCI Area The switch failed in Because of the one-out-A new switch was High Temp.
the closed position, of-two-taken-twice" calibrated and installed.
Switch 2-partially due to logic, this situation The procedure for 2371D excessive heat (2 switches failed testing the switch was applied during closed) would not have changed to reduce the testing.
prevented HPCI heat applied.
isolation, nor caused spurious HPCI isola-tion.
Q28446 83-16/03L HPCI Area See Above.-
See Above.
See Above.
High Temp.
(Q28445)
(Q28445)
(Q28445)
Switch 2-2370D Q28447 83-16/03L llPCI Area There was excessive The other three switches A new switch was liigh Temp.
Instrument drift.
in this area were calibrated and installed.
Switch 2-operable and would have 2370A tripped the alarm.
9 T
e
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IV, LICENSEE EVENT ' REPORTS The following is a tabular summary of all licensee event reports for Quad-Cities Units One and Two occurring during the reporting period, pursuant to the reportable occurrence reporting requirements as set forth in sections 6.6.B.1. and 6.6.B.2. of the Technical Specifications.
UNIT ONE Licensee Event Report Number Date Title of Occurrence 83-48/03L 12-29-83 Chimney Monitor System
. inoperable UNIT TWO 83-24/03L 12-1-83 Hechanical Snubbers Failed Functional Test
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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 l
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l.-
f
OPERATING DATA REPORT DOCKET NO.
50-254 UNIT ONE DATEJonuarv 6 COMPLETED BYAlex Hisak TELEPHONE 309-654-224ix194 OPERATING STATUS 0000 120183 1.
Reporting period 2400 123183 Gross hours in reporting period 744
- 2. Currently authorized power level (MWt): 2511 Max Depend capacity (MWe-Net): 769* Design electrical rating (MWe-Net): 789
- 3. Power level to which restricted (if any)(MWe-Net): NA
- 4. Reasons for restriction (if any):
This Month Yr.to Date Cumulative
- 5. Number of hours reactor was critical 744.0 8384.4 83555.6 6.
Reactor reserve shutdown hours 0.0 0.0 3421.9
- 7. Hours generator on line 744.0 8261.2 80347.9
- 8. Unit reserve shuttiown hours.
0.0 0.0 909.2
- 9. Gross thermal energy generated (MWH) 1806327 18893715 165106706 LO. Gross electrical energy generated (MWH) 597065 6136735 53258616 Li. Het electrical energy generated (MWH) 566248~
5777059,
49605967-
- 12. Reactor service factor 100.0 95.7 81.9
' 13. Reactor avo11ob111ty factor 100.0 95.7_
85 Q
- 14. Unit service factor 100.0_
94.3 78.7
- 15. Unit evollability factor 100.0 94.3 79.6
- 16. Unit capacity factor (Using MDC) 98.6 85.6 63.2
- 17. Unit copocity factor (Using Des.MWe) 96.i 83.4 61.6
- 18. Unit forced outoge rate 0.0 2. 0.
6.3
- 19. Shutdowns scheduled over next 6 nonths (Type,Date,and Duration of each):
- 20. If shutdown at end of report period, estimated date of startup
___NA j
_,________I 8The IOC ser be lower the.n 769 Ilue dering perleds of high onbient tenperatore doe to the thernal perfernance of the sorep canal.
8WilFFICIAL ColHilf 11130ERS ARE USD Ill 1HIS REPORT 6
OPERATING DATA REPORT DOCKET NO.
50-765 UNIT TWO DATEJonuarv 6 COMPLETED BYelex Hisok TELEP HONE 309-6_W4-2241 x i94 OPERATING STATUS 0000 120183
- 1. Reporting period:2_400 123183 Gross hours in reporting period 744
- 2. Currently authorized power level (MWt): 2511 Max. Depend.copacity (MWe-Net): 769* Design electrical rating (MWe-Net): 789
- 3. Power level to which restricted (if any)(MWe-Net): NA 4.
Reasons for restriction (if any):
This Month Yr.to Date Cumulative
- 5. Number of hours reactor was critical 0.0 5654.1 77917 5
- 6. Reactor reserve shutdown hours 0.0 0.0 2985.8 t.
7.. Hours generator on line 0.0 5621.7 75209.8
- 8. Unit reserve shutdown hours.
0.0 0.0 702.9
- 9. Gross thernal energy generated (MWH) 0 10790594 155382088
- 10. Gross electrical energy generated (MWH) 0 3398245 49435780 ii. Het electrical energy generated (MWH)
-836 3150493 46334060_,
- 12. Reactor service factor 0.0 64.5 77.0
- 13. Reactor avo11obility factor 0.0 64.5 80.0
- 14. Unit service factor 0.0 64.2 74.4
- 15. Unit availability factor 0.0 64.2 75.1
- 16. Unit capacity factor (Using HDC)
.2 46.8 59.6
- 17. Unit capacity factor (Using Des.MWe)
. E_
' '4 5. 7 58.1
- 18. Unit forced outage rate 0.0 1.8 8.6
- 19. Shutdowns scheduled over next 6 months (Type,Date,ond Duration of each):
- 20. If shutdown at end of report period,estinated date o f s t ar t u p _ l _19_-84__,____
SThe 14C not be louer then %9 Inie daring periods of high anblent tenperatore due to the thernal perfernance of the spret canal.
80RFFICIN. C5FAllY IRISERS ARE USED Ill TO REPORT
APPENDIX B AVERAGE DAILY UNIT POWER LEVFL DOCKET NO.
50-254 UNIT ONE DATEJonuoru 6 COMPLETED BYAlex Misak TELEPHONE 309-654-2241xi94 MONTH December 1983 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1.
767.0 17, 741.3 2.
770.5 18, 771.8 3.
768.8 19, 771.9 4,
751.6 20.
769.9 5,
760.4 21.
771.8 6.
779.7 22.
777.7 7.
768.2 23, 779.2 G.
770.5 24, 790.0 9.
768.8 25.
735.0 10, 555.0 26.
776.7 s
ii.
658.8 27.
794.3 12.
766.7 28.
764.8 13, 755.9 29.
776.1 14.
759.4 30.
778.0 15.
769.6 31, 766.9 16, 775.2 INSTRUCTIONS D this forn, list the notregt daily snit power level in llle-Met for each day in the reperting nenth. Compete to the neerest uhele negeustt.
These figeres will be used to plot a graph for each reporting nenth. Note that when narinen dependable capacitt is seed for the net electrical rating of the sait there ney be occasion: When the daily everage pwer level exceeds the 1901 line (or the restricted pwer level line).,In sich cases,the everage dolly snit power output sheet should he festnoted to esplein the opperent openely
APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-265 UNIT TWO DATEJnnue.rv 6 COMPLETED BYAlex Hisak TELEPHONE 309-654-2241xi94 MONTH December 1983 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1.
-1.7 17.
-1.5 2.
-i.6 18.
-2.0 3.
-1.5 19.
.-2.5 4.
-1.5 20.
-2.6 5.
-1.5 21.
-2.7 6.
-1.5 22.
-2.5 7.
-1. 5 _
23.
-1.8 B.
-i.5 24.
-i.6 9.
-i.6 25.
-1.8 10.
-1.8 26.
-i.9 11.
-2.0 27.
-2.0 i
12.
-1.2 28.
-2.7 13.
-1,6 29.
-2.2 14.
-1.5 30.
-2.3 15.
-1.3 31.
-2.6 16.
-1.4 INSTRUCTIONS On this fern, list the overage daily unit power level in fuit-Net for each day in the reporting nenth.Conpete to the neerest uhele ne tt.
These figeres vi 1 k esed to plot a graph f6r each reporting nenth. Note that when nasinen dependable capacity is used for the ut electrical.roting of the valt there may be occasiens when the daily eseroge peuer level excetes the ill! line (or the restrict 9 power level line),In sech cases,the neerage daily unit power outpet sheet sheeld be feetnoted to explain th< wrent onesely
M M
M M
FlTE EE9 M
M M
M M
M M
'M M
M
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ID/SA APPENDIX D QTP 300-S13 UNIT SilHTlXNNS AND POWER REDUCTIONS Revision 6 DOCKET NO. _050:25h Aaxust 1982 UNIT NAME _ Quad-Cities _ Unit _One COMPI.ETED RY JleL MISak._CKt 194 i
DATE
_JanuaryJ._l382L_
REPORT HONTil _DECE11BER_1983 TELErilONE 309-654-2241 m
8 e
sa sa sa Q
M LICENSEE R@
DURATION M
EVENT
$v gu w
g O
NO.
DATE (Il0URS)
REPORT No.
CORRECTIVE ACTIONS /CONNENTS o
ca
'h 83-87 831204 s
0.0 B
5 HA XXXXXX Reduced load to perforn weekly Turbine tests F
83-88 831210 s
0.0 B
5 RC CONROD Reduced load to adjust control rod pattern 83-89 831217 s
0.0 B
5 11A XXXXXX Reduced load to perform weekly Turbine tests 83-90 831217 F
0.0 A
5 llH PuttPXX Reduced load to u?tch Condensate pumps due to leakage from the inboard seal of I
the ID Condensate pump 1
1 83-91 831225 s
0.0 B
5 HA XXXXXX Reduced load to perform weekly Turbine 1
tests 83-92 831231 F
0.0 A
5 CB VALVEX neduced load for entering Drywell to investigate cause of increased leakage.
Leakace found to be from isolation valve packing on IB Recirculation pump discharge valve bonnet leak-off Iine.
4 x
APPROVED r
I' AUG 1 G l'J82
, (final) ygg3g 1
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O O
O O
M M
M M
M M
M M
M M
M O
We O
a.
ID/5A APPENDIX D QTP 300-S13 IINIT SillITDOWNS AND POWER REDilCTIONS Revision 6 DOCKET NO.
050-265 August 1982
__ uad-Cities Unit Two CONPl.ETED HY Alex Hisak, ext 194 UNIT NAME Q
DATE January 9, 1984 REPORT P10NTil DECEMBER 1983 TEl.EPil0NE 309-654-2241 5
m e
m x
20 y$
u
$m O.c O 3o M
I.ICENSEE-o mo n, o H
DURATION M
EVENT g
NO.
DATE (Il0URS)
"o REPORT No.
CORRECTIVE ACTIONS /COHilENTS u
o 83-66 830904 s
744.0 C
4 RC FUELXX Unit Two remains shutdown for End of Cycle Six Refueling and Maintenance 4
1 1
APPROVED l
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_ (final) ygg3g
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 for the reporting period.
B.
Control Rod Drive Scram Timing Data for Units One and Two There was no Control Rod Drive Scram Timing Data for the reporting period.
VII. REFUELING INFORMATION t
The following information about future reloads at Quad-Cities Station was requested in a January 26,19'i8, 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 Infomation",
dated January 18, 1978 l
l
~~Yh-ee.,
s-QTP 300-S32 R; vision I l-QUAD-CITIES REFUELit!G tiarch 1978 8
INFORMAT10f REQ' JEST s
{
L 1.
Unit:
Q1 Reload:
6 Cycle:
7 2.
Scheduled date for next refueling shutdown:
9-6-82 3
Scheduled date for restart following refueling:
12-18-82 4.
Will refueling or resumption of operation thereaf ter require a technicsl specification ange or other license amu dment:
Yes 5.'
Scheduled date(s) for submitting proposed licensing action and supporting
]
information:
8-19-82:
Tecn. Spec. changes submitted to the :2C.
iu 6.
Important IIcensing considerations associated with refueling, e.g., new or
' different fuel design or supplier, u,. reviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:
y a) All 7x7 fuel asse=blies vill be removed fr0= the core.
b) MAPI.HGR curves for fuel types in the core are being extended to LO,000 '."4D/S~'.
c) MCPR limits vill be detemined by GI's CD'ai ec=puter code.
d) The vessel pressure safety limit is being =cdified to accommodate the
~
potential for higher reactor pressures as calculated by CD'E!.
7 The number of fuel assemblies.
N a.
Number of assemblics in core:
724 J
b.
Number of assemblies in spent fuel pool:
1730 9
j 8.
The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is planned in number of fuel assemblies:
n 9
O a.
Licensed storage capacity for spent fuel:
3o57 b.
Planned increase in licensed storage:
O _,,
9 The projected date of the last refueling that can be discharged to the
~
spent fuel pool assuming the present licensed capacity: 2003
(
XPPROVED
'i-APR 2.01978 Q.c.o.S.R.
a
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QTP 300-S32 Rsvision 1 3
QUAD-CITIES REFUELING March 1978 I
L INFORMATION REQUEST L
1.
Unit:
Q2 Reload:
6 Cycle:
,7 2.
Schecuted date for next refueling shutdown:
9-5-83 3
Scheduled date for restart following refueling:
11-12-83 i
4.
Will refueling or resumption of operation thereaf ter require a technical specification change or other license amendment:
7 No, however, a change to the Technical Specifications is behq submitted
(
(see below).
5.
Scheduled date(s).for submitting proposed IIcensing action and supporting f
Information:
June 14, 1983 (Scheduled) 6.
Important licensing considerations associated with refueling, e.g., new or
' different fuel design or suppiler, unreviewed design or performance analysis
~
methods, significant changes in fuel design, new operating procedures:
p.
a)
All new fuel assemblies will be of barrier design; MAPLHGR curves will be re-labeled to include the barrier designation.
b)
The use of I.mproved assumptions in the load reject witacut bypass analysis resulted in a much improved MCPR operating limit.
Technical Specifications are being changed to provide this add'etional operating margin.
7.
The number of fuel assemblies.
a.
Number of assemblies in core:
724 p
b.
Number of assemblies in spent fuel pool:
412 8.
The present licensed spent fuel pool storage capacity and the size of any l
increase in licensed storage capacity that has been requested or is planned l
In number of fuel assemblies:
L a.
Licensed storage capacity for spent fuel:
3G97
(
0 b,
Planned increase in licensed storage:
9 The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2003
~
WPPROVED
- APR 2 01978 Q.C.C.S.R.
t o
VIII, CLOSSARY The following abbreviations which may have been used in the Monthly Report, are defined below:
Atmospheric Containment Atmospheric Dilution / Containment ACAD/ CAM Atmospheric Monitoring ANSI American National Standards Institute Average Power Range Monitor APRM Anticipated Transient Without Scram ATWS Bolling Water Reactor BWR Control Rod Drive CRD Electro-Hydraulle Control System EHC Faergency Operations Facility EOF Generating Stations Emergency Plan CS EP High-Ef ficiency Particulate Filter HEPA High Pressure Coolant Injection System HPCI High Radiation Sampling System HRSS Integrated Prima ry Containment Leak Rate Tes t IPCLRT Intermediate Range Monitor IRM ISI Inservice Inspection Licensee Event Report LER Local Leak Rate Test LLRT Low Pressire Coolant Injection Mode of RHRS LPCI Local Power Range Monitor LPRM Maximun Aver?ge Planar Linear Heat Cencration Rate MAPLHGR Minimum Critical Power Ratio MCPR MFLCPR Maximum Fraction Limiting Critical Power Ratio Maximum Permissible Concentration MPC Main Steam Isolation Valve MSIV National Institute for Occupational Safety and Health NIOSH PCI Primary Containment Isolation Precondition 1ng Interim Operating Management Recamnendations PCIOMR RBCCW Reactor Building Closed Cooling Water System Rod Block Monitor RBM Reactor Core Isolation Cooling System RCIC RHRS Residual Heat Removal System RPS Reactor Protection System Rod Worth Minimizer RRM Standby Cas Treatment System SBGTS Standby Liquid Control SBLC Shutdown Cooling Mode of RHRS SDC SDV Scram Discharge Volume Source Range Monitor SRM TBCCW Turbine Building Closed Cooling Water System Traversing Incore Probe TIP Technical Support Center TSC M
r O
4.
S, g Commonwe:lth Edison Quid Cities Nuclear Pow:r Station
) 22710 206 Av:nus North C
g Coraova, Illinois 61242 Telephore 309/654-2241 NJK-84-1 January 3, 1984 Director, Of fice of Inspection & Enforcenent United States Nuclear Regulatory Commission Washington, D. C.
20555 Attention: Document Control Desk Gentlemen:
Enclosed for your information is the Monthly Performance Report covering the operation of Quad-Cities Nuclear Power Station, Units one and Two, during the month of Decenber 1983.
Very truly yours, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION
^
/
/
//, ssdx/ ;
N. J, Ka11vlanakis Station Superintendent bb Enclosu re h,
Y
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