ML20215D207
| ML20215D207 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 11/30/1986 |
| From: | Kronich C, Robey R COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION (ADM), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| 0027H-0061Z, 27H-61Z, RAR-86-40, NUDOCS 8612160267 | |
| Download: ML20215D207 (22) | |
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QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT NOVEMBER, 1986 COMMONWEALTH EDISON COMPANY l
AND IONA-ILLIN0IS GAS & ELECTRIC COMPANY NRC DOCKET N05. 50-254 AND 50-265 l
LICENSE NOS. DPR-29 AND DPR-30 l
8612160267 861130 PDR ADOCK 05000254 R
ppg 0027H/00612 IEa9
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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.
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 V.
Data Tabulations A.
Operating Data Report B.
Average Daily Unit Power Level 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.
Glossary 0027H/0061Z
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it I.
INTRODUCTION Quad-Cities Nuclear 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 Commonwealth Edison Company and Iowa-Illinois Gas & Electric Company.
The Nuclear Steam Supply Systems are General Electric Company Bolling Water Reactors.
The Architect / Engineer was Sargent & Lundy, Incorporated, and the primary construction contractor was United Engineers & Constructors.
The Mississippi River is the condenser cooling water source.
The plant is subject to license
-numbers DPR-29 and DPR-30, issued October I, 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 Carol Kronich, telephone number 309-654-2241, extensions 2240 and 2157.
0027lI/0061Z
I g
II.
SUMMARY
OF OPERATING EXPERIENCE A.
Unit 1 November 1-14 Unit 1 began the month of November on Economic Generation Control (EGC).
At 0002 the unit was taken off of EGC until 0155. At 1400 the unit was taken off of EGC due to a Turbine bypass valve opening. The unit returned to EGC operation at 1445. At 2036 the unit was taken off of EGC and, at 2335 load was dropped to 700 MWe for Turbine surveillances.
On November 2, at 0345, load began an increase towards full power.
Full power was obtained at 1530. The unit returned to EGC operation on November 5, at 0750. The unit stayed on EGC until 1410, at which time an increase to full power began. At 2100 the unit was at 827 MWe and holding. The unit stayed at full power until November 7, at 1200, at which time it began a decrease due to a scheduled maintenance outage.
At 1875 the Turbine was manually tripped, and at 1930 the Reactor was manually scrammed and mode switch placed in SHUTDORN. The unit remained shutdown through November 14.
November 15-21 On November 15, at 1120, the mode switch was placed in STARTUP and at 2227 the Reactor went critical. At 2240 an Intermediate Range Monitor (IRM) problem developed; a control rod was inserted so the Reactor was not critical. On November 16, at 0023, the Reactor went critical and at 0026 the Reactor was not critical due to the IRM's.
At 0123 the Reactor went critical and, at 0200 the Reactor was not critical due to IRM's.
At 0708 the Reactor went critical and at 1314 the mode switch was placed in RUN.
At 1624 the Generator was placed on line at an initial load of 50 MWe.
The load was increased on through November 17, and at 1131 was at 625 MWe.
At 1515 an Unueual Event was declared due to the Torus not being completely inerted. At 1900 a unit shutdown commenced at 120 MWe/hr. At 2150 HPCI was declared out of service to repair valve MO 1-2301-5.
At 2335 load was holding at 145 MWe.
On November 18, at 0145, the Torus was inerted. At 0415 HPCI was declared operable and the Unusual Events were terminated.
The unit started increasing load at 0620, and on November 19, at 0700, was at 771 MWe.
At 2200 load was decreased 100 MWe/hr. until 600 MWe was obtained to do rod movements. On November 20, at 0045, unit load started increasing toward full power. The unit was at 833 MWe and holding at 0520 on November 21.
e 1
s A.
Unit 1 (cont.)
November 22-30 On November 22, at 1245, load was decreased to 775 MWe to change the Condensate Pumps combination. At 1322 load started increasing back to full power and at 1350 reached 831 MWe.
The unit stayed at full power until November 25, at 0000, at which time it decreased to 775 MWe to do Turbine surveillances. At 0055 load started increasing again until 0245 when 832 MWe was obtained. At 1210 load was decreased and placed on EGC. On November 26, at 0720, the unit was taken off of EGC to increase to full power. At 0735 the unit was at 833 MWe until 0925 then decreased for EGC operation. At 0940 the unit was on EGC. At 1320 the unit was taken off EGC to increase to full power. At 1410 the unit was at 833 MWe.
On November.27, at 1700, the unit load was decreased for EGC operation. At 1737 the unit was on EGC and remained there until November 30, at 1512, at which time it was taken off ECC.
At 1625 load was increased from 765 MWe to 800 MWe and remained there until 1640 at which time the unit went on EGC and remained on EGC for the rest of the month.
B.
Unit 2 Unit 2 remained shutdown for the entire month of November for the End of Cycle Eight Refueling and Maintenance Outage.
1*
s III.. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIEMNTS, AND SAFETY RELATED MAINTENANCE A.
Amendments to Facility License or Technical Specifications On August 15, 1986, the NRC issued Amendment numbers 96 and 92 to Facility Operating License numbers DPR-29 and DPR-30, for Quad-Cities Nuclear Power Station, Units 1 and 2.
These amendments change the Drywell concentration upper limit from 5 percent by weight to 4 percent by volume.
On October 28, 1986, the NRC issued Amendment number 97 to Facility Operating License number DPR-29, for Quad-Cities Nuclear Power Station, Unit 1.
This amendment revises the minimum critical power ratio (MCPR) limits contained in the Technical Specifications, and deletes MCPR limits for fuel types no longer in use.
B.
Facility or Procedure Changes Requiring NRC Approval There were no Facility or Procedure changes requiring NRC approval for the reporting period.
C.
Tests and Experiments Requiring NRC Approval l
There were no Tests or Experiments requiring NRC approval 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 Units 1 and 2 during the reporting period. This summary includes the following: Work Request Numbers, Licensee Event Report Numbers, Components, Cause of Malfunctions, Results and Effects on Safe Operation, and Action Taken to Prevent Repetition.
UNIT 1
MAINTENANCE SLANARY CAUSE RESULTS & EFFECTS W.R.
LER OF ON ACTION TAKEN TO NUEER NUWER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q46912 86-002 Rebuilt MSIV l-Inspection of the The unit was in the The leaking MS1V's were 203-2A internals of valve SHUTDOWN mode. Leakages rebuilt and the Station determined the of the A, B, i C steam has purchased a new valve cause of leakage lines were minimal due seat lapping machine.
was due to grime to the fact that the and pitting on outboard valves, 1-203-the valve seat.
2A, 2B, & 2C were shown to have the majority of the leakage. Therefore, safety implications were minimal.
Q52695 1-1705-2B A defective femto The
'B' Steam Line Rad The defective monitor was Replaced ammeter board in Monitor was in the replaced and the original Chassis on the Main Steam tripped condition until was sent to General Main Steam Line Radiation it was replaced & the Electric for repair. The Line Radia-Log-Rad Monitor other three steam line monitor was later rein-tion Monitor chassis was monitors were functional, stalled and functionally determined to be so safety implications tested.
the cause of the were minimal.
1/2 scram & Group I Isoletion.
Q53107 Disassembled There was no cause Safety implications were The 250 Volt DC Battery is and moved new of malfunction as minimal as the load from being replaced per Modifi-250 Volt this was a planned the Unit 1 Battery was cation 4-1-85-43, there-Battery course of action.
transferred to the Unit fore, no action is needed 2 Battery and Unit 1 to prevent repetition, was shutdown for a scheduled outage on 11-7-86.
0027H/0061Z
. UNIT
?.
MAINTENANCE
SUMMARY
CAUSE RESULTS & EFFECTS W.R.
LER OF ON ACTION TAKEN TO NUEER NUEER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q52492 Recalibrated The cause of the The safety implications A Work Request was written Transmitter, Level Recorder were minimal due to the to correct the problem.
Checked Sight-malfunction was availability of the The Level Transmitter was-glass & Cali-determined to be suppression chamber water recalibrated and tested.
bration; Checked setpoint drift.
level sightglass & a LR 2-1602-7 redundant narrow range suppression chamber level indicator.
Q52677 86-15 Installed New The cause of the The unit was in the SHUT-The Drywell Head gasket was Gasket on Dry-ILRT to be in DOWN mode with values replaced with Garlock #8364 well Head excess of Tech-well below that which seal material.
CECO is also nical Specifi-would cause serious impact doing research to determine cations was to public safety in an an even better material for determined to accident and only slightly this application.
be a poorly above the allowable.
sealed Drywell Head gasket.
i 0027H/0061Z
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 1 Licensee Event Report Number Date Title of Occurrence 86-31 11-8-86 Unit 1 Diesel Generator Auto Start 86-32 11-8-86 Unit 1/2 Diesel Genera-tor Output Breaker Auto Trip 86-33 11-12-86 Control Room Panels --
Inadequate Mounting 86-34 11-17-86 HPCI Inoperable --
1-2301-5 Valve Leaking 86-35 11-26-86 RCIC Inoperable Due to Speed Meter Bulb Unit 2 Licensee Event Report Number Date Title of Occurrence 86-17 11-5-86 Recirc 'K' Riser Pipe Crack 86-18 11-13-86 Standby Gas Treatment Auto Start -- RBV Auto Trip X-raying 86-19 11-28-86 One-fourth Scram with Rod Insertion 0027H/0061Z
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0 V.
DATA TABULATIONS The folloting data tabulations are presented in this report:
A.
Operating Data Report B.
Average Daily Unit Power Level C.
Unit Shutdowns and Power Reductions 0027H/00612
?-
o OPERATING. DATA REPORT DOCKET NO.
50-254 UNIT ONE DATEDECEMBER 8 1986 COMPLETED BYCAROL L KRONICH TELEPHONED 09) 654-2241 OPERATING STATUS 0000 110186
- 1. Reporting period 2400 113086 Gross hours in reporting period:
720 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 Cunulative 5.
Number of hours reactor was critical 517.1 5490.8 102152,2 6.
Reactor reserve shutdown hours 0,0 0.0 3421.9 7.
Hours generator on line 506,0 5309.0 90669,3 8.
Unit reserve shutdown hours.
0.0 0.0 909,2 9.
Gross thernal energy generated (MWH) 1160304 12596591 207561740
- 10. Gross electrical energy generated (MWH) 305611 4137980 67257351
- 11. Net electrical energy generated (MWH) 360016 3946047 62973390
- 12. Reactor service factor 71.0 60,5 00,0
- 13. Reactor ovallobility factor 71,8 60.5 02.7 L4. Unit service factor 70,3 67.2 77,3 1
- 15. Unit avo11obility factor 70.3 67.2 70,0
- 16. Unit capacity factor (Using MDC) 66,5 64.0 64,2
- 17. Unit capacity factor (Using Des.MWe) 64,8 62.4 62.5
- 10. Unit forced outage rate 0.0 1.7 5.7
- 19. Shutdowns scheduled over next 6 months (Type,Date,and Duration of each):
- 20. If shutdown at end of report period,estinated date of startup ___[_^_____,___
\\
0 OPERATING DATA REPORT DOCKET NO.
50-265 UNIT TWO DATEDECEMBER 8 1986 COMPLETED BYCAROL L KRONICH TELEPHONE (309) 654-2241 OPERATING STATUS 0000 110186
- 1. Reporting period:2400 113086 Gross hours in reporting period 720
- 2. Currently authorized power level (MWt): 2511 Max, Depend copocity (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 Cunulative
- 5. Number of hours reactor was critical 0,0 6448.0 97715,7
- 6. Reactor reserve shutdown hours 0.0 0,0 2985,0
- 7. Hours generator on line 0.0 6401.5 94699.1 8.
Unit reserve shutdown hours.
0,0 0,0 702.9
- 9. Gross thernal energy generated (MWH) 0 15230676 201339699
- 10. Gross electrical energy generated (MWH) 0 4942510 64359292
- 11. Het electrical energy generated (MWH)
-2373 4725590 60599849
- 12. Reactor service factor 0,0 80,4 77,1
- 13. Reactor avo11obility factor 0,0 00,4 79,5
- 14. Unit service factor 0, 0_
79,9 74,7
- 15. Unit avo11ob111ty factor 0.0 79,9 75,3 L6. Unit copocity factor (Using MDC)
,4 76,7 62,2 L7. Unit capacity factor (Using Des.MWe)
,4 74.7 60,6 LO. Unit forced outage rate 0,0
,7 7,7
- 19. Shutdowns scheduled over next 6 nonths (Type, Dote,ond Duration of each):
- 20. If shutdown at end of report period,cstinated dote o f c t o r t u p
__I _18 _8{_____
SWWFICIM. CSNIN IIANIS flIE 'NETTiflil5TEPiN t
T-APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-254 UNIT ONE DATEDECEMBER 8 1986 COMPLETED BYCAROL L KRONICH TELEPHONE (309) 654-2241 MONTH November 1986 DAY' AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1.
729,8 17.
527.0 2.
767.0 18, 463.8 3.
765.9 19.
700.2 4,
762.4 20, 715.0
'5.
777.7 21.
798.2 6.
787.0 22, 793.7 7.
493.3 23.
797.3 8.
-i.5 24.
793.1 i
9.
.i 25.
793.3 LO.
3 26.
785.6 11.
4 27, 769,2 12.
.7 28.
763.0 I
13.
-6.6 29.
769,9 14.
-6.0 30.
735.1 L5.
-9.i 16, 65.4 INSTRUCTIONS On this fern, list the everage daily snit powr level in Mit-Net for each day in the reporting nonth.Cenpete to the neerest whole mejouett.
These figeres will be esed to plot a graph for eeth repertina nanth. Note that whtn notinen dependable capacity is esed for the net tiettrical rating of the snit there not be %ctesiens when the daily cetragt power level esteeds the 100% line (or the restricted pow r level line).,In seth costs,the cetragt delly enit po wr cetpet sheet sheeld be feet;eted to tiplein the opperent onenoly
8 UNOFFICIAL COMPMY NUMBERS ME USED 1D THIS REPORT
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APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-265 UNIT TWO DATEDECEMBER 8 1986 COMPLETED BYCAROL L KRONICH TELEPHONE (309) 654-224_i MONTH November 1986 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1.
-5.8 17.
-2.8 2.
-6.0 18.
-2,8 3.
-5.9 19.
-2.7 4.
-2.9 20.
-2.8 5.
-3.8 21.
-2.7 6.
-3.6 22.
-2.5 7.
-3.5 23.
-2.6 8.
-6.3 24.
-2,7 9.
-6.7 25.
-2.7
-2.7 10.
-6.6
- 26.._
11.
-6.7 27.
-2.7 12.
-6.4 28.
-2.9 13.
-3.5 29.
-2.7 14.
-3.5 30.
-2.6 15.
-3.3 16.
-3,0 INSTRUCTIONS On this fern, list the average daily unit power level in MWe-Net for each day in the reperting ninth. Compute to the nearest whole negawatt.
These figeres will be ssed to plot a graph for toch reporting ninth. Note that when narinen dependable capacity is used for the net electrical rating of the init,there nay be occasions when the daily average power level exceeds the 100% line (or the restricted power level line).In such cases,the average daily snit power outpet sheet shevid be festnoted to explain the apparent onenaly
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ID/5A APPENDIX D QTP 300-S13 UNIT SIIUTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO.
050-254 August 1982 UNIT NAME Quad-Cities Unit 1 COMPLETED BY C Kronich DATE December 9, 1986 REPORT MONTil NOVEMBER 1986 TELEP110NE 309-654-2241 s
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EVENT
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DATE (IIOURS)
REPORT NO.
CORRECTIVE ACTIONS / COMMENTS l
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86-34 861101 S
0.0 B
5 HA TURBIN Reduced load to 700 MWe for Turbine surveillances l
86-35 861107 s
213.0 B
2 ZZ ZZZZZZ Unit shutdown for scheduled Maintenance Outage 86-36 861117 F
0.0 H
5 86-34 ZZ ZZZZZZ Commenced unit shutdown due to Unusual Event -- oxygen concentration greater than 4% in the Torus. Then llPCI was taken out of service to repair MO 1-2301-5 valve (shutdown terminated 1
approximately 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />).
APPROVED AUG 101982
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M,f ID/5A APPENDIX D QTP 300-S13 UNIT SIIUTDOWNS AND POWER REDUCTIONS Revision 6 DOCKET NO.
050-265 August 1982 UNIT NAME Quad-Cities Unit 2 COMPLETED BY C. Kronich DATE December 9, 1986 REPORT MONTil NOVEMBER 1986 TELEP110NE 309-654-2241 5
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REPORT NO.
CORRECTIVE ACTIONS / COMMENTS NO.
DATE (110URS) o A
86-44 861011 S
720.0 C
2 RC FUELXX Unit shutdown for End of Cycle Eight Refueling and Maintenance Outage APPROVED AUG 1 G 1982 i (final) ygg3g
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VI.
UNIQUE REPORTING REQUIREMENTS The following items are included in this report based on prior commitments to the commission:
A.
Main Steam Relief Valve Operations Relief valve operations during the reporting period are summarized in the-following table.
The table includes information as to which relief valve was actuated, how it was actuated, and the circumstances resulting in its actuation.
Unit:
1 Date: November 7, 1986 Valves Actuated No. & Type of Actuation 2-203-3A 1 Manual 2-203-3B 1 Manual 2-203-3C 1 Manual 2-203-3D 1 Manual i
2-203-3E 1 Manual Plant Conditions: Reactor Pressure - 918 psig Description of Events:
Surveillance, Technical Specification 4.5.D.l.a Unit:
1 Date: November 16, 1986
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Valve Actuated:
2-203-3E No. & Type of Actuation:
1 Manual Flant Conditions: Reactor Pressure - 923 psig Description of Events:
Surveillance, Technical Specification 4.5.D.l.a.
(Post Maintenance repaired ll4E Relay)
B.
Control Rod Drive Scram Timing Data for Units 1 and 2 There was no Control Rod Drive scram timing data for Units 1 and 2 for the reporting period.
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VII.
REFUELING INFORMATION The following information about future reloads at Quad-Cities Station was
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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.
0027H/0061Z
QTP 300-532 Revision 1 QUAD-CITIES REFUELING Narch 1978 INFORMATION REQUEST 1.
Unit:
Q1 Reload:
8 Cycle:
9 2.
Scheduled date for next refueling shutdown:
9-14-87 3
Scheduled date for restart following refueling:
12-21-87 4.
Will refueling or resumption of operation thereaf ter require a technical specification change or other license amendment:
NOT AS YET DETERMINED.
5 Scheduled date(s) for submitting proposed licensing action and supporting information:
AUGUST 21, 1987 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:
NONE PLANNED AT PRESENT TIME.
7 The number of fuel assemblies.
a.
Number of assemblies in core:
724 b.
Number of assemblies in spent fuel pool:
1683 l
8.
The present IIcensed spent fuel pool storage espacity and the size of any increase in licensed storage capacity that hss been requested or is planned I
in number of fuel assemblies:
a.
Licensed storage capacity for spent fuel:
3657 b.
Planned increase in licensed storage:
0 9
The projected date of the last refueling that can be discharged to the l
spent fuel poc1 assuming the present IIcensed capacity: 2003 XPPROVED 1-APR 2 01978 Q.c.o.S.R.
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QTP 300-532 Revision i QUAD-CITIES REFUELING March 1978 INFORMATION REQUEST 1.
Unit:
Q2 Reload:
7 Cycle:
8 2.
Scheduled date for next refueling shutdown:
3-5-88 3
Scheduled date for restart following refueling:
1_Ig_87 4.
Will refueling or resumption of operation thereafter require a technical specification change or other license amendment: YES -
TECHNICAL SPECIFICATION CHANGES WILL BE REQUIRED FOR NEW FUEL TYPES (MAPLHGR CURVES),
MCPR OPERATING LIMIT, AND A LICENSE AMENDMENT TO MOVE SINGLE LOOP OPERATION INTO 5
I5Ee@uD YajC,y,y,I "S,.r s bmitting proposed IIcensing action and supporting information:
SUBMITTED SEPTEMBER 18, 1986.
6.
Important Ilcensing considerations associated with refueling, e.g., new or
' different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:
NONE AT PRESENT TIME.
7 The number of fuel assemblies, a.
Number of assemblies in core:
0 b.
Number of assemblies in spent fuel pool:
1922 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:
a.
Licensed storage capacity for spent fuel:
3897 b.
Planned increase in licensed storage:
0 9
The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2003 4PPROVED APR 2 01978 C2. C:. c). !5. Ft.
..o O
VIII. GLOSSARY The following abbreviations which may have been used in the Monthly Report, are defined below:
ACAD/ CAM -
Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring ANSI American National Standards Institute APRM Average Power Range Monitor ATHS Anticipated Transient Without Scram BWR Bolling Water Reactor CRD Control Rod Drive EHC Electro-Hydraulic Control System EOF Emergency Operations Facility GSEP Generating Stations Emergency Plan HEPA High-Efficiency Particulate Filter HPCI High Pressure Coolant Injection System HRSS High Radiation Sampling System IPCLRT Integrated Primary Containment Leak Rate Test IRM Intermediate Range Monitor ISI Inservice Inspection LER Licensee Event Report LLRT Local Leak Rate Test LPCI Low Pressure Coolant Injection Mode of RHRS LPRM Local Power Range Monitor MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum CriticP1 Power Ratio MFLCPR Maximum Fraction Limiting Critical Power Ratio MPC Maximum Permissible Concentration MSIV Main Steam Isolation Valve NIOSH National Institute for Occupational Safety and Health PCI Primary Containment Isolation PCIOMR Preconditioning Interim Operating Management Recommendations RBCCH Reactor Building Closed Cooling Water System RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling System RHRS Residual Heat Removal System RPS -
Reactor Protection System RHM Rod Horth Minimizer SBGTS Standby Gas Treatment System SBLC Standby Liquid Control SDC Shutdown Cooling Mode of RHRS SDV Scram Discharge Volume SRM Source Range Monitor TBCCH Turbine Building Closed Cooling Water System TIP Traversing Incore Probe TSC Technical Support Center 0027H/0061Z
D/
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Commonwealth Edison Quad Cities Nuclear Power Station 22710 206 Avenue North Cordova, Illinois 61242 Telephone 309/654-2241 RAR-86-40 December 1, 1986 Director, Office of Inspection & Enforcement United States Nuclear Regulatory Commission Washington, D. C.
20555 Attention: Document Control Desk 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 November, 1986.
Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION f..
R. A. Robef Services Superintendent bb Enclosure 4
{V I
0027H/00612
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