ML20067B138
| ML20067B138 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/30/1990 |
| From: | Robey R COMMONWEALTH EDISON CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| RAR-90-55, NUDOCS 9102080146 | |
| Download: ML20067B138 (22) | |
Text
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/ Coraova liiinois 61242 9740 L/
Tele: bone XM54 2241 RAR-90-55 July 2, 1990 Director of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Hall Station PI-137 Washington, D. C.
20555 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 June, 1990, Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION O O.
c-h.
R. A. Robey Technical Superintendent RAR/LFD/r;h Enclosure n
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0027H/C061Z
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QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT June, 1990 COMMONHEALTH EDISON COMPANY AND IONA-ILLINOIS GAS & ELECTRIC COMPANY NRC DOCKET NOS. 50-254 AND 50-265 LICENSE NOS. OPR-29 AND DPR-30 0027H/00612
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/00612
I.
INTRODUCTION Quad-Cities Nuclear Power Station is composed of two Bolling Hater 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 nuibers 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 Lynne Deelsnyder and Verna Koselka, telephone number 309-654-2241, extensions 2185 and 2240.
0027H/00612
a II.
SUMMARY
OF OPERATING EXPERIENCE A.
Unit One Unit One began the month of June operating at full power.
Normal operational activities occurred and routine surveillances were performed throughout the month. The unit remained near full power or operated in Economic Generation Control (EGC) per the demands of the Chicago Load Dispatcher.
Power levels were adjusted accordingly.
On June 15, during turbine weekly testing, the 'A' master trip solenoid valve light failed to extinguish when the NSO attempted to test the 'A' side. The 'B' side was not tested. Weekly testing was terminated, and generator load was reduced to less than 40% to bypass the turbine trip reactor scram while troubleshooting was performed.
The 'A' main steam isolation valve was then retested and it worked properly.
It was exercised several times without a malfunction.- Power levels were increased to full load at 1040 hours0.012 days <br />0.289 hours <br />0.00172 weeks <br />3.9572e-4 months <br />.
On June 17, a load reduction to 350 MWe was taken to perform cotc rol rod drive hot scram timing.
The surveillance was successfully completed and the unit was taken to full power.
B.
Unit Two Unit Two began the month of June operating at full power.
Normal operational.
activities occurred and routine surveillances were performed.
On June 18, power levels were reduced to identify a minor leak which was discovered in the main steam isolation valve room on t..e safe shutdown makeup pump discharge line. A weekend outage was scheduled for the following weekend.
Power IcVels were increased to full load.
On June _23, at 0103 hours0.00119 days <br />0.0286 hours <br />1.703042e-4 weeks <br />3.91915e-5 months <br />, the main generator was taken off line and at 0108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br />, the reactor was scrammed.
The pteviously identified leak on the safe shutdown makeup pump discharge line was repaired and a recircu-lation pump seal was replaced. On June 24, at 2328 hours0.0269 days <br />0.647 hours <br />0.00385 weeks <br />8.85804e-4 months <br />, the reactor
-was made critical, and at 1006 h7urs, on June 25, the generator was synchronized to the grid.
On June 25, at 1649 hours0.0191 days <br />0.458 hours <br />0.00273 weeks <br />6.274445e-4 months <br />, hydrogen tion testing was begun. The testing l
continued throughout the remainder 01
,e month. Noreal operational activites continued through the end of June.
1-III.
PLANT OR PROCEDURE CHANGES. TESTS, EXPERIMENTS, AND SAFETY EELATEDMAINTENANCE A.
Amendr'en,tn to Facility License or Technical Specifications Technical Specification Amendment Nos. 124 and 121 were issued on May 23, 1990 to Facility Operating License DPR-29 a.nd DPF-30. These an endment s revise the Technical Specificacions to modify the requirements for jet pump flow indication.
B.
Facili;ty or Procedure Changes Requiring NRC Approval There were no Facility or Procedure changes requiring NRC approval for the reporting period.
C.
Tests and Exyeriments Requiring NRC Approval r
There were no Tests or Experiments requiring NRC approval for the reporting period.
b.
Correctivo Maintenance of Safety Related Equipment The following represents a tabular sumraary of the major afiety related maintenance performed on Units One and Two during the reporting period.
This surmaary includes Work Request numbers, system component description, and work performed.
y --
0027H/00612 l
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UNIT 1 MAINTENANCE
SUMMARY
WORK REQUEST SYSTEM EID DESCRIPTION WORK _ PERFORMED Q85400 7510 Valve SBGT "A" Suction Damper' As foui.d <_ondition: nicks on seal cup.
As left, condition:
to Fan rebuilt' operator, replaced softwear..
Q52049 5741 Damper Control Room.AFU "A"
. Readjust i,utlet damper to provide approximately 2,000 SCFM Booster Fan
' in order to clear alarm.
Replaced actuator for' damper
. with exception of open travel..stop' actuator to be rebuilt in WP. Q76621.
Q52027 5741
' Damper Control Room AFU "B"
'.ald tag 88-769.
Rebuild under Q74754. Replaced operator Booster Fan with new contromatic' operator..
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UNIT 2 MAINTENANCE
SUMMARY
i WGRK REQUEST
. SYSTEM-EID DESCRIPTION ~
WORK PERFORMED
.Q85424 0203 Valve Electromatic Relief' Replaced leaking electromatic.
Q85495~
0203 Valve Electromatic Relief Welded nipple inta flange of electrometic.
~
Q84988
-1001 Pump Residual Heat Removal Service ' Replaced'l/4" close nipple and.1/4-3/8 bushing.
Water Q85321 1201 Limitorque Reactor Water Cleanup
'Found open set at 2 1/2, changed to 2 3/4.'
Found close Recire Pump Suction Valve set at 2.1/2, changed to 2 3/4.
Q84452 1600 Hatch North Torus Replaced gaskets.and lubricated.
Q85160 2301 ~
Restricting Orifice HPCI Turbine Disassembled flanges, cleaned and inspected mating Stop Valve surfaces on flanges and orifice plate, installed gaskets and bolting.
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IV.
LICENSEF EVENT REPORTS l
-.The fol owing is a tabular summary of all licensee event reports-for. Quad-
. Cities Units One and Two occurring during one reporting period, pursuant to the reportable occurrence reporting reqb".rementa-as set forth in sections-6.6.B.1. and 6.6.B.2. of the Technical Specifications.
UNIT 1 i
Licensee Event Report-Number-Date Title of Occurrence 90-11 6/11/90 1/2 'A' Fire Diesel
-00S-Longer Than_7 Days 90-12 6/12/90 Control Room (CR) 'B'
- Train HVAC Inoperable; 90-13 6/26/90 Lightning Strike Causing Valve 1-220-45 to Close UNIT 2 i
90-08 6/2/90~
HPCI Flow Controller.
Failure u
1
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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.
Uiilt Shutdowns and Power Reductions 00;IH/0061Z
t
.L APPEND!!C 0FERAllN6 DATA REPORT DocketNo. 50-254 Unit One 3
-Late -July 3, 1990 Completed By Lynne Deelsnyder Telephone 309-654 2241
.0PERAllN6 STATUS.
0000-060190
- l. Reporting Period 2400 063090-GrossHoursinReportPeriod: M 2.'- Cur' ently Authorized Power level (MWtit Mll Max. Depend. C;ipacity (MWe hetle M hugn Electrical Rating (Ne Net): M
- 3. Power Level to Which Restricted (11 Any) (MWe-Net): M 4.- Reasons for Restriction (If any):-
THIS MONTH-YR10DATE
- UMULATIVE
~
' 5.- Nuaber of Hours Reactor Was Crl'ical 720.0-4228.6 128392.4 6.--Reactor. Reserve Shutdown Hours 0.0 0.0 3421.9
- 7. Hours 6enerator On'Line 720.0 4204.1 124294.7'
'8.
Unit Reserve Shutdown Hours
- 0.0 -
0.0 909.2 9.' ' Gross 'Thereal Energy 6enerated (MWh)--.
1665540.0.
9919219.0 265601580.0
- 10. Gross Electrical Energy Generated (M h)-
335140.0 3241709.0 86093266.0 II.~ Net Electrical-Energy 6enerated (Muhl 515237.0 3111131.0-80951517.0_
=
--12. Reactor Service Factor' 100.0 97.4 80.4
- 13. Reactor Availability Factor.
J00.0=
97.4-82.5 i
- 14.~ Unit Service Factor
-100.0 96.8 77.8 115. Unit ~AvailabilityFactor 100.0 96.8 78.4
- 16. Unit. Capacity Factor (Using MDC) 93.1 93.2 65.9
- 17. Unit. Capacity Factor (using Design MWe) 90.7
- 90.8 64.2
- 18. Unit Forced Outage Rate-0.0 1.9
- 5. 3 --
- 19. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):
[
-20. If Shut Doan at End of Report Period, Estisated Date of Startup:
21.UnitsinTestStatus(PriortoCoseercialOperation):
Forecast Achieved initial Criticality-
' Initial Ei;;.tricity
~Coesercial Operation 1.16-9
.~
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APPEND!! C OPERAi!N6 DATA REPORT-Docket No. 50-265 Unit'-Two Date July 3; M '
Completed By~ Lynne Deelsnyder-Telephone 309-654 2241 OPERATIN6 STATUS 0000 060190
- 1. Reporting Period 2400 063090 Gross Hors in Report Feriods 720 3
- 2. Currently Authorized Power Level (MWt): Ull Mar. Depend. Capacity.(Mue Net): 7J Design Electrical Rating (MWe-Neth lf9
'31 Pomer' Level to Nhich Restricted (If Any) (Me-Net): N/A 4.
Reasons for Restriction (If any): -
i
-l 2
THIS MONTH YRTODATE
- CUMULAllVE 5.-
Nueber of Hours Reactor Was Critical 673.7 2063.2 121447.0.
- 6. ~Reartor Reserve Shutdown Hours 0.0 0.0' 2985.8 9
7... Hours Generator On Line-663.0 2015.4
!!8113.9
~8..bnit Reserve Shutdown Hours -
_-0.0
- 0. 0 702.9
=)
- 9. 6ross Thereal Energy 6enrated (Mh) 1533276.0 4304798.0 253701415.0 10.=6ross E!ectrical Energy Generated (Mh) 499804.0; 1404211.0 81343295.0
~
.11.NetElectricalEnergyGenerated(Wh) 480500.0-1327162.0
- 76806811.0 93.6 47.5 76.8 -
-12c Reactor Service Factor
- 13. Reactor Availability Factor 93.6 47.5 78.7
'14.'UnitServiceFactor.
92.1 -
46.4 74.7-
!!. Unit Availability Factor 92.1-46.4
_75.1 j
.16. Unit Capacity Factor tusing MDC){
86.8 ~
= 39.7 -
63.2 1 l17. Unit Capacity Factor (Using Design Mel
.84.6 38.7' 61.6 y
- 18. Unit Forced Outage Rate.
- 0. o.
0.0 8.0 l
219. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each): '
20..li Shut Down at End of Report Period, Estiaated Date of Startups.
n 2ti tats in Test Status--iPrior to Cossercial Operation):
Forecast Achieved InitialCriticality initial Electricity Coseercial Operation 1.16-9
..m._.
. ~. - -. _.. _. _ _... _. _ _ _. = _.
4.
- e.-.
E APPENDIX $=-
l AVERA6EDAILYUNITPOWERLEVEL f
Docket No.- 50-254 UnitL One Date hly 3,1990 Coacleted By LynneDeelsnyder Telephone 3096542241 MONTH
-. {UK _
DAYAVERASEDAILYPOWERLEVEL DAYAVERASEDAILYPOWEKLEVEL
[
(Me-het)
(NWe-Net)
I i-725 17 586
-2 725' il 777
~
3 693-19
- 767
- 643 20 -
- 701 656.
- 21 696 6
727.-
22 728 17 732.
23 741 0:
~735~
24-646 9
724.
25 738
= 10-639:
26 765
' 692 27
- 769 760
. 28 695-1 13:
765-,
29 1753 I
i 14 773-
- 30 760' 15 -
584 l
.16 775
- INSTRUCTIONS On this fore, list the average dally unit' power level in MWe-Net for each day in the reporting conth.
4
- Cospute to the nearest whole negawatt.-
These. figures will_be used to plot a graph for each repcrting conth. Note that when satieue dependable, capacity is used for the net electrical; rating of thk ur.it, thtre say be occasions.when the daily average
- poser level exceeds the 1001 line -tor the restricted power level line). In such cases, the average daily-unit power output sheet should be footnoted to explain the apparent anosaly.
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APPENDil9 AVERA61DAILYUNITPOWERLEVEL DocketNo. 50-265 f
Unit two 1
Date July 3. 1990 i
CospletedBy Lynneteclonyder Telephone 309654224l I
MONTH
[M DAYAV!R46EDAILYPOWERLtVEL-DAYAVERAGEDAILYPOWERLEVEL (Me Nett (He Net) l 709 17 789 2
788 16; 678 3-778 19 646 4
7BB 20 709 5
789 21 689 6'
790 22 633 4
7 789-23
+1 8-789 24
-6 I
9 789 25 99 10 755 26 231 11-
~764-27 623-12 785 28 787 13 781 29 766 14=
792 30 776 i
15 166 16 790-1 q
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-INSTRUCT!DNS-
\\
On' this fore,-list the average daily unit power level.in Me Net for each day-in the reporting month.
Cospute to the nearest whole negawatt.-
These figures will be used to plot:a graph for each reporting month. Note that when manieue dependable capacity is used for the net electrical rating of the unit, there say be occasions when the daily average power level exceeds the 1001 line for the restricted power level line). In such cases, the average daily-unit power output sheet should be footnoted to explain the apparent anomaly.
1.16-0
~
APPENDIX'D UNIT SNUTDOWNS AND POWER REDUCTIONS DOCKET NO.
50-254
' UNIT NAME Unit One COMPLETED BY Lynne F. Deelsnyder DATE July 3, 1990 REPORT MONTH June. 1990 TELEPHONE 309-654-2241 sc DoO h
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E es -
LICENSEE-e@
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u DURATICN iii!
E EVENT o
u NO.
DATE
'(HOURS)
REPORT NO.
CORRECTIVE ACTIONS /COffENTS 90-5 900615 F-0.0 B
5 Power Reduction Taken to Bypass Turbine Trip Reactor Scram to Troubleshoot
'A' MTSV Light 90-6 900617 F
0.0 B
5
. CRDRVE Power Reduction Taken to Perform Control Rod Drive Hot Scram Timing 1 (final)
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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 a;6uation.
Unit! Two Date: June 25, 1990 Valves Actuated No. & Type of Actuation 2-203-3B 1 Manual 2-203-3D
'l Hanual Plant Conditions:
Reactor Pressure - 930.7 Description of Events:
Post Maintenance Testing, Manual Operation of Electromatic Relief Valves (QOS 201-SI).
Tech Spect -Ref. 3.5/4.5.D.I.a B.
Control Rod Drive Scram Timing Data for Units One and Two The basis for reporting this data to the Nuclear Regulatory Commission are specified in-the surveillance requirements of Technical Specifications 4.3.C.'l and 4.3.C.2.
The following table is a complete summary of Units one and Two Control' Rod Drive Scram Timing for the reporting period. All scram timing was performed with Reactor pressure greater than 800 PSIG.
[
o
RESULTS_OF SCRAM TICING MEASUREMENTS.
PERF0FWED ON UNIT 162 CONTROL' ROD DRIVES, FROM 1-1-90 TO 12-31-90 AVERAGE TIME. IN SEC0f0S AT %
MAX. TIME INSERTED FROM FULLY WITHDRAWN FOR 90%
INSERTION DESCRIPTION NIABER 5
20 50 90 Technical Specification 3.3.C.1 &
DATE OF RODS 0.375 0.900 2.00 3.5 7 sec.
3.3.C.2 (Average Scram Insertion Time) 1-22-90 1
0.27 0.65 1.42' 2.52
.N-10 Unit 1, Hot Scram Timing for test due (2.52) to 127 Valve Diaphragm Replacement 1-31-90 1
0.29 0.64 1.34 2.36 R-7 Unit 1, Hot Scram Timing for test due (2.36) to 127 Valve Diaphragm Replacement 5-09-90 177 0.30 0.68 1.45 2.54 H-ll Unit 2, Hot Scram Timing during Start Up (3.29)
Sequence A&B, Cycle 11 6-21-90 88 0.29 0.66 1.40 2.46-M-11 Unit 1. Hot Scram Timing for Sequence B (2.70)
Cycle 11, 1st Sequence of Cycle 0027H/0061Z
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a VII.
REFUELING INFORMATION The following information about future reloads at Quad-Cities Station was requested in a January 26, 1978, licensing memorandum (78-24) from D. E.
O'Brien to C. Reed, et al., titled "Dresden, Quad-Cities, and Zion Station--NRC Request for Refueling Information", dated January 18, 1978.
0027H/00612
QTP 300-$32 Revision 2 QUAD CITIES REFUELING October 1989 INFORMATION REQUEST 1.
Unit:
01 Reload:
10 Cycle: _
11 2.
Scheduled date for n6xt refueling shutdown:
10-27-90 3.
Scheduled date for restart following refueling:
1-4-91 4.
Will refueling or resumption of operation thereafter require a Technical Specification change or other 11 cense amendment:
NONE AS YET DETERMINED.
5.
Scheduled date(s) for submitting proposed licensing action and supporting information:
JULY 6, 1990 6.
Important licensing considerations assoc!ated 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:
724 b.
Number of assemblies in spent fuel pool:
1537 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:
3657 b.
Planned increase in licensed storage:
n L
9.
The projected date of the.last refueling that can be discharged to the i
spent fuel pool assuming the present licensed capacity:
2008 APPROVED 14/0395t
-I-OCT 3 0 f969 O.C.O.S.R.
QTP 300-532 Revision 2 QUAD CITIES REFUELING October 1989 INFORMATION REQUEST 1.
Unit:
02 Reload:
10 Cycle:
11 2.
Scheduled date for next refueling shutdown:
9-7-91 3.
Scheduled date for restart following refueling:
12-9-91 4.
Will refueling or resumption of operation thereafter require a Technical Specification change or other license amendment:
NOT AS YET DETERMINED.
S.
Scheduled date(s) for submitting proposed licensing action and supporting information:
NOT AS YET DETERMINED.
6.
Important licensing 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.
4.
Number of assemblies in core:
724 b.
Number of assemblies in spent fuel pool:
2011 8.
The present licensed spent fuel pool storage capacity and the size of any increase in licenseo 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: 2008 APPROVED m nal) 14/0395t GCI 3 01969 O.C.O.S.R.
VIII. GLOSSARY Th'e 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 APRM'
. Average Power Range Monitor Anticipated Transient Without Scram ATHS Boiling Hater Reactor
-BHR
-CRD' Control Rod Drive Electro-Hydraulic Control System EHC_
EOF Emergency Operations Facility GSEP-Generating Stations Emergency Plan
-HEPA High-Efficiency Particulate filter HPCI High Pressure Coolant Injection System
-High Radiation Sampling System HRSS.
-Integrated Primary Containment Leak Rate Test IPCLRT e
. IRM -
_ Intermediate Range Monitor ISI
' Inservice: Inspection Licensee Event Report LER LLRT' Local Leak. Rate Test LPCI.
Low Pressure Coolant Injection Mode of RHRS LPRM Local Power Range Monttor MAPLHGR Maximum-Average. Planar Linear Heat-Generation Rate Minimum Critical Power Ratio MCPR MFLCPR-Maximum fraction. Limiting Critical Power Ratio MPC Maximum Permissible Concentration
. Main Steam Isolation Valve MSIV NIOSH National Institute for Occupational. Safety and Health Primary-Containment Isolation
- PCI, Preconditioning Interim Operating Management Recommendations PCIOMR-Reactor Building Closed Cooling Hater System
-RBCCH
~-
RBM:
Rod Block-Monitor-Reactor Core Isolation. Cooling. System RCIC:
RHRS-Residual Heat Removal _ System RPS' Reactor Protection-System RHM Rod Horth' Minimizer Standby Gas-Treatment System SBGTS Standby. Liquid Control SBLC Shutdown ~ Cooling Mode of RHRS
-SDV:
Source Range Monitor _-
SRM,
TBCCH.
Turbine Building Closed Cooling Hater System Traversing Incore_ Probe TIP
~ Technical Support Center TSC 0027H/0061Z
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