ML20012B801
| ML20012B801 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/28/1990 |
| From: | Deelsnyder L, Robey R COMMONWEALTH EDISON CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 0027H-0061Z, 27H-61Z, RAR-90-26, NUDOCS 9003160313 | |
| Download: ML20012B801 (22) | |
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s w '2 Commonwealth Edison 4 - Quad Cities Nuclear Power Station 22710 206 Avenue North
- Cordova. lilinois $1242 Telephone 309/654-2241
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.RAR-90-26 March.2, 1990 Director of Nuclear Reactor' Regulations U. S. Nuclear Regulatory Commission.
Mail. Station P.1-137
.Hashington. 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 February, 1990.
Respectfully,
- COMMONHEALTH. EDISON COMPANY
-QUAD-CITIES NUCLEAR.. POWER STATION.
h[A..Robey--& R. 87 Technical Superintendent-
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Enclosure 0027H/00612 / s Il 1 9003160313 900223 -ra#, ii
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y, , tc ;. . . ,, PT .- ..- {> . p A: QUAD-CITIES NUCLEAR POWER STATION
, UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT February, 1990 COMMONHEALTH EDISON COMPANY t
AND IONA-ILLINOIS GAS-& ELECTRIC-COMPANY NRC DOCKET NOS. 50-254-AND 50-265 LICENSE NOS. DPR-29 AND DPR-30
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't TABLE OF CONTENTS
- 1. . Introduction v 'II . Summary of Operating Experience A. Unit One B. Unit Two p '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
~l IV. Licensee Event Reports .j V. Data Tabulations A. Operating Data Report i l B. Average Daily Unit Power Level. {
, C. Unit-Shutdowns and Power Reductions ] VI., Unique Reporting Requirements l A. ' Main Steam Relief Valve Operations '! B. Control' Rod Drive' Scram Timing Data VII. ' Refueling Information i
'VIII. Glossary -l l
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4-h b 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 Boiling 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 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,1respectively 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. V number 309-654-2241, extensions 2185 and 2240. 0027H/00612
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SUMMARY
OF OPERATING EXPERIENCE s c- [" A. - Unit:One . l Unit:On'e began the-month of February operating at full load. On February 2, n load was reduced for 1A Feedwater Regulating Valve maintenance. Unit was taken.to maximum power. level after maintenance was' completed. 10n' February 4, at 0754 hours,-a GSEP unusual event was declared due to the i ' failure of the Secondary Containment Capability Test. This is a test of' l secondary containment capability to maintain-an average 0.25 inch water-vacuum under calm wind conditions, with-a Standby Gas Treatment System filter' train flow of not more than 4000 cfm which must be demonstrated at each refueling outage prior to-refueling. Power was. reduced to 168'MWe. Temporary repairs-were made to several penetrations. The test was then n' successfully completed and the GSEP terminated at 1642 hours. The temporary repairs will be inspected periodically until permanent repairs can be-completed. . Power' levels were increased to 395 MWe, then held constant due to.10 Feed Pump problems. , .On February 6, an' ascent'to full power was taken. For the remainder of-the month, normal operational' activities occurred and routine surveillances
.were performed. Power levels remained near full power or were adjusted-
- , according to the demands of the Chicago Load Dispatcher.
.i B. Unit Two ;
"~ Unit Two. began the month of February operating'near maximum attainable power .) , . levels in coast down operation. On February 4, at 0116 hours,'a manual scram was inserted and reactor shut-e down was commenced to begin the End of Cycle Ten Refueling Outage for Unit Two.- Normal refueling. activities were performed. including the removal of p the reactor head and steam dryer and core unloading. These activities con-tinued throughout the remainder-of the month. 1 5 ( 4 N.
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h$2[' p j , f. , f b t i: h' . III. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED FMINTCNANCE A. Amendments to Facility License or Technical Speci fications t There were no Amendments to the Facility License or Technical Specifi-cations for the reporting period. 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
, There were no Tests or Experiments requiring NRC approval for the reporting' period.
D. Corrective hbintenance of Safety Related Equipment l i The folicwing summary lists the corrective -maintenance completed on !
- safety related equipment on Unit One and Two during this reporting i- period. This summary includes the following
- Work Request Numbers,
. Licensee Event Report Numbers, Components, Cause of Malfunctions, )
Results and Ef fects on Safe Operation, and Action Taken to Prevent Repetition. Due to the post maintenance reviews, the work may have been performed .l several months earlier than this report. The station is changing the j way maintenance information is gathered to more-accurately reflect the , work.done during the reporting period. i l 1
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. UNIT 1 MAINTENANCE
SUMMARY
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. WORK REQUEST No.: Q79526
[' LER NUMBER: .N/A COMPONENT: System 8350 - Technical Staff found cell #69 of the Unit 1 250V battery leaking.. The battery was' declared inoperable placing Unit Two in a Limited Condition of Operation (LCO) per Technical Specification 3.9.C.3. CAUSE OF MALFUNCTION: The leakage was due to a hairline crack found on cell
#69. Interviews with the battery contractors who were working in the area.
O Land station electricians who'were installing-an overhead lifting rig in the
. area deny dropping anything within the area of the battery. Whichever group caused the. incident, the root cause is attributed to management deficiency.
I- iThe contractors work package did not contain precautions that warned them about not performing this work over the battery. The station electricians did not have any written guidance for hanging the lift on the beam. l RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plant and personnel was not affected by this event. The electrolyte level in cell #69 was above the top of the plates leaving suf ficient ele ';rolyte to allow current to flow without damage to the cell. 3 I ACTION TAKEN'TO PREVENT REPETITION: The immediate corrective action was to : declare the battery inoperable,' transfer the remaining Division II loads to the Unit Two battery, and replace the damaged Unit One battery all per Work ; Request Q79526. l l WORK' REQUEST NO.: Q79954, Q79955, Q79956, Q79957 i LER NUMBER: N/A COMPONENT: . System 3900 - It was discovered during leakage testing of the Unit . One Residual Heat Removal Service Water (RHRSW) vault sump discharge check valves- ; that the:1-3999-515A, 1-3999-516A, 1-3999-515C, and 1-3999-516C check valves ;
-leaked when pressurized in the reverse direction. !
CAUSE OF MALFUNCTION: The cause of this event was the failure of the check 4
- valves in a-system where the fluid is normally dirty and often contains debris I and-foreign particles which can collect in the check valves, j RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of this event l
are minimal. The check valves prevent backflow of sump water into the RHRSW ., vaults. The sump pumps would cycle more often to ensure that the water that leaked in would be pumped out. ]; 4
' ACTION TAKEN TO PREVENT REPETITION: Mechanical Maintenance. disassembled, inspected and repaired the check valves under Work Requests-Q79954, Q79955, Q79956, and
~Q79957. The valve seats were lapped and cleaned.
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t E , , WORK REQUEST NO. - Q72279 , 7 (' - LLER NUMBER:. N/A ~ ECOMPONENT: System 1400 - Instrument Maintenance (IM) personnel were performing-r _Q 1S 23-2, Core Spray Pump Discharge Pressure Functional.- While testing pressure: switch, PS-1-14620, the switch would not reset.1 This switch provides a low-- pressure Emergency Core Cooling System pump-running permissive to the auto blow-g down system. _The IM's repressurized the line in an effort to get the switch E .to reset. :These ' attempts' failed. The switch was then manually reset by pushing ,
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;on the bourdon tube. The alarm and switch reset. ~ QIS 23-1, Core-Spray Pump ,
. Discharge. Pressure Calibration, was done to verify that the switch still tripped -;
i at the proper setpoint. The function test,>QIS 23-2, was. repeated several' times. > E -with no repeat of the original failure. CAUSE OF MALFUNCTION:' The~cause of this event was due to the failure of PS-1-1462C. RESULTS & EFFECTS ON SAFE OPERATION: The safety significance of this event , is minimal. At no time was the logic for-auto blowdown inoperative. , l
-ACTION TAKEN TO PREVENT REPETITION: Under Work Request Q72279, PS-1-1462C was replaced with a like-for-like replacement. 'The new pressure switch was then calibrated and functionally tested per QIS 23.
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SUMMARY
p I 7: WORK REQUEST No.: ~ Q74656 h a
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F - p COMPONENT: . System 700 ^A half scram was received on RPS A channel due to APRM'
. spike.- LPRM 32-49D had'an upscale alarm but was only reading 32 watts /cm2 , -A-qualified l Nuclear Engineer was contacted and permission was given to bypass LPRM 32-49D. A caution card was placed on the bypassed LPRM and Work Request :
-Q74656.was initiated for that:LPRM. .
> r CAUSE OF MALFUNCTION: The cause of this event is a problem with the LPRM's r s ,
,. connector. . LPRM 32-49D used'an Amphenol connector. This type of connector has been found-to be susceptible'to moisture and mechanical problems.
p, RESULTS~& EFFECTS ON SAFE OPERATION: The safety significance of RPS channel A lL tripping due to a LPRM Hi. signal to APRM channel 3 is minimal. The'RPS receives [ :two channels (A and B) to trip before the reactor is scrammed. A spurious signal 1 causing'one channel to trip is not significant due to this two channel design Sof the'RPS. ' n
' ACTION TAKEN TO PREVENT REPETITION: The corrective action implemented was re-
, placing the Amphenol. connector per Work Request Q74656.
UWORK" REQUEST NO.: Q80778
- LER NUMBER: 'N/A COMPONENT: System 1000 - The Instrument Maintenance (IM) Department discovered, while investigating a work-request, that the pressure on the sensing lines off ,
.X-52F was reading approximately l'.05 psig.
The pressure on the lines off of
-X-52E and-_all;the lines going.to the 2201-6 rack were reading' proper.drywell (DW); pressure oflapproximately 1.29 psig.
CAUSE OF MALFUNCTION: The cause of this event is the blockage'of the instrument lines with water. The instrument-lines had~been installed so that moisture could possibly collect inside the line. The probable cause of the water.accumu- 3 lation was;due to condensation of the humid drywell atmosphere inside the sensing
'line.
RESULTS &-EFFECTS ON SAFE OPERATION: The safety of the public and plant personnel - was not effected by this event. The unit was put in the conservative condition
.by tripping the drywell pressure switches associated with penetration X-52F.
y L ACTION-TAKEN TO PREVENT REPETITION: The immediate corrective action was to determine.the: pressure switches that were effected by the blockage of the sensing line:and place the affected instruments in the tripped condition to meet Technical Specifications. IM personnel then purged the sensing line per Work Request
; Q807:7 2. - Surve111ance' tests were performed to verify the correct operation of all pressure switches to initiate their corresponding logic prior to returning
.them to service.
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; Units.Oneland;Two r occurring during the.-. reporting period, pursuant to'the. reportable-N ~ i occurrence reporting requirements;:as Jaet forth ~1n'sectionf 6.6~.B.1=.Eand - 6.6' B.2. . , ;a ..d.
a 1 s of;the Technical: Specifications. Tc.t ::' O : C..M:1,. UNIT'I bP p1 ,
- C2 Licensee ~-Event" k,. - : Report Number' . Da t e '. Title of Occurrence -
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1 APPEM)IX B ;
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U, AVERAGE DA!LY IMIT POER LEVEL R , h i kdfW m" Docket No. 50-254 J Unit One I Date March 6, 1990 .j '
, Completed By Lynne Deelsnyder- .
Telephone. 309-654-2241 ;
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MONTH- FEBRUARY
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' DAY MERAGE DAILY POWER LEVEL ' DAY'AVERFDE DAILY POE R LEVEL' (We-Net) - (We-Net) t 1- 717 17 667 2 756 -1B 794 3 784- 19 - 796 3 '4 446 20- 7%
t E , 5 .382- 21 797 $ , o 6
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-9~ 767 25 796
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12' 7% = 2B .764-
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- INSTRUCTIONS ,
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, , :On this' form, list the average daily unit power level in MWe-Net for each day in the reporting month.
Compute to the nearest whole megawatt.. ,
..These figures will be used to plot a graph for each reporting month. Note that when maximum dependable capacity is used for the net electrical rating of the unit, there may be occasions when the daily average
- power level exceeds the 100% line (or the restricted power level line). In such cases, the average daily
- f unit power output sheet should be footnoted to explain the apparent anomaly.
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-.v AVERAGE DAILY (Mli POER LEVEL 4 4 %
9 r.y Docket No.i 56-265-. b)p:A Unit 1To
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Date~ March 6, 1990 6' W ~ w ' Completed By Lynne Deelsnyder h -Telephone- 309-654-2241-i: , . jt MONTH.- FEBRUMY
' DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DA.ILY POER LEVEL
-(MWeSet) Olie-Net)
- 1 602; 17z -7
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2- 683 18 ; -7 . - 3~ :485 19. :
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- INSTRUCTIONS w . . . .-
I< .On this form, list lthe average daily unit power level in WNet for each day in the reporting month. ltf ii-- s ' Compute to the nearest whole negawatt. - , b These figures will be used to plot a graph for each reporting month. Note that when maximum dependable ) Ns - K , - espacity 'is used for the net electrical rating of the unit, there may be occasions when the daily average -
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g '; . . power level exceeds the 185 line (or the restricted power level line). In such cases, the average daily - K .: unit power output sheet should be footnoted to explain the apparent anomaly. .; A 1.16-8 1
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APPENDIX D - j UNIT SHUTDOWNS AND POWER REDUCTIONS . . .
' DOCKET NO. 50-254 l
l l UNIT NAME Quad Cities Unit One COMPLETED BY 'Lynne Deelsnyder '. ' l l l DATE ' March 6, 1990 REPORT MONTH February, 1990 TELEPHONE 309-654-2241' .g i l .. -g DoO $ w $' cm 3 E hoe ' E se l h@w E h@ LICENSEE m@ 2@ DURATICN @ g"g EVENT' '@u ga NO. DATE REPORT NO. " (HOURS) g CORRECTIVE ACTIONS / COMMENTS 90-2 900204 F 0.0 A 5 90-02 PENETRX- Power Reduction Due to Failure of Secondary Containment Capability Test - Temporary Repairs Made to Several Penetrations e 4 S e e e
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.A-. N.: APPENDIX D- T. UNIT SHUTDOWNS AND POWER. REDUCTIONS 1 DOCKET NO. 50-265 , i- . UNIT NAME Cuad Cities Unit Two COMPIX.TED BY Lynne' Deelsnyder '- DATE March 6. 1990 REPORT MONTH February, 1990 TEIE. PHONE 309-654-2241 -2 t ce z 0 h. zo m se -o o-< w na Ese
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u 3" NO. ~DATE (HOURS) REPORT NO. CORRECTIVE ACTIONS /C0t9ENTS 7 g 90-2 900203 .F- 622.7- LC 1 RC FUEllX End of Cycle. Ten Refuel Outage O e . O
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-VI.. UNIQUE REPORTING REQUIREMENTS- -[ ,.
e In L-p F ' The following items are included in this report based on: prior.comm1tments to the; commission:--- [,' , 'A.. Main Steam" Relief' Valve Operations N There were~no Main Steam Relief Valve Operations for-the-reporting. ,
.. period.
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t G , ll Ii; ' B. Control Rod Drive-Scram Timing Data for Unita One and Two- ' i There was-no Control Rod Drive Scram Timing-Data-for Units _One-- k- and Two for the? reporting period. b . j i s g
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VII..-REFUELING INFORMATION g > r A
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,Thelfo11owing information.about. future reloads atLQuad-Cities Station was. .
%l requested in, a; January? 26,;1978,: licensing memorandum T(78-24) f rom:D.E. O'Brien . .
N, ito C.; Reed,'.et'a1.',1 titled,"Dresden Quad-Cities. and~ Zion ~ Station--NRC. Request j
.for. Refueling Information",Ldated January 18 -1978.
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OTP 300-lu Revision 2 QUAD CIT!!$ htFutLING October 1989 INFORMATION REQU($T
- 1. Unit: 01 koload: 10 1
Cycle: _ 11
- 2. . Scheduled date for nest refueling shutdown:
)
10-6-90 3. Scheduled date for restart following refueling: 12-11-90 4. Will refueling or resunetton of operation thereafter require a Technical ' Specification change or other license amendment. l NOT AS YET DETERMIAED. s 5. Scheduled date(s) for subeltting proposed licensing action and supporting information: JUIY 6, 1990 6. Important Ilconsing considerations associated with refueling e.g., new or different fuel design or supplier, unreviewed design or pe,rformance j analysis methods, significant changes in fuel design, new operating procedures: ' NONE AT PRESENT TIME.
- 7. The number of fuel assemblies. '
- a. Number of asseabiles in core: 726
- b. Number of assemblies in spent fuel pool: 1537
- 8. The present Ilconsed spent fuel pool storage capacity and the site of any increase in 11 censed storage capacity that has been requested or is p14nn6d in number of fuel assemblies:
i
- a. Licensed storage capacity for spent fuel: 3657 i
- b. Planned increase in licensed storage: o
- g. l The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present Ilconsed capacity: 2008 APPROVED ,
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DCT 3 01189 Q.C.O.S.R.
f' QTP 300.$32 ) Revision 2 QUA0 CITIts REFUEL!hG October 194g INFORMATION REQUEST
- 1. Unit: 02 Reload: 9 Cycle: 10
- 2. Scheduled date for next refueling shutdown: 2-3-90
- 3. Scheduled date for restart following refueling: 5-5-90
- 4. N111 refueling or resumption of operation thereafter require a Technical Specification change or other license amendment:
NOT AS YET DETERMINED. I
- 5. Scheduled date(s) for submitting proposed Itcensing action and supporting information: i NOVEMBER 2, 1990
- 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. i
- 7. .The number of fuel assemblies,
- a. Number of assemblies in core: _ 0 '
l b. Number of assemblies in spent fuel pool: 2732
- 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 t
- b. Planned increase in licensed storage: 0 f'
- 9. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2008 4 r
4 APPROVED i 14/0395t N Q.C.O.S.R.
i o . f.. . . . i VIII. GLOSSARY l The following abbreviations which may have been used in the Monthly Report, are defined below: l ACAD/ CAM - Atmospheric Containment Atmospheric Dilution / Containment . Atmospheric Monitoring ANSI - American National Standards Institute APRM - Average Power Range Monitor
, ATHS - Anticipated Transient Without Scram l BWR - Boiling Water Reactor CRD - Control Rod Drive EHC - Electro-Hydraulic Control System EOF - Emergency Operations facility i 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 Critical Power Ratio MFLCPR - Maximum Fraction Limiting Critical Power Ratio MPC - Maximum Permissible Concentration MSIV - Main Steam Isolation Valve , NIOSH -. National 'nstitute for Occupational Safety and Health PCI - Primary Containment Isolation PCIOMR - Preconditioning Interim Operating Management Recommendations , RBCCW - Reactor Building Closed Cooling Hater System ; RBM - Rod Block Monitor . RCIC - Reactor Core Isolation Cooling System RHRS - Residual Heat Removal System RPS - Reactor Protection System RWH - Rod Horth Minimizer SBGTS - Standby Gas Treatment System i 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}}