ML20245G243

Monthly Operating Repts for Jul 1989 for Quad-Cities Nuclear Power Station Units 1 & 2

ML20245G243
Person / Time
Site:	Quad Cities
Issue date:	07/31/1989
From:	Deelsnyder L, Robey R COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
RAR-89-50, NUDOCS 8908150300
Download: ML20245G243 (28)

Text

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Cordova, Illinois 61242-9740 Telephone 309/054 2241 RAR-89-50 August 1, 1989 Director of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Mail Station P1-137 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 montn of July, 1989.

Respectfully, COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION f.h.

R. A. Robey Technical Superintendent RAR/LFD/vmk Enclosure gvf 8908150300 epo731 FDR ADOCK 05000254 R

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4 QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT JULY, 1989 COMMONWEALTH EDISON COMPANY AND IOWA-ILLINOIS GAS & ELECTRIC COMPANY NRC DOCKET N05. 50-254 AND 50-265 LICENSE NOS. DPR-29 AND DPR-30 O

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QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT JULY, 1989 COMMONWEALTH EDISON COMPANY AND IOWA-ILLINOIS GAS & ELECTRIC COMFt.:,'Y NRC DOCKET N05, 50-254 AND 50-265 LICENSE NOS. DPR-29 AND DPR-30 O

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1 TABLE OF CONTENTS I.

Introduction II.

Summary of Operating Experience A.

Unit One B.

Unit Two II I..

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 l

VII.

Refueling Information VIII.

Glossary l

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INTRODUCTION Quad-Citiet lu; lear Power Station is composed of two Boiling Water Reactors, each v,.a 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, respectively were October 18, 1971, and Apri t i:6,1972. 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.

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III. PLANT OR PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE A.

Amendments 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 requiring NRC approval for the reporting period.

C.

Tests and Experiments Requiring NRC Approval There were no Tests or E.:periments 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 One and Two 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.

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SUMMARY

OF OPERATING EXPERIENCE l

A.

Unit One Unit One began the month of July in the Shutdown mode. An investigation revealed the cause for this event was a loose connection on the condenser

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low vacuum pressure switch irdicating lamp. When the lens cover for the lamp was put on, the loose wire induced a voltage in the K2D18 relay and energized the master trip bus. This resulted in a turbine trip. The connection for the indicating light was tightened, and further testing of che lead could not duplicate the turbine trip signal. All the connections 1

associated with the Electro Hydraulic Control system were checked for tightness on July 1.

The low vacuum indicating light connection was the only identified loose connection.

At 1924 hours0.0223 days <br />0.534 hours <br />0.00318 weeks <br />7.32082e-4 months <br />, on July 2. reactor startup was commenced. At 2105 hours0.0244 days <br />0.585 hours <br />0.00348 weeks <br />8.009525e-4 months <br />, the reactor was made critical and at 1043 hours0.0121 days <br />0.29 hours <br />0.00172 weeks <br />3.968615e-4 months <br /> on July 3, the main generator was synchronized to the grid.

Increases in power levels were begun using control rods per the request of the Chicago Load Dispatcher. However, load was held constant at 270 MWe while investigations were made into a steam leak discovered on the 1A Moisture Separator Drain Tank Vent Line.

There was also mechanical damage to the moisture separator drcin tank.

The damage appears to be caused by excessive vibration. At 1340 hours0.0155 days <br />0.372 hours <br />0.00222 weeks <br />5.0987e-4 months <br />, load was reduced to 150 MWe.

At 2147 hours0.0248 days <br />0.596 hours <br />0.00355 weeks <br />8.169335e-4 months <br />, the main generator was taken off-line and at 2217 hours0.0257 days <br />0.616 hours <br />0.00367 weeks <br />8.435685e-4 months <br />, a manual scram was inserted due to the leak on the vent line. The line was repaired and on July 4 reactor startup was commenced. At 1816 hours0.021 days <br />0.504 hours <br />0.003 weeks <br />6.90988e-4 months <br />, the reactor was made critical, and at 1425 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.422125e-4 months <br /> on July 5, the main generetor was synchronized to the grid.

Control rod maneuvers were begun to increase power levels beginning at 100 MWe, Power levels were held constant at 250 MWe until July 6.

At 0725 hours0.00839 days <br />0.201 hours <br />0.0012 weeks <br />2.758625e-4 months <br />, load was increased to 785 MWe with control rods. At 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />, a power reduction was taken to 450 MWe.

Power levels were'neld constant until July 7.

At 1035 hours0.012 days <br />0.288 hours <br />0.00171 weeks <br />3.938175e-4 months <br />, load was increased to full power at the request of the Chicago Load Dispatcher. At 1812 hours0.021 days <br />0.503 hours <br />0.003 weeks <br />6.89466e-4 months <br />, power levels were reduced due to increasing river temperatures.

From July 8 to July 13, high river temperatures continued and power levels were adjusted accordingly.

On July 13, at 0945 hours0.0109 days <br />0.263 hours <br />0.00156 weeks <br />3.595725e-4 months <br />, the unit was taken to 500 MWe.

On July 14, at i

1005 hours0.0116 days <br />0.279 hours <br />0.00166 weeks <br />3.824025e-4 months <br />, the unit was taken to maximum attainable power.

From July 15 thru 19, power levels were adjusted according to the demands of the Chicago Load Dispatcher. At 0835 hours0.00966 days <br />0.232 hours <br />0.00138 weeks <br />3.177175e-4 months <br />, a feedwater heater transient caused main steam drain tank levels to peg high. Reactor power was reduced to 635 MWe to help prevent a turbine trip until the tripped heaters could be relatched.

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g Heaters were put back to normal'and power ascension was begun at 0910 hours0.0105 days <br />0.253 hours <br />0.0015 weeks <br />3.46255e-4 months <br />.

At 1807 hours0.0209 days <br />0.502 hours <br />0.00299 weeks <br />6.875635e-4 months <br />, the unit was placed in Economic Generation Control (EGC).

On July 20, at 0055 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br />, EGC was tripped and recirculation pumps placed in manual. A load reduction to 500 MWe was taken at the request of the Load Dispatcher. At 0858 hours0.00993 days <br />0.238 hours <br />0.00142 weeks <br />3.26469e-4 months <br />, a load increase to 745 MWe was taken.

On July 21, at'0230 hours0.00266 days <br />0.0639 hours <br />3.80291e-4 weeks <br />8.7515e-5 months <br />, a power reduction was taken to 450 MWe per the Load Dispatcher. At 0820 hours0.00949 days <br />0.228 hours <br />0.00136 weeks <br />3.1201e-4 months <br />, power was increased with recirculation pumps to 755 MWe.

On July 22, at 0639 hours0.0074 days <br />0.178 hours <br />0.00106 weeks <br />2.431395e-4 months <br />, the unit was placed in EGC.

F' rom July 22 thru the remainder of the month, the unit operated near maximum power or remained in EGC while normal operational activities and routine surveillance were conducted. Maximum attainable thermal power continues to decrease due to coastdown operation.

B.

Unit Two Unit Two began the month of July operating at 450 MWe.

From July 1 thru July 5, normal operational activities occurred.

Power levels were adjusted according to the requests of the Chicago Load Dispatcher. Routine sur-veillances were performed. On July 6, at 1023 hours0.0118 days <br />0.284 hours <br />0.00169 weeks <br />3.892515e-4 months <br />, the control room received spurious Channel B Main Steam Tunnel High Temperature alarms and Group I Isolation Channel Trip Alarms. All alarms were reset and the 1/2 Group I alarm was reset. Upon investigation, it was discovered that the alarms were from apparent failure of the main steam isolation valve room temperature switch.

Investigations are being made under the "near miss" program.

On July 7, at 0405 hours0.00469 days <br />0.113 hours <br />6.696429e-4 weeks <br />1.541025e-4 months <br />, power levels were adjusted and the unit was placed in Economic Generation Control (EGC). At 0901 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.428305e-4 months <br />, EGC was tripped due to Load Dispatcher comptroller problems. On July 8, at 1740 hours0.0201 days <br />0.483 hours <br />0.00288 weeks <br />6.6207e-4 months <br />, power was reduced to 250 MWe due to increasing river temperature limitations.

For the next three days, until July 11, power levels were held below 300 MWe due to continued high river temperatures. On July 12 and 13, power levels remained constant while routine surveillance were performed. On July 13, at 1210 hours0.014 days <br />0.336 hours <br />0.002 weeks <br />4.60405e-4 months <br />, a load increase was begun. 500 MWe was achieved at 1625 hours0.0188 days <br />0.451 hours <br />0.00269 weeks <br />6.183125e-4 months <br />.

On July 14, at 1030 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91915e-4 months <br />, a load increase to 800 MWe was taken.

Between July 15 and July 18, normal operational activities and routine surveillance were conducted. Power levels were adjusted accordingly, and the demands of the Chicago Load Dispatcher were met.

Beginning July 18, load was held at approximately 650 MWe due to core thermal limits calculation program difficulties. On July 23 the computer program was corrected and at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, 800 MWe was reached.

l For the remainder of the month, normal operational activities and routine surveillance were performed. Power levels were adjusted accordingly, and the demands of the Load Dispatcher were met.

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i-UNIT 1 MAINTENANCE

SUMMARY

WORK REQUEST NO.:

Q73382 LER NUMBER: N/A COMPONENT: System 2400 - While in RUN mode at 100% of rated core thermal power, the "H2-02 Monitor Common Failure" alarm was received due to a low sample chamber temperature on the IB channel of Containment Atmospheric Monitoring-(CAM). The 1B CAM was, therefore, declared inoperable.

CAUSE OF MALFUNCTION: The IB CAM was inopreable due to a wire obstructing the switch's resetting capabilities. The wire could have been positioned during assembly such that a small amount of vibration caused the wire to interfere with the operation of the switch.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal since High Radiation Sampling System (HRSS) was operable as a redundant system during the time that the CAM systems were inoperable.

If all drywell hydrogen monitoring capability had been lost, continued reactor operation would have been permissible for up to seven days according to Technical Specifications.

In addition, the low temperature alarm insured that the problem was discovered quickly, and the IB CAM was inoperable only a short time.

ACTION TAKEN TO PREVENT REPETITION: The IB CAM was repaired by repositioning the wire that was obstructing its resetting capabilities under Work Request Q73382.

The IB CAM was tested and declared operable..

WORK REQUEST NO.:

Q75125, Q75136, Q75137 LER NUMBER:

89-004 COMPONENT:

System 203 - While performing QOS 0201-S1, " Auto-Pressure Relief System Manual Operation of Relief Valves", the 1-203-3D valve stuck open. After numerous unsuccessful attempts to close the valve, the reactor was manually scrammed per procedure.

All other systems operated as expected. An Unusual Event was declared.

CAUSE OF MALFUNCTION: The cause of the event was determined to be component failure. The relief valve stayed in the open position due to a combination of failures. The pilot valve showed signs of steam leakage past the seat, and the 1/16" drain orifice in the disc retainer of the main valve was plugged with a small piece of metal. The steam leakage past the pilot valve seat was normal wear.

RESULTS & EFFECTS ON SAFE OPERATION: The safety significance of the event I

was minimal. All Engineered Safeguard Feature (ESF) actuations occurred as expected to bring the reactor to a safe shutdown condition. The relief valve closed when the spring pressure overcame reactor pressure, and the Unusual Event was terminated i

when the reactor was in a cold SHUTDOWN condition.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to replace the 1-203-3D electromatic relief valve, including its pilot valve. This was done under Work Request Q75124.

The pilot valves for the 1-203-3B and 1-203-3E were replaced under Work Requests Q75136 and Q67927, respectively. The pilot valves j

were rebuilt under Work Requests Q75137 and Q75125.

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WORK REQUEST NO.:

Q75836, Q76079 LER NUMBER: N/A l

COMPONENT:

System 1600 - While performing QOS 1600-25, Weekly Suppression Chamber Level Verification, it was discovered that the' suppression chamber narrow range level recorder 1-1602-7 differed from the actual suppression chamber level by more than 0.5 inches, thus rendering the recorder inoperable.

The recorder reading was 0.6 inches and the sightglass actual reading was 0 inches. Work Request Q75836 was written to investigate and repair the recorder.

A separate incident showed that the level recorder 1-1602-7 had steadily increased from 0.1 inches to 0.6 inches. The actual suppression chamber level was found to be 0 inches. The recorder was again declared inoperable. Woik Request Q76079 was written to investigate and repair the recorder.

CAUSE OF MALFUNCTION: The apparent cause of these events was moisture in the dry leg of the instrument line.

Possible sources of the moisture were:

1.

Infrequent draining of the sensing line; 2.

Leaking valves; 3.

Procedure that introduces moisture during calibration.

In procedure QIS 54-2, the instrument manifold bypass valve is opened which could introduce a small amount of moisture into the dry leg of the instrument during each calibration.

RESULTS & EFFECTS ON SAFE OPERATION: The safet:r significance of the event was minimal because_the suppression chamber level differed from the level indicated on the recorder by no more than 0.6 inches. Thus, Technical Specifi-cation limits were not exceeded.

In addition, redundant equipment was available.

A local sightglass and narrow range level indication and two wide range indicators were available.

ACTION TAKEN TO PREVENT REPETITION: From a similar event delineated in Daviation Report (DVR) 4-1-88-071, a procedure change was initiated to delete the opening of the bypass valve, require more frequent draining of the dry leg, and to note the amount of water found during the dry leg draining.

WORK REQUEST NO.: Q76346, Q76347, Q76348, Q76349, Q76358

,LER NUMBER:

89-008 COMPONENT:

System 1600 - While performing QOS 1600-15, Pressure Suppression Systems Power-Operated Valve Fail-Safe Testing Quarterly, the air-operated (AO)

.1-1601-23 valve would not fail in the closed (fail-safe) position. A0 1-1601-23 is the inboard vent isolation valve to the drywell and is required to fail in the closed position for primary containment isolation. A0 1-1601-23 was declared inoperable, and QOS 1600-01, Containment System Isolation Valve Inoperable Outage Report, was initiated. An investigation of the manifold on the' air operator for valve 1-1601-23 showed that the check valves installed in the air supply line to valves 1-1601-24 and 1-1601-63 were not all the same.

Several valves were check valves and several were flow / check valvee.

CAUSE OF MALFUNCTION: The apparent cause of the failure of valve 1-1601-23 air operator was the use of a flow / check valve in place of a check valve. The flow / check valve was installed initially in the air operator. When installed backwards with the needle valve fully closed, the flow / check valve acts like a check valve.

RESULTS & EFFECTS ON SAFE OPERATION: This event was potentially significant

.because in the event of a loss of instrument air, the flow / check control valves would not have acted as check valves. This would have potentially prevented

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the valve from failing in the safe position.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to take the affected valves out of service in their respective fail-safe positions.

The flow / check control valves were replaced with check valves in the affected air operators under the following Work Requests:

A0 1-1601-21 Q76346 A0 1-1601-22 Q76347 A0 1-1601-23 Q76283 A0 1-1601-24 Q76348 A0 1-1601-56 Q76349 A0 1-1601-60 Q76358 l

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UNIT.' MAINTENANCE

SUMMARY

WORK REQUEST No.: Q65830

.LER NUMBER: 88-007 COMPONENT:. System 220 - Inboard feedwater check valve 2-220-58B failed local l

leak rate test (LLRT).

l l-CAUSE OF MAIFUNCT10N: The cause of the valve leaking could not be determined until all repairs had been completed and tested. At the time of completion of repairs, a supplemental report would be submitted.

RESULTS 6 EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal since local leak rate testing is a conservative method of measuring containment leakage. During accident conditions, the actual leakage would be less than that determined by local leak rate testing since some lines would l

be pressurized with water instead of air and some non-primary containment isolation valves would perform to isolate the primary containment.

ACTION TAKEN TO PREVENT REPETITION:

No corrective actions were completed at the time of the report, as a supplemental report was being awaited for a listing of necessary repairs.

WORK REQUEST No.: Q76018 LER NUMBER: N/A COMPONENT: System 203 - While operating personnel were in the drywell, it was discovered that the outlet flange drain line on electromatic relief valve (ERV) 2-203-3D was broken off. The valve was taken out of service, and Work Request Q76018 was written to repair the valve at the next outage. During a short outage the drain line was replaced and the valve was declared operable.

CAUSE OF MALFUNCTION: The cause of the broken drain 1 r not known at 4

the time of.this report. The fractured area of the dral

.aae was sent to Systems Material Analysis Department (SMAD) for analysis and evaluation. A supplemental report is to be issued when the results are received.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal because Technical Specifications do not place any Limiting Conditions for Operating (LCD) on the unit until two ERV's are declared inoperable. Techni-cal Specifications state that the loss of one ERV does not significantly reduce the pressure relieving capability.

In addition, the other three ERV's were operable, as well as all eight safety valves and the Target / Rock safety / relief valve.

ACTION TAKEN TO PREVENT REPETITION:

The drain line was replaced like-for-like under Work Request Q76018. A root cause failure analysis is being done by SMAD, and Engineering is evaluating the adequacy of t he supports on the ERV drain lines.

WORK REQUEST No.:

Q66221 LER NUMBER: N/A COMPONENT:

System 1100 - While performing QOS 1100-4, " Standby Liquid Control (SBLC) Relief Valve Setpoint Check", relief valve RV-2-1105B was found not to fully lift within a system pressure of 1,455 to 1,545 psig.

System pressure was raised to 1,700 psig before being throttled back down. The relief valves are installed to prevent over-pressurization of the SBLC systems.

CAUSE OF MALTL'NCTION: The cause of the deviation was that the valve's resent pressure was too high. The blowdown guide adjusting ring, which controla rescat pressure, was found to be set lower than the manufacturer's specifications.

The valve was then lifting and. seating at approximately the same pressure.

This would not allow a large enough volume of water through the valve and caused the valve to not relieve within its range. The valve was disassembled in August, 1987 and a new valve disc was installed. The blowdown guide adjusting ring was not placed in its original position during reassembly of the valve.

In addition, the valve was not tested per QOS 1100-4 after being reinstalled in August, 1987.

RESULTS & EFFECTS ON SAFE OPERATION: The safety implications of this condition were minimal because the redundant pump, 2A-1102, and relief valve, RV-2-1105A, were fully functional.

Relief valve, RV-2-1105B, was not relieving below its required setpoint; therefore, the 2B pump was demonstrated to be capable of meeting the 40 gpm flow requirement at 1,275 psig as required by Technical Specifications.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to initiate Work Request Q66221 to investigate and repair the problem. The valve was disassembled and inspected by Mechanical Maintenance. However, the relief and blowdown ring setpoints were not adjusted. The valve again did not fully open after re-assembly. Work Request Q66635 was written to further investigate the problem. The blowdown ring setpoint was adjusted and the valve finally i

opened fully.

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1 WORK REQUEST No.: Q69723-LER NUMBER: N/A L

COMPONENT:

System 1000 - While the Instrument Maintenance Department was perform-L ing QIS 6-1, "High Drywell Pressure SCRAM Calibration", it was discovered that pressure switch 2-1001-88A tripped at 2.56 psig, instead of less than or equal to 2.5 psig as required by Technical Specifications. The switch was immediately recalibrates to 2.41 psig.

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CAUSE OF MALFUNCTION: The cause of the event was instrument setpoint drift.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of plant and personnel was 1.

not affected during this event.

High drywell pressure switches 2-1001-88A, B, C and D provide SCRAM and Group 2 isolation functions. These switches serve as a backup to the SCRAM function and the closure of Group 2 isolation valves.

These functions would not have been inhibited by the setpoint drift of this single pressure switch because the switches are arranged in a one-out-of-two taken twice logic and the other three switches were within lechnical Specifi-cation limits.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to calibrate the pressure switch to 2.41 psig.

The switch was replaced under Work Request Q69723. The removed pressure switch was sent to Static-0-Ring for examination because of the large number of instances of setpoint drift in these pressure switches.

WORK REQUEST No.: Q74621, Q75344 LER NUMBER: N/A COMPONENT:

System 1000 - While performing Temporary Procedure 5737, " Unit 2 B Loop Residual Heat Removal Service Water (RHRSW) Pump Operability Surveillance",

I it was found that the 2C RHRSW pump would not meet Technical Specifications requirements of 3,500 gallons per minute at 198 psig. A 3,500 gpm flow rate was obtained at a pressure of 180 psig.

CAUSE OF MALFUNCTION: The cause of this deviation was normal pump wear. The cause was determined to be normal wear as a result of steady flow rate drops observed in surveillance over the past two years.

RESULTS & EFFECTS ON SAFE OPERATION:

The safety significance of this event was minimal. The "B" containment cooling and "B" shutdowns cooling loops were still available with the operable 2D RHRSW pump. All other components of contain-ment cooling were also operable.

ACTION TAKEN TO PREVENT REPETITION:

Engineering analysis determined that resizing the impellers vould reduce the cavitation.

The impellers were resized for modification 4-2-87-002C, and they were replaced under Work Request (WR) Q74621.

Post-maintenance performance testing and investigations were done under WR Q75344.

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. WORK REQUEST No.: Q76350, Q76351, Q76352, Q76353, Q76354, Q76357 LER NUMBER:

89-008

' COMPONENT:

System 1600 - While performing QOS 1600-15, Pressure Suppression Systems Power-Operated Valve Fail-Safe Testing Quarterly, the air-operated (AO) 1-1601-23 valve would not fail in the closed-(fail-safe) position. A0 1-1601-23 is the inboard vent isolation valve to the drywell and is required to fail in

.the closed position for primary containment isolation. AD 1-1601-23 was declared inoperable, and QOS-1600-01, Containment System Isolation Valve Inoperable Outage Report, was initiated. An investigation of the manifold on the air operator for valve 1-1601-23 showed that the check valves installed in the air supply line to valves 1-1601-24 and 1-1601-63 were not all the same.

Several valves were check valves and several were flow / check valves.

CAUSE OF MALFUNCTION: The apparent cause of the failure of valve 1-1601-23 air operator was the use of a flow / check valve in place of a check valve.

The flow / check valve was installed initially in the air operator.

When installed backwards with the needle valve fully closed, the flow / check valve acts like a check valve.

RESULTS & EFFECTS ON SAFE OPERATION: This event was potentially significant because in the event of a loss of instrument air, the flow / check control valves would not have acted as check valves. This would have potentially prevented the valve from failing in the safe position.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to take the affected valves out of service in their respective fail-safe positions.

The flow / check control valves were replaced with check valves in the affected air operators under the following Work Requests:

A0 2-1601-21 Q76350 A0 2-1601-22 Q76351 A0 2-1601-23 Q76352 A0 2-1601-24 Q76353 A0 2-1601-56 Q76354 A0 2-1601-60 Q76357 i

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IV.

LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for l

Quad-Cities Units One and Two occurring during the reporting period, l

pursuant to the reportable occurrence reporting requirements as set forth l

In sections 6.6.B.1. and 6.6.B.2. of the Technical Specifications.

UNIT 1 Licensee Event Report Number Date Title pf Occurence 89-010 6-29-89 Reactor Scram from Turbine Trip 89-011 7-7-89 Diesel Generator Vent.

Supply Damper Not Working Properly 1-9472-032 UNIT 2 There were no Licensee Event Reports for Unit 2 for this reporting period.

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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 0027H/0061Z

A APPENOlX C OPERATING DATA REPOR7 OOCKET NO.

50-254 UNIT One OATE Aueust 7, 1989 COMPLETED NY LYnne Deelsnyder l

TELEPHONg 309-654-2241 OI43tATifMBSTA'11aB 0000 070189 2!.00 073189 744

1. REPORTINe PEnl00:

GROSE MOURE IN REPORTINO PER800:

2511 MAX. 0EPENO. CAPACITY lemuretss) 769

8. CURRENTLY AUTHORIGEO POWER LEVEL gg9 DEE40N ELECTRICAL RATING lamurasso: -

N/A

3. POWER LEVEL 70 WHICH RETRICTED llP ANY) (MuupNot):

4. REASOIS POR REETRICT80N (47 ANYl:

TMitMONTH YM TO OATE CUlfULATIVE 4806.3 122348.5 E. NUMEGR OF MOURE REACTOR WAS CRITICAL.............. 678. 9 0.0 0.0 3421.9 E. REACTOR RESERVE EMUTDOWN MOURE.................. 644.7 4703.0 118362.2 T. MOURE GENERATOR 088 UNE..........................

0.0 0.0 909.2 E. UNIT RESERVE EMUTIF)W,4 MOURE......................

1224886 10431645 252121724

9. GROSE THERMAL ENERGY GENERATED IMWM) 378704 3364314 81721927

10. GROSE ELECTRICAL ENERGY GENERATED IMWHI.............

359049

'3213661 76778935 1t. NET ELECTRICAL ENERGY GENERATE 0 (MWM) 81.0l

~

91.3 94.5 a uCTOR.ERViCE P ACT0R..................

91.3 94.5 83.3

13. REACTOR Av AILAtiLITY P ACTOR......................

86.7 92.5 78.4

14. UNef SERVICE F ACTOR..............................

86.7 92.5 79.0

15. UNIT AVAILABILITY P ACTOR..........................

62.8 82.2 66.I it. UNIT CAPACITY P ACTOR lusing MOCl.....................

17. UNIT CAPACITY PACTOR (Uung Design MWel.................

tt. UNIT PORCEO OUTAGE RATE,.........................

18. SMUTDOWNE SCHEOUL50 CVER NEXT E MONTME (TYPE. DATE. ANO QURATION OP EACN):

20. IP EMUT DOWN AT END OF REPORT PERIOD. ESTIMATED DATE OP STARTUP:

21. UNITE IN TEST STATUS (PRIOR TO COMMERCI AL OPERATIONi:

FORECAST ACHIEVED INITIA6.CRITICAUTY INITIAL ELECTRICITY i

CO.M.

CIAL OPERAT,0 4

)

)

1.lM

o

+

APPENOtX C OPERATING DATA REPORT DOCKET NO.

50-265 L".YI UNIT Tw DATE Aunust 7,1989 COMPLETED BY Lynne Deelsnyder l

TEl.E*NONE 309-654-2241 OPERATING STATUS 0000 070189 2400 073189 744

1. REPORTlass PERs00:

GROes M0uRs IN REPORTING PERIOD:

2. CURRENTLY AUTHORl280 POWER LEVEL (tRett. 2511MAX. OEPEND. CAPACITY teReo 8esel. 769 OSB8GN ELECTRICAL RATifeG teres 8esti:

789 N/A

3. POWER LEVEL TO WHICH REETRlCTED (IP ANYI tenus.sesti:

4. REASONE POR RSBTRICTt006 44P A88Yl:

TMtB MO8fTM VR TO DATE CutsuLATIVE 4970.6 115920.5 E. NutsesR OF Mouns REACTOR WAS CRITICAL.............. 744.0 0.0 0.0 2985.8

8. REACTOR RESERVE SMUT 00gWe MOURE...................

4928.6 112660.3 T. MOURS GENERATOR ON L8988......................... 74 4. 0 0.0 0.0 702.9

8. UNIT RESERVE SMUTOOWN MOURS......................

9. GROSS THERMAL ENERGY GENERATED (MWMi............. 1431158 10733676 241643949 448346 3476955 77410426

10. Gross ELECTRICAL ENERGY GENERATED IMWM).............

426530 3325525 73062102

11. NET ELECTRICAL ENERGY GENERATEO IMWHI 97.7 77.2

12.

• EACTOR SERVICE P ACTOR.......................... 100. 0 100.0 97.7

, 9. 2_

13. REACTOR AV AILASILITY P ACTOR......................

96.9 75.0

14. UNIT SERVICE F ACTOR............................. 10 0. 0 96.9 75.5

15. UNIT AVAILABILITY P ACTOR......................... 100. 0

18. UNIT CAPACITY PACTOR (Using MOCl.....................

85.0 63.3 74.6 72.7 82.9 61.7

17. UNIT CAPACITY PACTOR tumag Osnan iduel.................

0.0 3.1 8.2

18. UNIT PORC80 0UTAGE RATE..........................

19. SMUT 00WN8 SCHEOULEO OVER NExT E MONTHS (TYPE. DATE. ANO DURATION OF EACHl:

20. IF SMUT 00W84 AT END OF REPORT PEh100. ESTIMATED DATE OP STARTUP:

21. UNITS IN TEST STATUS (PRIOR TO COMMERCI AL QPERATIONi:

FORECAST ACHIEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMesERCIAL OPERATl000 1.lM

q.

o

,M!1.-

APPENDIX B AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-254 1

UNIT one DATE Aucust 1,'1989 COMPLETED BY Lvnna Deelsnyder TELEPHONE 309-654-2241

. MONTH July, 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POP ER LEVEL (MWe-Net)

(MWe Net; 1

-14 17 643 2

-13 13 724 52 3'

jg 688 4

-14 662 g

g?

59 629 5

gi 469 8

22 700 7

614 23-651 503 6

34 682 g

226 25 670 367 648 10-28 11 229 643 2,

12 226-g 644 13 377 650 3

624 623 14 g

659 15 33 626 Ig 634 INSTRUCTIONS On this form, list the average daily unit power level in MWe. Net for each day in the reporting month. Compute to the neascst whole megawatt.

These figures will be used to plot a graph for cach 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't line (or the restruted power level line). In such cases, the average daily unit power output sheet should be footnoted to explam the apparent anomaly.

1.16 8

~

b 1

APPENDIX B,

I AVERAGE DAILY UNIT POWER LEVEL J

DOCKET NO.

50-265

{

UNIT Two l

DATE Aucost 1, 1989 l

COMPLETED BY.lgmne Deelsnyder TELEPHONE 309-654-2241 l

MONTH July, 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 625 1

37 624 2

609 538 gg 3

630 613 g,

575 4

a 619 g@

576 5

g 613 674 621 6

22 y

669 604 23 423 664 3

24 g

241 3

698 10 390 716 28 237 11 gy 636 12 242 671 g

740 13 330 g

14 617 684 y

15 676 gj 673 16 583 INSTRUCTIONS On this form, list the average daily unit power level in MWe Net for each day in the reporting month. Compute to the neassst whole megawatt.

These figures will be used to plot a graph for cach reporting month. Note that when maximum dependable capacityis used for the net electrical rating of the unit, there may be occasions when the daily average power level exceeds the 1001 line (or the restruted power level line). In such cases, the average daily unit power output sheet should be footnoted to explain the apparent anomaly.

1.164

1. 4

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^

VI. UNIOUE 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 The basis for reporting this data to the Nuclear Regulatory Commission are specified in the surveillance requirements of Technical Specifications 4.3.C.1 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.

0027H/0061Z

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~

f 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.

i 0027H/0061Z E______

f.

~

QTP 300-532 Revision 1 QUAD-CITIES REFUELING March 1978 INFORMATION REQUEST 1.

Unit:

01 Reload:

9 Cycle:

10 2.

Scheduled date for next refueling shutdown:

9-9-89 3

scheduled date for restart following refueling:

12-11-89 4.

Will refueling or resumption of operation thereafter require a technical specification change or other license amendment:

NOT AS YET DETERMINED.

5.

Scheduled date(s) for submitting proposed IIcensing action and supporting information:

JUNE 10, 1989 6.

Important IIconsing considerations associated with refueIIng, 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 assembiles in core:

724 b.

Number of assemblies in spent fuel pool:

1773 8.

The present IIcensed 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:

0 9

The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present IIconsed capacity: 2008 3E f* F8 Ft C) if Et C)

. APR 2 01978 C3. CL (). S. Ft.

7.-

g QTP 300-532 Revision 1 QUAD-CITIES REFUELING March 1978 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:

5-7-90 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:

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.

7.

The number of fuel assemblies.

a.

Number of assemblies in core:

724 b.

Number of assemblies in spent fuel pool:

1475 8.

The present IIcensed 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: 2008 XPPROVED

-i-APR 2 01978 i

Q. C. O. S. R.

____I___

_.~I.__.___

3 5..

i*

l 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 BWR Boiling 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 Critical Power Ratio MFLCPR Maximum Fraction Limiting Critical Power Ratio Maximum Permissible Concentration MPC MSIV Main Steam Isolation Valve NIOSH National Institute for Occupational Safety and Health Primary Containment Isolation PCI PCIOMR Preconditioning Interim Operating Management Recommendations RBCCW

. Reactor Building Closed Cooling Kater System RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling System RHRS Residual Heat Removal System RPS Reactor Protection System RWM Rod North Minimizer SBGTS Standby Gas Treatment System SBLC Standby Liquid Control SDC Shutdown Cooling Mode of RHRS SDV Scram Discharge Volume SRM Source Range Monitor TBCCW Turbine Building Closed Cooling Water System TIP Traversing Incore Probe TSC Technical Support Center 0027H/0061Z

- _ _ _ _ _ _ _ _