ML20247D145

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

ML20247D145
Person / Time
Site:	Quad Cities
Issue date:	04/30/1989
From:	Deelsnyder L, Robey R COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RAR-89-31, NUDOCS 8905250184
Download: ML20247D145 (31)

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RAR-89-31, i

May.- 1,11989 Director;of Nuclear Reaccor Regulations U.. S. - Nuclear Regulatory Commission

' Mail 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 April, 1989.

- Respectfully,

. COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR' POWER STATION a

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1. A. Robey Services' Superintendent RAR/vmk/djb Enclosure 8905250184 890430 DR ADOCK 05000254 i

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QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2-i MONTHLY PERFORMANCE REPORT APRIL, 1989 COMMONWEALTH EDIS0N COMPANY AND l

l IONA-ILLIN0IS GAS & ELECTRIC COMPANY NRC DOCKET N05. 50-254 AND 50-265 i

LICENSE N05. DPR-29.AND DPR-30 i

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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 feductions VI.

Unique Reporting Requirements A.

Main Steam Relief Valve Operations B.

Control Rod Drive Scram Timing Data VII.

Refueling Information VIII.

Glossary n

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INTRODUCTION' Quad-Cities Nuclear Power Station ~is composed of two Boiling Water

' Reactors, each with a Maximum Dependable Capacity of 769 MWe Net, located in 1

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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 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.

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5 II.

SUKHARY OF OPERATING EXPERIENCE A.

Unit One Unit One began the month of April operating at 720 MWe to shuffle control rods and perform the weekly turbine-generator surveillance.

At 1020 hours0.0118 days <br />0.283 hours <br />0.00169 weeks <br />3.8811e-4 months <br />, a load increase to full power was taken with control rods and recirculation pumps.

At 2025 hours0.0234 days <br />0.563 hours <br />0.00335 weeks <br />7.705125e-4 months <br />, all testing was completed and the unit was placed in Economic Generation Control (EGC). The unit remained in EGC until April 3.

At 0330 hour0.00382 days <br />0.0917 hours <br />5.456349e-4 weeks <br />1.25565e-4 months <br />s: EGC was tripped to perform the monthly High Pressure Coolant Injection (HPCI) surveillance.

At 0430 hours0.00498 days <br />0.119 hours <br />7.109788e-4 weeks <br />1.63615e-4 months <br />, the surveillance was completed, and the unit was placed in EGC.

At 0554 hours0.00641 days <br />0.154 hours <br />9.160053e-4 weeks <br />2.10797e-4 months <br />, the HPCI system was declared inoperable due to problems with being able to engage the turning gear.

At 0625 hours0.00723 days <br />0.174 hours <br />0.00103 weeks <br />2.378125e-4 months <br />, ECC was tripped and a load increase to 725 MWe was taken. At 0800 houro, the Chicago Load Dispatcher requested full load.

At 0855 hours0.0099 days <br />0.238 hours <br />0.00141 weeks <br />3.253275e-4 months <br />, the HPCI system was declared operable.

Power levels were held constant and on April 4 at 0025 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />, the unit was placed in EGC.

The unit remained in EGC until April 5.

At 0620 hours0.00718 days <br />0.172 hours <br />0.00103 weeks <br />2.3591e-4 months <br />, EGC was tripped and a load increase to full power was taken to perform Traversing In-Core Probe Set.

Power levels were held constant until April 7 due to Master Controller problems.

At 1440 hours0.0167 days <br />0.4 hours <br />0.00238 weeks <br />5.4792e-4 months <br />, power levels were adjusted and the unit was placed in EGC at the request of the Load Dispatcher.

The unit remained in EGC until April 9.

At 0754 hours0.00873 days <br />0.209 hours <br />0.00125 weeks <br />2.86897e-4 months <br />, a " Turbine Bypass Valve Open" alarm was received in the control room and at 0758 hours0.00877 days <br />0.211 hours <br />0.00125 weeks <br />2.88419e-4 months <br />, EGC was tripped to investigate the continuous problems with the master controller. At 0830 hours0.00961 days <br />0.231 hours <br />0.00137 weeks <br />3.15815e-4 months <br />, power levels were adjusted and per the request of the load dispatcher, the unit was placed in EGC.

The unit remained in EGC until April 10.

At 0735 hours0.00851 days <br />0.204 hours <br />0.00122 weeks <br />2.796675e-4 months <br />, EGC was tripped and a load increase to full power was taken per the request of the Load Dispatcher. At 1847 hours0.0214 days <br />0.513 hours <br />0.00305 weeks <br />7.027835e-4 months <br />, power levels were adjusted and the unit was placed in EGC.

On April 11 at 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />, EGC was tripped and a load increase to full power was taken at the request of the Load Dispatcher. At 0525 hours0.00608 days <br />0.146 hours <br />8.680556e-4 weeks <br />1.997625e-4 months <br />, the steam seal feed valve was isolated and left in the closed position due to a steam leak.

At 2221 hours0.0257 days <br />0.617 hours <br />0.00367 weeks <br />8.450905e-4 months <br />, a turbine bypass valve opened while the unit was operating at 750 MWe, EGC was tripped and recirculation pumps were placed in MANUAL. At 2233 hours0.0258 days <br />0.62 hours <br />0.00369 weeks <br />8.496565e-4 months <br />, the turbine bypass valve again opened. Load was reduced to 730 MWe with recirculation flow, and the valve closed. At 2349 hours0.0272 days <br />0.653 hours <br />0.00388 weeks <br />8.937945e-4 months <br />, the bypass valve again opened, so control rods were inserted and power was reduced to 635 MWe.

On April 12, Instrument Maintenance concluded that there was a problem with the bypass valve small close bias potentiometer.

An adjustment was made to increase the bias signal to the bypass valves, and a load increase to 700 MWe was taken.

As reactor power was increased, the

1. 1 bypass valve opened. Another adjustment was made to the potentiometer and the valve closed. At 0312 hours0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br />, the #1 bypass valve opened fully and the #2 bypass valve opened 50 percent.

Adjustments to the potentiometer were ineffective.

Continuous problems with the turbine bypass valves existed and at 1136 hours0.0131 days <br />0.316 hours <br />0.00188 weeks <br />4.32248e-4 months <br />, a manual reactor scram was inserted.

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The expected water level transient due to the collapse of the voids following the. scram caused reactor vessel level to drop below the +8 inches which caused Group II and III Primary Containment Isolations, Reactor Building Ventilation and Control Room Ventilation Isolations, and Standby Gas Treatment initiation.

Reactor water level was restored automatically by the Feedwater System and a normal scram recovery proceeded.

Instrument Maintenance began troubleshooting the Electro-Hydraulic Control system.

Between April 12 and April 15, a circuit board in the Electro-Hydraulic j

Control (EHC) system was replaced. A circuit board within the combined i

maximum flow limit circuit had a decreasing output.

The 1 ard limits f

the opening of control valves and as a result of_the decreasing output, j

caused the control valves to close. The bypass valves were opening as designed to control reactor pressure.

On April 15 at 1051 hours0.0122 days <br />0.292 hours <br />0.00174 weeks <br />3.999055e-4 months <br />, the reactor was made critical. At 1730 hours0.02 days <br />0.481 hours <br />0.00286 weeks <br />6.58265e-4 months <br />, the reactor was manually scrammed due to a steam leak discovered on the elbow of the continuous heat vent line which could not be isolated.

Repairs were made and at 1450 hours0.0168 days <br />0.403 hours <br />0.0024 weeks <br />5.51725e-4 months <br /> on April 16 the reactor was again made critical. Startup procedures were commenced..On April 17 at 0235 hcurs the mode switch was placed in RUN.

While performing testing of the electromatic relief valves, the 3D electromatic relief valve was discovered stuck in the open position.

Several attempts were made to close the valve with the keylock switch on panel 901-3 but these were unsuccessful. At 0330 hours0.00382 days <br />0.0917 hours <br />5.456349e-4 weeks <br />1.25565e-4 months <br />, the reactor was manually scrammed. At 0331, an Unusual Event per Emergency Action Level No. 14, Failure of Relief Valve to Reseat was initiated in accordance with the General Site Emergency Plan. At 0753 hours0.00872 days <br />0.209 hours <br />0.00125 weeks <br />2.865165e-4 months <br />, the unit reached cold shutdown and the unusual event was terminated. The unit remained shutdown while the electromatic relief valve and pilot valve were replaced.

On April 18 at 0830 hours0.00961 days <br />0.231 hours <br />0.00137 weeks <br />3.15815e-4 months <br />, startup procedures were commenced, and at 1157 hours0.0134 days <br />0.321 hours <br />0.00191 weeks <br />4.402385e-4 months <br />, the reactor was made critical. At 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br />, the mode switch as placed in RUN.

On April 19 at 0120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />, the main generator was synchronized to the grid. A load increase to 250 mew was taken and held constant to perform weekly turbine / generator surveillance. At 0530 hours0.00613 days <br />0.147 hours <br />8.763227e-4 weeks <br />2.01665e-4 months <br />, all testing was completed and an ascent to full load was begun, using control rods. At 1305 hours0.0151 days <br />0.363 hours <br />0.00216 weeks <br />4.965525e-4 months <br />, 820 MWe was achieved.

On April 20 at 1621 hours0.0188 days <br />0.45 hours <br />0.00268 weeks <br />6.167905e-4 months <br />, power levels were adjusted and the unit was placed in EGC. At 2341 hours0.0271 days <br />0.65 hours <br />0.00387 weeks <br />8.907505e-4 months <br />, EGC was tripped and a power reduction to 500 MWe was taken at the request of the Chicago Load Dispatcher.

On April 21 at 0515 hours0.00596 days <br />0.143 hours <br />8.515212e-4 weeks <br />1.959575e-4 months <br />, e load increase to full power was taken per the Load Dispatcher. At 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br />, 820 MWe was achieved. At 2050 hours0.0237 days <br />0.569 hours <br />0.00339 weeks <br />7.80025e-4 months <br />, a power reduction to 300 MWe was taken for a drywell entry. At 0030 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> on April 22, the load drop was completed.

.g At 0435' hours, an ascent to full power was begun with recirculation pumps

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and control rods. At 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br />, full power was achieved. On April 23 at 0045 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />, a power reduction to 776 We was taken at the request of the Load Dispatcher. At 1020 hours0.0118 days <br />0.283 hours <br />0.00169 weeks <br />3.8811e-4 months <br />, the unit was placed in EGC.

The unit remained in EGC or operated near full power until April 27.

At 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />, EGC was tripped due to the turbine control valves spiking 100 percent open. At 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, power levels were adjusted and the unit was placed in EGC.

On April 28 at 1010 hours0.0117 days <br />0.281 hours <br />0.00167 weeks <br />3.84305e-4 months <br />, the unit was taken off of.EGC, and a power reduction to 750 W e was tkaen so that Instrument Maintneance could perform testing on the Electro-Hydraulic Contr-1 (EHC) system.

On April 29 at 0025 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />, power levels were further rc. :ed to 450 We at the request of the Load Dispatcher.. At 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br />, another load reduction to 200 We was taken for an EHC system board replacement due to continuous turbine control valve problems.

On April 30 at 0105 hours0.00122 days <br />0.0292 hours <br />1.736111e-4 weeks <br />3.99525e-5 months <br />, the main generator was taken off-line to replace the control valve op amp circuit board. At 0341 hours0.00395 days <br />0.0947 hours <br />5.638227e-4 weeks <br />1.297505e-4 months <br />, the main generator was synchronized to the grid, and a load increase to 450 We was taken and held constant through the remainder of the month.

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Unit Two Unit Two began the month of April operating in Economic Generation Control (EGC). The unit remained in EGC until April 6 with minor interruptions to perform normal operational activities.

At 0332 hours0.00384 days <br />0.0922 hours <br />5.489418e-4 weeks <br />1.26326e-4 months <br /> on April 6, an unanticipated reactor scram occurred while performing weekly turbine / generator testing due to a master trip solenoid valve failure. The "A" pilot solenoid valve of the turbine master trip solenoid valve failed in the de-energized condition. Due to a stuck limit switch, the light indication continued to show the pilot solenoid valve energized. Thus, when the "B" master trip solenoid was tested, a turbine trip occurred. The failed solenoid was rebuilt and the coil and limit switch were replaced.

At 1445 hours0.0167 days <br />0.401 hours <br />0.00239 weeks <br />5.498225e-4 months <br /> on April 6, the reactor was made critical.

On April 7 at 0120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />, the main generator was synchronized to the grid. An ascent to full power was begun, using control rods. At 1855 hours0.0215 days <br />0.515 hours <br />0.00307 weeks <br />7.058275e-4 months <br />, 818 We was achieved. On April 8 at 1650 hours0.0191 days <br />0.458 hours <br />0.00273 weeks <br />6.27825e-4 months <br />, a power reduction to 750 We was taken and at 1855 hours0.0215 days <br />0.515 hours <br />0.00307 weeks <br />7.058275e-4 months <br /> the unit was placed in EGC.

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From April 8 through April 14, the unit remained in EGC or operated near j

full power with minor interruptions to perform normal operational activities j

and surveillance. On April 14 at 2031 hours0.0235 days <br />0.564 hours <br />0.00336 weeks <br />7.727955e-4 months <br />, EGC was tripped to perform l

the monthly control rod drive surveillance. On April 15 at 1631 hours0.0189 days <br />0.453 hours <br />0.0027 weeks <br />6.205955e-4 months <br />, power levels were adjusted, and the unit was placed in ECC.

The unit j

remained in EGC or operated near full power for normal operational activities until April 22.

At 2223 hours0.0257 days <br />0.618 hours <br />0.00368 weeks <br />8.458515e-4 months <br />, the unit was taken of EGC, and a power i

I reduction to 300 We was taken at the request of the Chicago Load Dispatcher.

A drywell entry was made to repair leaking valve packing. At 0730 hours0.00845 days <br />0.203 hours <br />0.00121 weeks <br />2.77765e-4 months <br /> on April 23, an ascent to full power was taken.

Full power was achieved at 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br />.

For the remainder of the month, normal operational activities occurred, with the unit operating near full power or remaining in ECC with minor q

interruptions to perform routine surveillance.

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

A.

Amendments to Facility Lietnse or Technical Specifications l

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 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 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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P UNIT 1 MAINTENANCE

SUMMARY

WORK REQUEST NO.:

Q64149 LER NUMBER:

87-026 COMPONENT:

System 1024 - FSAR noncompliance of piping support for line 1-1024A-20"-LX on drawing M-1604-40 Rev. B.

CAUSE OF MALFUNCTION: The cause of the problem was found to be on improperly set spring can.

RESULTS & EFFECTS ON SAFE OPERATION:

Safety of the plant and personnel were unaffected because, although the supports were outside FSAR compliance. They were still within operability compliance. RHR was therefore still considered operable.

ACTION TAKEN TO PREVENT REPETITION:

Corrective action was to reset spring can in accordance with support drawing M-1606-40 Rev. B.

WORK REQUEST NO.:

Q66065 LER NUMBER: NA COMPONENT _:_ System 1400 - One Spray Discharge Header Hi/Lo pressure instrument broke and was not able to be calibrated.

CAUSE OF MALFUNCTION: While performing a QIF-37 calibration, instrument 1-1467B broke while it was being calibrated.

It therefore would not be calibrated to within Tech Specs. The root cause of the pressure switch failure is unknown but speculated to be a buildup of dirt and other foreign material in the switch actuator.

RESULTS & EFFECTS ON SAFE OPERATION:

The safety implications of this event are minimal due to the monitoring program initiated immediately after the failure. At no time did the "B" Core Spray discharge pressure drop below the alarm setpoint of 46 psig.

Therefore the discharge piping remained filled as required by Tech Specs.

ACTION TAKEN TO PREVENT REPETITION: Work Request Q66065 was initiated to replace the switch. The switch was replaced with a Technically identical switch. Two failures at this switch were found at the Quad-Cities Utation in the past ten years and seven were found in an industry NPRDS search. Most of the failures were due to unknown causes.

1 WORK REQUEST NO.: Q69590 j

LER NUMBER: NA COMPONENT:

System 1400 - 1A Core Spray pump upper motor bearing oil cooler had leak allowing water to leak into oil.

CAUSE OF MALFUNCTION:

The 1A. Core Spray was declared inoperable due to water found in the pump motor upper bearing oil reservoir. The cause of the event was found to be a small leak in the sealing material at the fittings of the cooling coil.

RESULTS & EFFECTS ON SAFE OPERATION:

Safety significance was minimal because all other ECCS systems were operable as demonstrated by the completion of QOS 1400-01, Core Spray Subsystem Outage Report.

ACTION TAKEN TO PREVENT REPETITION: As a preventive action, the cooling coil is inspected during each refuel outage and operators check the oil level once per shift. Electrical Maintenance was writing a procedure to assist' in trouble-shooting such problems.

This procedure will have the cooling coil' pressurized prior to removal.

This assures that the sealing material is also tested.

WORK REQUEST NO.:

Q70310 LER NUMBER: NA COMPONENT:

System 1600 - Suppression Pool level recorder gives different reading than that of sight glass.

EPN LR-1-1602-7 CAUSE OF MALFUNCTION:

Suppression chamber level recorder LR-1-1602-7 was found to be reading +2.1 inches during a HPCI surveillance: After performing a suppression chamber level verification, it was determined that torus level was below + 2 inches. The apparent cause of the event was actual oscillations in torus level.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal because the HPCI surveillance was stopped when a high level reading was received.

In addition, since local indication reported a proper torus level, no safety problem occurred.

ACTION TAKEN TO PREVENT REPETITION: Work request Q70310 was written to recalibrates LR-1-1602-7 to agree with other torus level indications.

In addition, further operator training will be given to insure that operators are aware of torus level oscillations while performing the HPCI surveillance.

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WORK REQUEST No.: Q70404 LER NUMBER: NA COMPONENT:

System 2400 - 1B DW Rad Monitor failed upscale and downscale for no apparent reason.

EPN 1-2419B CAUSE OF MALFUNCTION: The 1B Drywell Radiation Monitor failed downscale and caused a 1/2 Group II isolation logice trip. The apparent cause of the event was unknown at the time of this report. The cause of the failure was to be known after the completion of Work Request Q70409.

RESULTS & EFFECTS ON SAFE OPERATION:

The safety consequences of this event were minimal.

Since the monitor was inoperable for only 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, no pre-planned alternate method of monitoring was required.

i ACTION TAKEN TO PREVENT REPETITION:

Immediate corrective action was to replace the monitor and to reset the Group II isolation.

Further corrective action was expected in a supplemental report, but it had not been completed at the time of this report.

WORK REQUEST NO.:

Q70581 LER NUMBER: NA COMPONENT:

System 1700 "B" MSL Rad-Monitor failed downscale.

CAUSE OF MALFUNCTION: The "B" MSL Rad Monitor failed downscale.

The alarm was reset.

It was found that one of the two Low Voltage Power supplies had l

failed. The monitor then failed again downscale, and a manual 1/2 scram was initiated. The apparent cause of the failure was an open capacitor in the LVPS which was the cause for the degraded power output. The MSL monitor itself was declared to be operable.

RESULTS 6 EFFECTS ON SAFE OPERATION:

The safety consequences of the event were minimal because all of the MSL rad monitors were functional. The rad monitor with the degraded LVPS had a redundant LVPS which kept the unit fully operable.

In addition, operators were able to operate the plant in a safe, stable condition throughout the event.

ACTION TAKEN TO PREVENT REPETITION: Work Request Q70581 was written to investigate and repair the "B" MSL rad monitor.

The monitor was replaced with a calibrated spare and it was functionally tested. Therefore the system was declared operable. A review of industry NPRDS data indicated no problem with capacitors.

Therefore this event is considered to be an isolated random failure.

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WORK REQUEST No.: Q70586 LER NUMBER: NA COMPONENT:

System 1700 "B" MSL Rad Monitor Power Failure CAUSE OF MALFUNCTION: The "B" MSL Rad Monitor failed downscale. The alarm was reset.

It was found that one of the two Low Voltage Power supplies had failed. The monitor then failed again downecale, and a manual 1/2 scram was initiated. The apparent cause of the failure was an open capacitor in the LVPS which was the cause for the degraded power outpet. The MSL monitor itself was declared to be operable.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of the event were minimal because all of the MSL rad monitors were functional. The rad monitor with the degraded LVPS had a redundant LVPS which kept the unit fully operable.

In addition, operators were able to operate the plant in a safe, stable condition throughout the event.

ACTION TAKEN TO PREVENT REPETITION: Work Request Q70586 was initiated to repair.the LVPS for the rad monitor. A bad capacitor was found and replaced.

A review of industry NPRDS data indicated no problem with capacitors. Therefore this event is considered to be an isolated random failure.

WORK REQUEST No.: Q70891 LER NUMBER: NA COMPONENT:

System 263 - Reactor Level Indicator 1-263-106A was replaced.

CAUSE OF MALFUNCTION: The "A" loop of the 2/3 Core Reactor water level monitor was out of service to perform a partial modification consisting of recalibration.

An LCO was entered because the "B" loop was also out for the modification.

The 1-263-106A could not be calibrated to the tolerances specified in the modification. The root cause of the event was inadequate planning. The partial mod could not be completed until both loops were successfully calibrated and tested.

RESULTS & EFFECTS ON SAFE OPERATION: The safety significance of this event was considered minimal.

During the time that the "A" loop could not be recalibrates the "B" loop was fully functional. Although the "B" loop could not be tested and declared operable, it was within tolerance and would have performed its required function.

t ACTION TAKEN TO PREVENT REPETITION:

Immediate action was to repair the 1-263-106A indicator. After failure to repair the indicator, the station decided to perform an incomplete mod.

This allowed the "B" loop to be tested separately and to be returned to service. Work Request Q70891 was written to repair the "A" loop indicator.

i WORK REQUEST No.: Q72403 LER NUMBER:

NA COMPONENT:

System 1000 - Valve 1-1001-34B valve would not open.

"B" loop inoperable.

CAUSE OF MALFUNCTION: Upon completion of surveillance testing in preparation for taking the shared Diesel Generator out of service, the RHR MOV-1-1001-34B lost its light indication in the control room when the valve went fully closed.

An Equipment Attendant was sent to MLC 19-4 and found the breaker tripped.

Three attempts were made to reopen the valve, all resulting in tripping the breaker. The cause of the event was component failure due to sheared motor tie bolts.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences of the event were miniaml because the 1-1001-34B valve failed in its safe, normally closed position.

Therefore, primary containment was isolated.

The valve could have been opened manually if it had been needed.

Core Spray and the RHR "A" loop were both operational.

ACTION TAKEN TO PREVENT REPETITION: The MOV was replaced under Work Request Q72425. A procedure had been written to limit the number of times a breaker can be reset and operated after a trip. This problem was an isolated event.

WORK REQUEST No.:

Q73551 LER NUMBER: NA C0HPONENT:

System 1700 - Repair MSL Rad Monitor from 1-1705-2C and return it to 1-1705-2C after repair. EPN 1-1705-2C CAUSE OF MALFUNCTION: A fault in the "C" Main Steam Line Rad Monitor caused a Channel "A" 1/2 Scram and a 1/2 Group I isolation channel trip. An IM Technician found the main chassis fuse blown.

It was determined that the cause was a faulty power supply, which blew the fuse and prevented the monitor display from coming on.

RESULTS & EFFECTS ON SAFE OPERATION: The safety consequences were minimal because the other three MSL Rad Monitors were functional to perform this primary function. The blown fuse resulted in a failure in the conservative direction.

ACTION TAKEN TO PREVENT REPETITION: Work Request Q73541 was written to investigate and repair the "C" MSL Rad Monitor.

Immediate repair was replacement with a calibrated spare. Work Request Q73551 was written to repair the original rad monitor.

Five failures of the power supplies were experienced at the Quad Cities Station, and some failed LVPS's have been submitted to GE for analysis.

t WORK REQUEST NO.: Q7156/

LER NUMBER:

NA COMPONENT:

System 1000 - 1B RHRSW pump inoperable due to low flow.

CAUSE OF MALFUNCTION: During an RHR pump operability surveillance, it was found that the IB RHRSW pump only pumped 2800 gpm @ 301 psig discharge. This valve was 300 gpm less than the established reference valve. The pump was then declared inoperable.

Work Request Q70743 was written to inspect the i

pump.

A small amount of debris was found in the suction of the pump. However, it is not believed that the debris could have caused the pump degradation.

The root cause of the problem was not found at the time of this report.

RESULTS & EFFECTS ON SAFE OPERATION: The safety significance was minimal.

All other components on containment cooling were operable per Tech Specs.

l ACTION TAKEN TO PREVENT REPETITION: Work Request Q70743 was written to remove the debris in the pump. Work Request Q71567 was written to repair the pump.

Further corrective action was dependent on the results of a supplemental investigation.

WORK REQUEST NO.: Q72327 LER NUMBER: NA COMPONENT:

System 1000 - MOV-1-1001-36A lost its light indication and tripped thermals.

t CAUSE OF MALFUNCTION: While performing the RHR containment cooling valve operability test, the 1-1001-36A MOV lost its light indication in the control room panel while being stroked.

It was found that the thermal overload relay for the valve had tripped. The "A" RHR loop was declared inoperable.

Initial l

cause was thought to be undersized thermal overload heaters.

Further investigation found that the main contactors were not fully energized during operation, j

causing an extend heating of the overload heaters.

1 RESULTS & EFFECTS ON SAFE OPERATION:

Safety consequences of the event were minimal because the valve could have been stroked by hand.

In addition the upstream valve MOV-1001-34A was fully operable, therefore the primary containment 1

was maintained.

The "B" loop was also available for use during the event.

ACTION TAKEN TO PREVENT REPETITION:

Immediate action was to insure that the valve was closed. Work Request Q72327 was initiated to upsize the thermal heaters.

In addition, the auxiliary contacts were cleaned and lubricated to prevent binding. As a safety precaution, all other MOV's are being checked for proper thermal heater sizing.

1 l

l

__________J

^

j 1

WORK REQUEST NO.:

Q73717 LER NUMBER:

NA COMPONENT:

System 1700 - Sping on 912-4 indicated failure of service water rad monitor.

Sample line was found plugged.

CAUSE OF MALFUNCTION: A Liquid Process Radiation Monitor Failure alarm was

. received in the Control Room.

In addition, a " FAIL" light on the SPING terminal was received.

The Unit One Service Water Radiation M<:nitor was declared inoperable after a lack of flow was discovered in the sample line. The cause of the event was the plugging of the sample line to the Service Water Rad Monitor Flow switch.

RESULTS & EFFECTS ON SAFE OPERATION:

The safety consequences of this event are minimal due to the Radiation Chemistry Departments routine analysis on the service water return header. Grab samples are pulled every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, meeting the Tech Specs requirements for an inoperable Service Water Radiction Monitor.

ACTION TAKEN TO PREVENT REPETITION:

Immediate action was to remove the Rad Monitor from service. A flush was performed on the sample line, which was plugged with mud. The rad monitor was taken out of service under Work Request Q73717. As a follow-up corrective action, a special test was to be written to look into upgrading the supporting piping of the Rad Monitor, i

i I

j

li 3.'

4 UNIT 2 MAINTENANCE

SUMMARY

V l

WORK REQUEST NO.:

Q67628 LER NUMBER: NA COMPONENT:

System 730 - Tip Machine #3 did not fully withdraw during PCI Group II and III test.

EPN 2-0730-3 CAUSE OF MALFUNCTION: The #3 TIP machine would not withdraw its detector during PCI Group 2 isolation testing. The cause of the event was attributed to a blown fuse in TIP machine #3 that prevented the TIP Detector from being withdrawn from the reactor core.

The inability to withdrawal the detector prevented the closing of the TIP ball valve.

RESULTS & EFFECTS ON SAFE OPERATION:

Safety consequences of this event were minimal because the key lock shear valve located downstream of the ball valve was fully functional.

In the event that a Group II Isolation occurred with the failed fuse in place, the the TIP machine #3 detector in the core, the instrument line could have been isolated by actuating the shear valve.

ACTION TAKEN TO PREVENT REPETITION: Work Request Q67628 was initiated to investigate the problem. The brown fuse was replaced, but there were no other problems found. The Station deemed that no further corrective action

.was required.

WORK REQUEST NO.:

Q67675 LER NUMBER:

88-023 COMPONENT:

System 7800 - Testing auto transfer for MCC 28/29-5 CAUSE OF MALFUNCTION: During a modification test for M-4-2-88-06A, motor control center MCC 28/29-5 would not automatically transfer from the Bus 29 to the Bus 28 feed. A wire was found not landed per the approved electrical drawing. The cause was believed to be installation error during original plant construction.

RESULTS & EFFECTS ON SAFE OPERATION:

The auto transfer function of MCC 28/29-5 is needed in the event of a LOCA concurrent with a loss of offsite power and a failure of the Unit 2 Diesel.

In this event, a failure would result

{

in an inability to supply power to the RHR injection valves. However, MCC j

28/29-5 can still be changed manually.

j ACTION TAKEN TO PREVENT REPETITION: Work Request Q67675 was initiated to land the cable number 22373 to terminal point E-76 in the 902-8 panel.

A visual inspection was done on the Unit One wiring to insure that the same problem did not exist in the MCC 18/19-5 transfer logic. An action plan for reviewing untested components for failure was also initiated.

u WORK REQUEST NO.: Q69316, Q69317 LER NUMBER:

88-013 COMPONENT:

System 5700 - Replaced fan belts on 2A and 2B CS Room Cooler.

EPN 2-5748-A and 2-5748-B CAUSE OF MALFUNCTION: While performing normal operating rounds per procedure QOS 005-S14, The EA found the Unit One "A" Core Spray room cooler was off and was unable to be started manually.

Inspection showed one belt to be broken and the other to be off the pulley. The cause of the event was found to be insufficient preventative maintenance.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plant and operating personnel was not affected during this event. The Unit One HPCI and LPCI systems were successfully tested after finding the 1A Core Spray room cooler inoperable. Therefore, all other ECC. systems including the "B" loop of Core Spray were operable throughout the event.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to re-install the loose fan belt and operate the room cooler temporarily on one fan belt. Work requests have been written for both Units 1 and 2 to replace room coolers on the HPCI, RHR, and Core Spray Systems. The fan belts will be replaced every refuel outage. Work Requests Q69316 and Q69317 were written to replace the U-2 2A and 2B CS room cooler belts.

WORK REQUEST No.: Q69320 LER NUMBER:

88-013 COMPONENT:

System 5700 - Replaced fan belts on HPCI Room Cooler. EPN 2-5747.

CAUSE OF MALFUNCTION: While performing normal operating rounds per procedure QOS 005-S14, the EA found the Units One "A" Core Spray room cooler was off and was unable to be started manually.

Inspection showed one belt to be broken and the other to be off the pulley. The cause of the event was found to be insufficient preventative maintenance.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plant and operating personnel was not affected during this event. The Unit Onn WPCI and LPCI l

systems were successfully tested after finding the 1A Core Spray room cooler inoperable. Therefore, all other ECC systems including the 'B" loop of Core Spray were operable throughout the event.

l ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to re-install the loose fan belt and operate the room cooler temporarily on one fan belt. Work requests have been written for both Units 1 and 2 to replace room coolers on the HPCI, RHR, and Core Spray Systems. The f an belts will be replaced every refuel outage. Work Request Q69320 was written to replace the U-2 HPCI room cooler belts.

i

e WORK REQUEST NO.:

Q69323, Q69324 LER NUMBER:

88-013 COMPONENT:

System 5700 - Replaced fan belts on 2A and 2B RHR Room Cooler.

EPN 2-5746-A and 2-5746-B CAUSE OF MALFUNCTION:

While performing normal operating rounds per procedure QOS 005-S14, the EA found the Units One "A" Core Spray room cooler was off and was unable to be started manually.

Inspection showed one belt to be broken and the other to be off the pulley. The cause of the event was found to be insufficient preventative maintenance.

RESULTS & EFFECTS ON SAFE OPERATION: The safety of the plar.t and operating personnel was not affected during this event. The Unit One HPCI and LPCI systems were successfully tested after finding the 1A Core Spray room cooler inoperable. Therefore, all other ECC systems including the "B" loop of Core Spray were operable throughout the event.

ACTION TAKEN TO PREVENT REPETITION: The immediate corrective action was to re-instal?. the loose fan belt and operate the room cooler temporarily on one fan belt. Work requests have been written for both Units 1 and 2 to replace room coolers on the HPCI, RHR, and Core Spray Systems. The fan belts will be replaced every refuel outage. Work Requests Q69323 and Q69324 were written to replace the 2A and 2B RHR room cooler belts.

l l

1

- - A

1 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 I

l 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 89-003 4-12-89 Manual Scram, EHC problems89-004 4-17-89 Manual Scram, Stuck open on Relief valve 89-005 4-17-89 Control Room Energency Air Filtration Unit inoperable UNIT 2 89-001 4-06-89 Turbine Trip - Reactor Scram while testing Turbine Master trip solenoid i

i I

l 0027H/0061Z l

L.

4 1

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 I

0027H/00612

.o*

s APPENDIX C OPERATING DATA REPORT OOCKET NO, 50-254 UNIT One M"I 4' l989 DATE COMPLETED SY Lynne Deelsnyder l

TELEPHONg 309-654-2241 OPERATING STATUS 0000 040189 2400 043089

1. REPORTING PERICO:

GROSE HOURS IN REPORTING PERICO:

2511 MAX. OEPENO. CAPActTY tamme.sese 769

2. CURRENTLY AUTM0Rl200 POWER LEVEL (gY DE880N ELECTRICAL RATING leAWs Net):

N/A

3. POWER LEVEL TO WHICH RETRICTED (IP ANY) (HWo Neil:

& REA0008 POR RESTRICTION llP ANY):

TMit MONTM YR TO DATE CUBAULATfvE 594.0 2754 120296.2

5. NURAGER OF MOURE REACTOR WAS CRITICAL...............

0.0 0.0 3421.9 E. REACTOR RESERVE sMUTOOWN MOURs...................

558.7 2718.7 116377.9

7. MOURE GENERATOR ON UNE..........................

0.0 0.0 909.2 E. UNIT RS$8RVE SMUT 00WN MOURS..........,...........

1251955 6354331 248044410

s. GROEs THERMAL ENERGY GENERATED (MWM) 406635 2073635 80431248

10. GROSS ELECTRICAL ENERGY GENER ATED (MWM)....

I2%

11. NET ELECTRICAL ENERGY GENERATED (WWM1 82.6 95.7 80.9 12.

EACTOR SERVICE P ACTOR.....,...............

82.6 95.7 83.2_

13. REACTOR AV AILAstLITY P ACTOR........................

77.7 94.4 78.2

14. UNIT SERVICE P ACTOR........................

77.7 94.4 78.8

19. UNIT AVAILABILITY P ACTOR......................

.0

18. UNIT CAPACITY P ACTQm lueing MOCl.....................

17. UNIT CAPACITY PACTOR (Ussag Demon MWel................

22.3 5.6 5.4

18. UNIT PORCED OUTAGE RATE...................

19. SHUTDOWNS SCHEDULEO OVER NEXT e MONTHS ITYPE. DATE. AND DURATION OF EACHl:

20. IP SMUT 00WN AT END OF REPORT PERIOO. ESTIMATED DATE OF STARTUP:

21. UNITS IN TEST STATUS (PRIOR TO COMMERCIAL OPERATION):

PORECAST ACMfEVED INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERAT1000 1.1H

e APPENOlX C OPERATING DATA REPORT DOCKET NO.

50-265 UNIT Tun DATE May 4, 1989 COMPLETED gy Lynne Deelsnyder l

TELEPHONg 309-654-2241 OPERATING STAM 0000 040189 719 2400 043089 GROBE Hours IN REPORTl88G PER100:

3, REPORT 1880 PER800:

2. CURRENTLY AUTHORISED POWER LEVEL Ig0Yh. 2511 max. oGPENO. CAPACITY (ARe>80sd. ' 769 0884088 ELECTRICAL RATING (Ants.Ned:

N/A

3. POWER LEVEL TO WMcCH RWTRICTED llP ANY1 (GNWu.Ned:

4. REABONE FOR RESTRICTION (IP ANYh THIE M088TN YRTO DATE CUMULAftVE 707.8 2810.7 113760.6

5. NUMSER OF HOURE REACTOR WAE CRITICAL...............

0.0 0.0 2985.8

8. REACTOR RESERVE $NUTOOWN MOURE...................

697.2 2791.8 110523.5

7. HOURE GENERATOR ON LINE.......................

0.0 0.0 702.9 E. UNIT RESERVE SMUTDOWN MOURS,...................... 1590552 6423104 237333377

9. GROSE THERMAL ENERGY GENERATED IMWM) 519257 2103021 76036492

10. GROBE ELECTRICAL ENERGY GENER ATED (MWM)...........

497324 2015816 71752393

11. NET ELECTRICAL ENERGY GENERATED (MWH1.............

'I 98.4 97.6 76.9

12. EACTOR SERVICE P ACTOR.......................

98.4 97.6 7 8. 9 __

12. REACTOR AV AILABILITY P ACTOR.......................

97.0 97.0 74.7

14. UNIT SERVICE P ACTOR..............................

97.0 97.0 75.2

15. UNIT AVAILAEI LITY P ACTOR..........................

89.9 91.1 63.1

18. UNIT CAPACITY P ACTOR (Using MOCl....................,

87.7 88.7 M.s

17. UNIT CAPACITY PACTOR (UnaE Design MWH.................

3.0 3.0 8.3

18. UNIT PORCED OUTAGE RATE...........

1E. SHUTDOWN 8 SCHEDULE 0 CVER NEXT E MONTHS ITYPE. DATE. AND DURATION OF EACMh

20. IF SHUT DOWN AT END OF REPORT PERIOD. ESTIMATED DATE OP STARTUP:

21. UNITS IN TEST STATUS (PRIOR TO COMMERCIAL OPERATIONh PORECAST ACHIEVED I

INITIAL CRITICALITY 1

INITIAL ELECTRICITY i

COMMERCIAL OPERAT1000 i

1.1M 4

L ;.*

APPENDIX 8 I f, AVERAGE DAILY UNIT POWER LEVEL l

DOCKET NO.

50-254 UNIT one DATE 5-5-89 COMPLETED BY Lvnne Deelsnyder TELEPHONE 309-654-2241 MONTH April, 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 727 10 1

gy 707 16 2

3, 744 527 3

19 722 744 4

20 758 666 5

21 796 634 8

gg 739 709 y

gg 715 719 g

g 728 722 3

5 10*

703 717 3

763 735 11 27 289 724 12 3

13

- 11 541 3

- 10' 282 14 g

- 11 15 31-

- II 16 INSTRUCTIONS On this form, list the aversy daily unit power level in MWe Net for each day in the reporting month. Compute to the nessest whole megawatt.

These figures will be used to plot a graph for cach reporting month. Note that when muximum dependable capacityis

- ugd for the net electrical rating of the unit, there may be occasions whesi the daily averap power level exceeds the

~

1001 line (or the restrwted power level line). In such cases, the average daily unit power output sheet should be footnoted to explaus the apparent anomaly.

l.16 4 O

O O

g*

r-g I

i.i -

APPENDIX B -

• o n

L..-

' AVERAGE DAILY UNIT POWER LEVEL

i DOCKET NO.

50-265 UNIT Two L. -

DATE 5-5-89 COMPLETED BY tynne Deelsnyder TELEPHONE 309-654-2241

' MONTH-April, 1989 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL.

(MWe-Net)

(MWe-Net) 1 716 747 37 685 723

.2 gg 707 727 3

jg 714 704 4

f 712

.722 5

88 706

.S gg 578 588 7

3 63 747 S-24 722 756 g

3

~~

10-733 g

712 11 745 710 27 748 706 12 g

714-672 93 g

720 684 14 30 787 15 -

39 708 18 INSTRUCTIONS On this form, list the averap daily unit power level in MWe. Net for each day in the reporting month. Compute to the peassst whole megawatt.

These figures will be used to plot a graph for each reporting month. Note that when munimum dependable capacityis ugd for the net electrical rating of the unit, there may be occasions when the daily average power level exceeds the 800'A line (or the restricted power lewt line). In such cases, the average daily unit power output sheet should be footnoted to explam the apparent anomaly.

1.164

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VI. UNIQUE REPORTING RE0VIREMENTS The following items are included in this report based on prior commitments to the commission:

i l

l A.

Main Steam Relief Valve Operations l

Relief valve operations during the reporting period are summarized in the following table. The table includes information as to which l

relief valve was actuated, how it was actuated, and the circumstances i

resultng in ita actuation.

Unit: One Date: April 17, 1989 Valves Actuated No. 6 Type of Actuation 1-203-3A 1 Manual 1-203-3B 1 Manual 1-203-3C 1 Manual 1-203-3D 1 Manual Plant Conditions: Reactor Pressure - 921 psig Description of Events: Technical Specification 4.5.D.I.a Unit:

One Date: April 18, 1989 Valves Actuated No. & Type of Actuation 1-203-3B 1 Manual 1-203-3D 1 Manual 1-20.:-3E 1 Manual Plant Conditions: Reactor Pressure - 920 psig Description of Events: Technical Specification 4.5.D.l.a B.

Control Rod Drive Scram Timing Data For Units One and Two There was no Control Rod Drive Scram Timing Data for Units One and Two for the reporting period.

0027H/0061Z

r 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

.r QTP 300-S32

=

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 thereaf ter require a technical specification change or other license amendment:

l NOT AS YET DETERMINED.

5.

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

JUNE 10, 1989 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:

1773 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:

q I

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 WL' ps F8 ft () )/ Et E)

, AP8 2 01978 Q.C.O.S.R.

f 4

QTP 300-S32 Revision 1 j

QUAD-CITIES REFUELING March 1978 INFORMATION REQUEST

~

1.

Unit:

02 Reload:

9 Cycle:

10 2.

Scheduled date for next refueIIng shutdown:

2-3-90

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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 IIcense amendment:

J NOT AS YET DETERMINED.

5.

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

NOVEMBER 2, 1990 6.

Important IIcensing 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 i

b.

Number of assemblies in spent fuel pool:

1475 8.

The present IIconsed 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 Ilconsed storage:

0 9

The projected date of the last refueling that can be discharged to the spent fuel rw I assuming the present IIconsed capacity: 2008 XPPROVED, APR 2 o1978 Q.C.O.S.R.

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VIII. GLOSSARY l-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

-Average Power Range Monitor APRM Anticipated. Transient Without Scram ATHS I

BWR Bolling Water Reactor CRD Control Rod Drive EHC Electro-Hydraulic Control Syst9m 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 Integrated Primary Containment Leak Rate Test IPCLRT IRM Intermediate Range Monitor Inservice Inspection ISI Licensee Event Report LER 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 Institute for Occupational Safety and Health PCI Primary Containment Isolation PCIOMR Preconditioning Interim Operating Management Recommendations RBCCW 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 RWM Rod Worth Minimizer l

SBGTS Standby Gas Treatment System SBLC Standby Liquid Control SDC Shutdown Cooling Mode of RHRS SDV Scram Discharge Volume SRM Source Range Monitor j

TBCCW Turbine Building Closed Cooling Water System

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TIP Traversing Incore Probe 1

TSC Technical Support Center

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0027H/0061Z

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