ML20150B965
| ML20150B965 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 02/29/1988 |
| From: | Jensen H, Labruna S Public Service Enterprise Group |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8803170210 | |
| Download: ML20150B965 (16) | |
Text
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AVERAGE DAILY UNIT POWER LEVEL DOCKET No.86-354 UNIT Hope Creek DATE 3/15/88 COMPLETED BY H.
Jensen TELEPHONE (609) 339-5261 MONTH February 1988 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (HWe-Net)
(MWe-Net) 1 1005 17 0
2 1011 18 0
3 1018 19 0
4 1021 20 0
i 5
1014 21 0
6 1010 22 0
7 1011 23 0
8 1014 24 0
9 992 25 0
j 10 996 26 0
l 11 993 27 0
12 914 28 0
13 20 29 0
14 0
15 0
16 0
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S 8803170210 800229 DR ADOCK 0500 4
OPERATING DATA REPORT DOCKET No.86-354 UNIT Hope Creek DATE 3/15/88 COMPLETED BY H.
Jensen TELEPHONE (609) 339-5261 a
OPERATING STATUS 1.
REPORTING PERIOD Feb 1988 GROSS HOURS IN REPORTING PERIOD __f96 2.
CURRENTLY AUTHORIZED POWER LEVEL (HWt) 3293 HAX. DEPEND. CAPACITY (HWe-Net) 1067 *
(MWe-Gross) 1118 DESIGN ELECTRICAL RATING (HWe-Net) 1067 3.
POWER LEVEL TO WHICH RESTRICTED (IF ANY) (HWe-Net)
None 4.
REASONS FOR RESTRICTION (IF ANY)
THIS YR TO HONTH DATE CUMULATIVE 5.
NO. OF HOURS REACTOR WAS CRITICAL 301.0 1045.0 8903.1 6.
REACTOR RESERVE SHUTDOWN HOURS 0
0 0
7.
HOURS GENERATOR ON LINE 293.9 1037.9 8783.0 8.
UNIT RESERVE SHUTDOWN HOURS 0
0 0
9.
GROSS THERHAL ENERGY GENERATED (HWH) 937,997 3,378,284 27,186,852 10.
GROSS ELECTRICAL ENERGY GENERATED (HWH) 314,464 1,133,985 9.045,682 11.
NET ELECTRICAL ENERGY GENERATED (HWH) 296,858 1,085,075 8,650,113 i
12.
REACTOR SERVICE FACTOR 43.2 72.6 84.9 13.
REACTOR AVAILABILITY FACTOR 43.2 72.6 84.9 14.
UNIT SERVICE FACTOR 42.2 72.1 83.7 15.
UNIT AVAILABILITY FACTOR 42.2 72.1 83.7 16.
UNIT CAPACITY FACTOR (Usinc Design MDC) 40.0 70.6 77.3 17.
UNIT CAPACITY FACTOR (Using Design HWe) 40.0 70.6 77.3 18.
UNIT FORCED OUTAGE RATE 0.0 0.0 8.0 A9.
SHUTDOWNS SCHEDULED OVER NEXT 6 HONTHS (TYPE. DATE. & DURATION):
None 20.
IF SHUT DOWN AT END OF REPORT PERIOD. ESTIMATED DATE OF STARTUP:
4/8/88
- Data obtained in August 1987 is under management review.
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.I, OPERATING DATA REPORT UNIT SMUTDOWNS AND POWER REDUCTIONS i
DOCKET NO.86-354 i
UNIT Moce Creek i
DATE 3/15/88 f
COMPLETED BY R. Ritzman REPORT HONTH Feb. 1988 TELEPHONE (609) 339-3737 I
METMOD OF SHUTTING DOWN THE i
T'iPE REACTOR OR F 1.'OKCED DURATIO!f REASON REDUCING CORRECTIVE ACTION /
NO.
DATE S 3CHEDULED-(HOURS)
(1)
POWER (2)
COMMENTS 1
2/13 S
402.1 C
2 REFUEL OUTAGE I
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i SUNNARY
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l HOPE CREEK GENERATING STATION HONTHLY OPERATING
SUMMARY
(
FEBRUARY 1988 i
l Hope Creek entered the month of February operating at j
e,pproximately 100%
power.
The unit did not experience any shutdowns or reportable power reductions until it was taken off-line for refueling on February 13 at 5:55 am.
At that
- time,
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the plant had completed its 64th day of continuous power
[
operation.
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SUMMARY
OF CHANGES, TESTO, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION FEBRUARY 1988 l
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The following Design change Packages (DCPs) have been evaluated to determine:
1) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety i
previously evaluated in the safety analysis report may be increased; or i
2) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or a
3) if the margin of safety as defined in the basis for any technical d
specification is reduced.
lione of the DCPs created e new safety hazard to the plant nor did they affect the safe shutdown oi the reactor.
These DCPs did not change I
the plant effluent releases and did not alter the existing l
environmental impact.
The Safety Evaluations determined that no i
unreviewed safety or environmental questions are involved.
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QQE Description of_Desian Chance Packacs 4-EMC-86-0650 This DCP added support to floor penetration seals which had sepsrated from their. sleeve.
These supports resolved seismic II/I concerns.
4-HMM-86-1152 This DCP made several modifications to the Solid Radwaste system.
It _ changed the slope of the Centrifuge Metering Pump Discharge Lines to keep the flow meters filled in order to meter flow more accurately.
It also removed Decant Screens from the Centrifuge Feed Tank, increased the size of a
nozzle on the Centrifuge Feed Tank to improve system performance, and installed tubing and a
vent to accommodate a 3-way manual valve in a low radiation area.
4-HMM-86-1282 This DCP relocated the Reactor Auxiliaries Cooling System Water Supply and Return piping for the Emergency Air Compressor from downstream of the Containment Isolation Valves to upstream of the valves.
This will improve the reliability of the Emergency Instrument Air Compressor start-up during Loss of Offsite Power.
4EC-1030/02 This DCP installed monorails, lifting lugs, and rigging points for Torus Access Hatch covers.
This will facilitate removal of the hatch covers and increase cost effectiveness.
4EC-1055/03 This DCP instelled a Motor Control Center, Control Console, and Sequence Controller to support the installation of the Semi-Automatic Control Rod Drive Removal and Installation Equipment.
When complete this DCP will greatly reduce radiation exposure associated with Cuatrol Rod Drive changeout.
4EC-1082/09 This DCP revised the wiring to the Reactor Protection System "A"
and "B"
Out of Service Switches to provide the appropriate indication.
These switches provide indication only.
This discrepancy was identified during the Control Room Design Review process and is part of a commitment to improve human factors in the Control Room.
4HC-0054 This DCP provided a shielded Fuel Transport Chute and lifting equipment for use in the reactor cavity during refuel outaces.
The transfer chute J
is a temporary portable shielding device that is installed prior to the transfer of irradiated fuel bundles from the reactor to the spent fuel storage pool.
The fuel bundles are passed through the transfer chute to reduce radiation levels in the upper drywell area, allowing continuous personnel access.
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l QqE Description of-Desian-Chance Packace 4-HC-0125/01 This DCP installed the. ' mechanical equipment. and-piping required to support' the-installation.of a
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third Reactor Auxiliaries : Cooling System. Pump.'
When the related instrument and controls DCP (4HC-0125/02) isi-installed, the. additional. pump will improve operating, reliability of the Reactor Auxiliaries cooling. System.and - prevent forced shut-down of the plant.
The. additional pump will not increase the' system pressure beyond itsidesign-pressure.
The' pump foundation and related ' piping; are' seismically supported and will maintain their integrityLduring a seismic event.
4-MM-0047 This DCP-changed the
.5 amp fuses on the Data.
Driver Cards with-1 amp fuses.
This change. will' increase the. availability of thefReactor Manual Control.
System.
The
. change will prevent unnecessarily blowing the fuse to.the transmitter' cards in the Rod _ Select
- Module, while still providing adequate overload protection.
4HM-0066 This DCP changed out Asco fixed deadband switches on the Emergency Diesel Generators with ASCO adjustable deadband switches.
This change out-l will result in a substantial cose-savings over the life of the plant.
-The adjustable
'deadband switches maintain the necessary accuracy.and are consistent with the vendors original design.
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4HM-0136 This DCP increased the Standby Liquid' Control Tank Sodium Pentaborate. solution concentration and lowered the high level and low level slarm setpoints in the Standby Liquid Control Tank.
These changes implement a Technical-Specification amendment and ensure. compliance with 10CFR50.62.
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4HH-0191 This DCP modified the-impingement plate at Steam Jet Air Ejector After Condensor'Second Stage Steam Inlet Nozzles.
The previous design resulted in tube failure due to the plate breaking away from the After Condenser Shell.
This DCP will increase the strength at the impingement baffle to prevent similar failures.
4HH-0204 This DCP removed the cast steel plug in the upper inspection port of the Moisture Separttor in the.
"A" and "B"
loops and replaced it with a carbon i
steel plug and seal weld.
This modification will improve the integrity of the Steam Jet Air-Ejector-system by providing absolute sealing of the i
Moisture Separators upper inspection port area.
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ECE Descriotion of Desian Chance rackaoe 4HM-0246 This DCP moved the Main Hoist Normal Up Stop on the Refueling Bridge up 8
inches to provide adequate clearance for fuel bundles in tranrit across the fuel transfer chute.
Raising the Up Stop by 8 inches does not exceed maximum allowable height, does not affect the bundle drop accident
- analysis, and maintains adequate radiation shielding.
4HM-0263 This DCP raises the "A"
and "B" Steam Jet Air Ejector Steam Inlet High Pressure Alarm setpointn.
The setpoint changes will eliminate spurious alarm indications caused by the previous.
setpoints proximity to the normal operating range.
It will also provide the Control Room with a more accurate indication of Steam Jet Air Ejector trouble.
4HM-0291 This DCP recalibrated the core flow summer based on a new gain adjustment factor.
The core flow indication was higher than actual core flow.
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J The following Temporary Modification Requests (THRs) have been evaluated to determine:
1) if the. probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or
- 2) if a possibility for an accident or malfunction of a different ty;9 than any evaluated.previously in the safety analysis report may be created; or 3) if the margin of safety as defined in the basis for any technical specification is reduced.
None of the THRs created a new safety hazard to the plant nor did they affect the safe shutdown of the reactor.
These THRs did not change the plant effluent releases and did
.not alter 'the existing environmental impact.
The Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
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Safety Evaluation Descriotion of Temporary Hodification Recuest (TMR1 87-0087 This THR adds a Portable Arrowhead Flexmate Oil Removal System to the Chilled Water System.
Oil removal will be accomplished through two Charcoal Filter Demineralizers and one Mixed Bed Demineralizer.
Total failure of the supply and return hoses will induce a
1 1/4" diameter equivalent leak or loss of that quantity of coolant from the Chilled Water System.
This would have little or no impact since this closed cooling system has a
4" diameter demineralized water make-up line to its head tank, thus having a
greater capacity than the possible leak.
- Also, the 4" diameter floor drain piping system would allow adequate removal of water from the local area without causing floodina.
88-0009 This THR connects a temporary power load center to a non-lE 480 volt AC unit substation.
The load center will be supplied with 480 volt, 600 amp power by temporary cables connected to a
spare breaker and will be used to support the Main Generator Retaining Ring work during the refuel outage.
This installation was done in accordance with applicable standards and procedures.
The addition of the temporary load center to the spare breaker does not exceed the capacity of the associated transformer.
88-0024 This TMR provides a 480 volt Non-lE temporary power source to a 125 volt DC Class lE battery charaer durina the "C"
4.16 KV Class 1E bus outage.
This modification is to be used only when "C" Channel is inoperable and the plant is in operational conditions 4,5, and Therefore, this THR is for convenience and is not being relied on to provide any safety function.
88-0026 This THR provides a 480 volt Non-1E temporary power source to a 125 volt DC Class lE battery charger durina the "C"
4.16 Kv Class 1E bus outage.
This modification is to be used only when "C" Channel is inoperable and the plant is in operational conditions 4,5, and Therefore, this THR is for convenience and is not being relied on to provide any safety function.
88-0029 This THR provides a 480 volt Non-1E temporary power source to a 120 volt AC Class lE security inverter durina the "C"
4.16 Kv Class lE bus outage.
This modification is to be used only when "C" Channel is inoperable and the plant is in operational conditions 4,5, and Therefore, this THR is for convenience and is not beina relied on to provide any safety function.
Eafety Evaluation Degyriotion of Temocrary Modification Recuest (THR) 88-0032-This THR provides a 480 volt Non-lE temporary power source to a 480 volt AC Non-1B Motor Control Center during the "C" 4.16 KV Class lE bus outage.
This modification is to be used only when "C"
Channel is inoperable and the plant is in operational conditions 4,5, and Therefore, this TMR is for convenience and is not being relied on to provide any safety function.
88-0033 This THR provides a 120 volt AC temporary source to a 24 volt DC battery charger during the "C"
4.16 kV Class lE bus outage.
This modification is to be used only when "C" Channel is inoperable and the plant is in operational conditions 4,5, and *.
Therefore, this THR is for convenience and is not being relied on to provide any safety function.
88-0037 The Reactor Vessel Water Level Transmitter normally provides control room indication on a
scale of 0-400 inches (relative to vessel zero).
This scale is normally satisfactory as the top of the vessel is at 372 inches.
However, when the cavity is fully
- flooded, the level is at 492 inches.
This TMR provides a new transmitter and indicator scale of 0-550 inches.
The new transmitter is qualified to the same standards as the previous one and enhances the control room indication.
88-0043 This THR provides a temporary power source to the Reactor Protection System "A" bus during the 4.16 kv switchgear maintenance outage.
This THR shall only be in place during operationa: conditions 4,
5, and
- and shall be removed when normal power to the Reactor Protection System "A"
bus is available.
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The following Deficiency Requests (DRs) have been evaluated to determine:
1) if.the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased; or 2) if a possibility for an accident or malfunction of a.different type than any evaluated previously in the safety analysis report may be created; or 3) if the margin of safety as defined in the basis for any technical specification is reduced.
None of the DRs created a new safety hazard to the plant nor did they affect the safe shutdown of the reactor.
These DRs did not change the plant effluent releases and did not alter the existing environmental impact.
The Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
Safety Evaluation Description of Deficiency Report (DR) 88-0014 The results of a dye penetrant check performed on a Reactor Building Nitrogen Supply Valve were not properly documented.
This valve was purchased as an ASME N-stamped component, but is being used in a non-ASME system.
In the event of a failure of the valve
- disc, the valve could be isolated.
Therefore, this the valve could be isolated.
Therefore, this DR may be dispositioned "use-as-is".
88-0017 A Code Job Package Liquid Penetrant Test was performed on a valve in the Reactor Core Isolation Cooling Turbine Steam system.
At the time that this test was performed, there was no block on the form for recording the valves surface temperature.
- However, the technician verified that the temperature was within the required
- range, therefore, this valve may be "used-as-is".
88-0018 A Code Job Package Liquid Penetrant Test was performed on a
check valve in the Core Spray system.
At the time that this test was performed, there was no block on the form for recording the valves surface temperature.
- However, the technician verified that the temperature was within the required range, therefore, this valve may be "used-as-is".
88-0019 A Code Job Package Liquid Penetrant Test was performed on a
check valve in the Core Spray system.
At the time that this test was performed, there was no block on the form for recording the valves surface temperature.
- However, the technician verified that the temperature was within the required range, therefore, this valve may be "used-ac-is".
88-0041 During operational conditions, one of the Hydraulic Control Units exhibited frequent High Level Alarms.
These alarms were caused by water leakina past tae piston seals into the Nitrogen side of the accumulator.
During disassembly of the Hydraulic Control Unit, it was discovered that the leaks could be attributed to blemishes in the cylinder wall at the bottom end cap "O"
ring seating surface.
The Hydraulic Control Unit may be "used-as-is" because both Hich Water Level and Low Nitrogen Pressure provide an alarm.
The alarm can be cleared by draining the water and/or recharging the Nitrogen in the accumulator.
Safety Evaluation Description of Deficiency Report (DR) 88-0042 During operational conditions, one of the Hydraulic Control Units exhibited frequent High Level Alarms.
These alarms were caused by water leaking past the piston seals into the Nitrogen side of.the accumulator.
Durino disassembly of the Hydtaulic Control Unit, it was discovered that the leaks could be attributed to blemishes in the cylinder wall at the bottom end cap "O"
ring seating surface.
The Hydraulic Control Unit may be "used-as-is" because both High Water Level and Low Nitrogen Pressure provide an alarm.
The alarm can be cleared by draining the water and/or recharging the Nitrogen in the accumulator.
88-0044 The proper baseline data was not taken during the pre-operational testing of the Solid Radwaste Slurry Metering Pumps.
This missing data was baseline data only and did not affect acceptance criteria.
Additionally, the missing data deals with pump speeds significantly faster than the Slurry Hetering Pumps' operating speeds.
88-0045 While drilling a hole through the disc nut and bolt of a check valve in the Feedwater system the drill broke, damaging a thread on the bolt.
The remainder of the bolt is long enough for satisfactory engagement and may be "used-as-is".
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O PSEG Pubhc Service Electric and Gas Company P.O. Box L Hancocks Bridge, New Jerse,08038 Hope Creek Operations March 15, 1988 U.
S. Nuclear Regulatory Commission Document Control Desk Washington, DC.
20555
Dear Sir:
MONTHLY OPERATING REPORT HOPE CREEK GENERATING STATION UNIT 1 DOCKET NO. 50-354 In compliance with Section 6.9, Reporting Requirements for the Hope Creek Technical Specifications, the operating statistics for February are being forwarded to you.
In addition, the summary of changes, tests, and experiments for February 1988 are included pursuant to the requirements of 10CFR50.59(b).
Sincerely yours, S.
LaBruna General Manager -
Hope Creek Operations RAR:tib Attachment C Distribution SR(;
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The Energy People 95-2173 ti t Vi t 2 85