ML20235P125
| ML20235P125 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 06/30/1987 |
| From: | Jensen H, Labruna S Public Service Enterprise Group |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8707200348 | |
| Download: ML20235P125 (13) | |
Text
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.S AVERAGE. DAILY' UNIT POWER LEVEL'86-354
. DOCKET NO.
l UNIT Hope Creek Gen. Station DATE 87/07/87 H. Jensen COMPLETED BY
~
(609) 339-5261 TELEPHONE i
une, 1987 NONTH DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL
'(MWe-Net)
(MWe-Net) 927 914 1
'A.
17 911 1940 2
i 18 884 859 9
865 20 945 975 21 853 6
22 895 23 972 1964 p4 865 1 57 9
25 939 26 932 1910 11 27 940 28 13 29 927 30 937 15 31 934
.16 8707DOO340 B70630 j'4 k
DR ADDCK 0500
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OPERATING DATA REPORT DOCKET NO.
R6-194 UNIT Hnpa craak Gen. Sta.
DATE R7 /07 /R7 h
COMPLETED BY M-
.Tannan TELEPHONE (6091190-;?61 OPERATING STATUS 1.
REPORTING PERIOD June 1987 GROSS HOURS IN REPORTING PERIOD 720 2.
CURRENTLY AUTHORIZED POWER LEVEL (MWt) 3293 MAX. DEPEND. CAPACITY (MWe-Net)__
1067*
DESIGN ELECTRICAL RATING (MWe-Net) _
1867 3.
POWER LEVEL TO WHICH RESTRICTED (IF ANY) (MWe-Net)_
Mnna 4.
REASONS FOR RESTRICTION (IF ANY)
THIS YR TO MONTH DATE CUMULATIVE 5.
NO. OF HOURS REACTOR WAS CRITICAL 720 4177 (460 6.
REACTOR RESERVE SHUTDOWN HOURS 0
0 0
7.
HOURS GENERATOR ON LINE 7 2 H___,__ _414 7.1 4435.1 8.
UNIT RESERVE SHUTDOWN HOURS A
n n
9.
GROSS THERMAL ENERGY GENERATED (MWH)
?_?12.3?A 19.410.71n la tal_11R 10.
GROSS ELECTRICAL ENERGY GENERATED (MWH) 727.666 4.156.724 4_?;4.502 11.
NET ELECTRICAL ENERGY GENERATED (MWH) 644.764 3.964,512 4 ?E4 502 12.
REACTOR SERVICE FACTOR 100 06 1 QA 1 13.
REACTOR AVAILABILITY FACTOR 10A QA 1 QA 1 14.
UNIT SERVICE FACTOR 1An on E on A 15.
UNIT AVAILABILITY FACTOR 10A on R 05;R 16.
UNIT CAPACITY FACTOR (Using Design MDC)_
40,7 R6 A 86;4-.
f 17.
UNIT CAPACITY FACTOR (Using Design MWe) on 7 AA n Ag _A 18.
UNIT FORCED OUTAGE RATE A
a.4 A_3 j
SCHEDULED 19.
SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, & DURATION): SURVEILLAUCb 9/18/87, 15 days j
20.
IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:
i o
Extimated Value.
Actual value to be determined in July or August 1987.
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OPERATING DATA REPORT dNIT' SHUTDOWNS AND POWER REDUCTIONS86-354 DOCKET NO.
UNIT'NAME Hope Creek Gen.'Sta.;
-DATE 07/07/87" COMPLETED BY H. Jensen gj/
REPORT MONTH TELEPHONE
'(609)339-5261 June 1987 METHOD OF SHUTTING DOWN THE TYPE REACTOR OR F FORCED DURATION REASON REDUCING CORRECTIVE ACTION /
N0; DATE S SCHEDULED (HOURS)
(1)
POWER (2)
COMMENTS None i
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SUMMARY
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SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS i
i FOR THE HOPE CREEK GENERATING STATION i
JUNE 1987 l
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Description of Temporary Modification Safetv_ Evaluation Reguest (TMR) 87-0041 The Temporary Modification Request (TMR) associated with this Safety Evaluation added a temporary, portable sample pump to drain the North Plant Vent effluent through a
Particulate Filter and an Iodine Filter.
This would allow the normal analyses to be made.
Continued release through the North Plant Vent would also be permitted.
Temporarily replacing the out-of-service North Plant Vent Sample Pump with the portable sample pump did not create a new safety hazard to the plant ncr did it affect the safe shutdown of the reactor.
This TMR did not change the plant effluent releases 5
and did not alter the existing environmental impact.
Therefore, no unreviewed safety or environmental questions are involved.
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87-0043 The Temporary Modification Request (TMR) l associated with this Safety Evaluation 1
jumpered out the Thermal overload j
Protection for some Service Water Valves J
until the required bypass circuitry is installed.
The bypass circuitry of the thermal overloads is to ensure that the equipment operates in a safe manner under i
accident conditions.
This temporary modification ensures that the valves will operate as intended and did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.
j This TMR did not change the plant effluent releases and did not alter the l
existing environmental impact. Therefore, I
no unreviewed safety or environmental j
questions are involved.
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.s Safety' Evaluation Description 21 ' Temporary Modification-Request (TMR) 87-0044 The Temporary Modification Request -(TMR)-
associated 'with this SafetyL Evaluation jumpered out the Thermal:
Overload-Protection' for,some, Safety
. Auxil'iary Cooling. System until the requiredLbypass circuitry.-is installed.
The bypass
. circuitry of.the thermal overloads 11s'.to>
ensure that the equipment. operates in a
safe-manner under accident' conditions.
This temporary modification ensures.that:
,the valves operate as-intended and. did-not create.a new safety hazard-to the plant nor did it affect the safe shutdown of the reactor.
This:TMR did'not change the. plant effluent releases and did not'
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alter the existing environmental impact.
Therefore, no unreviewed safety.
or environmental. questions are involved.
87-0048-The Temporary Modification' Request (TMR) associated with this' Safety Evaluation evaluated plant-operation with the~
"B"
. Reactor.
Pressure-Regulator in control.
This evaluation was. satisfactory and did not create a new safety hazard to the plant nor did.it affect the safe. shutdown of the reactor..This TMR did not change the piant effluent releases and did not alter the existing environmental - impact.
Therefore, no unreviewed safety or environmental questions-are involved.
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DCE Description 2f DARISD Change Package 4-EMC-86-0960 This DCP provided a compensated watt-hour feed to the load dispatcher in Newark instead. of
'the previous uncompensated' watt-hour feed.
l Providing the compensated watt-hour feed did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.
This DCP did not change the plant effluent releases and did not alter the existing environmental impact.
Therefore, no unreviewed safety or environmental questions are involved.
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Safety Evaluation Description of Deficiency. Report IDEl 87-0018 Ninety Leak Detection System Temperature Elements were
- cleaned, lubricated, and retorqued.
The Temperature Element covers were to have been torqued to 50 foot-pounds;
- however, the torque wrench was determined to be out of calibration.
The actual torque value was 46.8 foot-pounds.
This torque value is still within design specifications and did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.
The deficiency did not change the plant effluent releases and did not alter the exicting environmental impact.
Therefore, no unreviewed safety or environmental questions are involved.
87-0042 Flow transmitters in the Filtration, Recirculation, and Ventilation systers have a tolerance of 0.00125" w.c..
The Measuring and Test Equipment Laboratory did not have equipment available to calibrate the transmitters to this tolerance.
They were calibrated to a
tolerance of.002" W.C..
The maximum accuracy error that this actual tolerance would allow is insignificant and well within Technical Specification limits.
EAR's were issued to correct the calibration tolerances on the ICD cards for these and other similar flow transmitters.
This deficiency did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.
It did not change the plant effluent releases and did not alter the existing environmental impact.
Therefore, no unreviewed safety or environmental questions are involved.
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l-l Safety Evaluation Descripti2n gf Deficiency BRp2It (DR)
.87-0045 General Electric issued a notification that the installed Turbine Control Valve Fast Closure Pressure Switches performing Reactor Safety Functions were not manufactured to GE's latest specification for o-ring and-back-up ring materials.
The installed switches have a
reliable performance history and undergo surveillance tests monthly.
Continued operation with these o-rings and back-up rings until the next scheduled outage will not compromise the ability of the pressure switches to perform their designed function.
This deficiency did' not create a new safety hazard to the pla " nor did it affect the safe shutdown of t.,e reactor.
The deficiency did not change the plant effluent releases and did not alter the existing environmental impact.
Therefore, no unreviewed safety or environmental questions are involved.
87-0049 The Extruder Evaporator Vent Hood Exhaust Fans were reported to be in the "slightly rough" region during vibration testing.
The manufacturer considers the results of l
the tests to be satisfactory; therefore, the vibration did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.
The deficiency did not change the plant effluent releases and did not alter the l
existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.
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j Safety Evaluation Description of Deficiency Renort (DR)
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l 87-0051 The South Plant Vent Wide Range Gas Monitor contains a
Gamma Checksource Assembly that allows for periodic verification of detector operability.
The Checksource Assembly contains a lead block to provide shielding.
The ' lead block dropped down onto the checksource, I
damaging the checksource and preventing the checksource from testing the l
detection.
The checksource tests are performed daily.
The problem was discovered and corrected.
The Safety l-Evaluation upgrades the Checksource l
Assembly to prevent future occurrences.
Since this component is a monitor and not process
- control, the deficiency did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.
The deficiency did not change l,
the plant effluent releases and did not alter the existing environmental impact.
Therefore, no unreviewed safety or environmental questions are involved.
87-0054 A
Steel gasket in a strainer in the Keepwarm Pump Suction Line of an Emergency Diesel Generator required replacement.
The gasket was not listed on the vendor's parts list and therefore, no spare parts were available.
A copper gasket was installed until the correct part can be obtained and installed.
The temporary installation of the copper gasket did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.
The deficiency did not change the plant effluent releases and did not alter the existing environmental impact.
Therefore, no unreviewed safety or environmental questions are involved.
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Safety Evaluation Description 91 Deficiengy.E222It (DR) 87-0055 A
Thrust Monitoring Device used to monitor motor-operated valves was tested independently of its associated mainframe.
The components should be tested together.
- However, since they were both within calibration parameters the monitoring results are valid and the independent testing did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.
The deficiency did not change the plant effluent releases and did not alter the existing environmental impact.
Therefore, no unreviewed safety or environmental questions are involved.
87-0056 An Emergency Diesel Generator Combustion Air High Temperature Switch failed to reset within its specified 12 F
range.
(The 12' limit was exceeded by 1.2 F).
Since the sole function of the switch is to provide a low priority annunciation it did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.
The deficiency did not change the plant effluent releases and did not alter the existing environmental impact.
Therefore, no unreviewed safety or environmental questions are involved.
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BOPE CREEK GENERATING STATION MONTHLY OPERATING
SUMMARY
. JUNE 1987 The Hope Creek Unit entered the month operating at approximately 90%
power with this output being. administrative 1y limited by the Electrical System Operator based'on electrical system stability considerations.
Varying limits are'placed on the output of the Bope Creek and-Salem Generating Stations as a result of the loss of.the Hope Creek /Keeney 500KV transmission line.
(See the Monthly Operating. Summary,. March 1987, for additional details.)
The. unit operated throughout the. month and did not experience any reportable shutdowns or power reductions.
On June 30, the unit completed its 122nd day of continuous power operation.
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'O PSLEG Public Service Electric and Gas Company P.O. Box L Hancocks Bridge, New Jersey 08038 Hope Creek Operations l
July 15, 1987 l
i U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 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 June are being forwarded to you.
In addition, the summary of
- changes, tests, and experiments for June 1987 are included pursuant to the requirements of 10CFR50.59(b).
Sincerely yours, cW<-r S,
LaBruna General Manager -
Hope Creek Operations Attachment C Distribution The Energy People 95 2173 411M; 5-67
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