ML20073B565
| ML20073B565 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/31/1994 |
| From: | Hovey R, Lyons D Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9409220100 | |
| Download: ML20073B565 (11) | |
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P O PSEG Public Service Electnc and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 s
Hopo Creek Generating Station September 15, 1994 U.
S.
Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
Dear Sir:
MONTHLY OPERATING REPORT HOPE CREEK GENERATION 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 August are being forwarded to you with the summary of changes, tests, and experiments that were implemented during August 1994 pursuant to the requirements of 10CFR50.59(b).
Sincerely yours, f
.J.
R. J. Hovey General Manager -
Hope Creek Operations
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The Energy People
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'I INDEX l
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NUMBER SECTION OF PAGES Average Daily Unit Power Level.
1 Operating Data Report 3
Refueling Information.
1 Monthly operating Summary.
1 Summary of Changes, Tests, and Experiments.
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OPERATING DATA REPORT DOCKET NO.
50-354 UNIT Hope Creek DATE 09/13/94.
i COMPLETED BY D.
W.
Lyons i
TELEPHONE (609) 339-35 7 t
OPERATING STATUS 1.
Reporting Period Avaust 1994 Gross Hours in Report Period 744 2.
Currently Authorized Power Level (MWt) 3293 Max. Depend. Capacity (MWe-Net) 1031 Design Electrical Rating (MWe-Net) 1067 3
Power Level to which restricted (if any) (MWe-Net)
None 4.
Reasons for restriction (if any)
This Yr To Month Date Cumulative 5.
No. of hours reactor was critical 655.0 4390.3 57213.3 6.
Reactor reserve shutdown hours 222 212 222 7.
Hours generator on line 643.0 4286.9 56319.4 8.
Unit reserve shutdown hours 222 0.0 0.0-9.
Gross thermal energy generated 2057303 13702415 179665785 (MWH)
- 10. Gross electrical energy 676900 4532690 59496644 generated (MWH)
- 11. Net electrical energy generated 644605 4321656 56849340
(
(MWH)
- 12. Reactor service factor 88.0 75.3 84,8
- 13. Reactor availability factor 88,0 75.3 84.8
- 14. Unit service factor 86,4 73.5 83.5
- 15. Unit' availability factor 86.4 73 5 83 5
- 16. Unit capacity factor (using MDC) 84.0 71.9 81.7
- 17. Unit capacity factor 81,2 69.5 78.9 (Using Design MWe)
- 15. Unit forced outage rate 13.6 129 4.6
- 19. Shutdowns scheduled over next 6 months (type, date, &= duration):
None
- 20. If shutdown at end of report period, estimated date of start-up:
N/A i
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l OPERATING DATA REPORT-UNIT SHUTDOWNS AND POWER REDUCTIONS f
DOCKET NO.
50-354 UNIT Hope Creek j
DATE 09/12/94 i
COMPLETED BY D.
W.
Lyons TELEPHONE (609) 339-3517 MONTH Aucust 1994 1
METHOD OF l
SHUTTING DOWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE NO.
DATE S= SCHEDULED (HOURS)
(1)
POWER (2)
ACTION / COMMENTS.
1 1
8/1 F
63 A
3 AUTOMATIC SCRAM DURING IRM SURVEILLANCE DUE TO FAULTY TEST EQUIPMENT.
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2 8/30 F
38 H*
3 UNIT STILL SHUT DOWN AT END OF REPORTING PERIOD.
t (SYNCHRONIZATION WAS
'ON 09/01/94)
AUTOMATIC SCRAM CAUSED BY GENERATOR RUNBACK DUE TO LOSS OF STATOR WATER COOLING.
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- Reason is still i
under review. Will update next period j
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l AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
5.0-354 UNIT Hope Creek
('-
DATE 09/12/94 COMPLETED BY D.
W.
Lyons TELEPHONE (609) 339-3517 MONTH Aucust 19Ei DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 1.
EEE 17.
1032 2.
Q 18.
1029 3.
p 19.
1030 4.
11E 20.
1025 5.
111 21.
1022 6.
1034 22.
1034 7.
1041 23.
1043 8.
1043 24.
1R13 9.
1035 25.
1212 10.
1036 26.
1040 l
11.
1036 27.
1035 12.
1037 28.
1009 13.
1015 29.
1030 14.
1023 30.
ARE 15.
1036 31.
2 16.
1Rlg
REFUELING INFORMATION DOCKET NO.
50-354 UNIT Hope Creek 1 DATE Sept 13.1994 COMPLETED BY R.
Schmidt TELEPHONE (609) 339-3740 MONTH Auaust 1221 1.
Refueling information has changed from last month:
Yes No X 2.
Scheduled date for next refueling:
9/16/95 3.
Scheduled date for restart following refueling:
10/31/95 4.
A.
Will Technical Specification changes or other license amendments be required?
Yes No X
B.
Has the Safety Evaluation covering the COLR been reviewed by the Station Operating Review Committee?
Yes No X
If no, when is it scheduled?
Auqust 28, 1995 5.
Scheduled date(s) for submitting proposed licensing action:
H2% aghgduled.
6.
Important licensing considerations associated with refueling:
ULb 7.
Number of Fuel Assemblies:
A.
Incore 764 B.
In Spent Fuel Storage (prior to refueling) 1240 C.
In Spent Fuel Storage (after refueling) 1472 8.
Present licensed spent fuel storage capacity:
4006 Future spent fuel storage capacity:
4006 9.
Date of last refueling that can be discharged 5/3/2006 to spent fuel pool assuming the present (EOC13) licensed capacity:
(Qqng allow for full-core offload)
(Assumes 244 bundle reloads every 18 months until then)
(Does Dgt allow for smaller reloads due to improved fuel)
HOPE CREEK GENERATING STATION MONTHLY OPERATING
SUMMARY
August 1994 Hope Creek entered the month of August at 100% reactor power.
The Unit shutdown at 20:55 on August 1 1994 caused by-a. trip during a surveillance test due to faulty test-equipment..The unit returned to operating status on August 4, 1994.
It continued to operate at full power from August 5th through August.30th when the Reactor' tripped at 0944 following a loss of stator water cooling and remained shutdown for the remainder of the period.
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SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION August 1994 The following items 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.
The 10CFR50.59 Safety Evaluations showed that these items did not create a new safety hazard to the plant nor did they affect the safe shutdown of the reactor.
These items did not change the plant effluent releases and did not alter the existing environmental impact.
The 10CFR50.59 Safety Evaluations determined that no unreviewed safety or environmental questions are involved.
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'DCE Summary gi Safety Evaluation AME-0066:
This Design Change replaces existing PVC voice and data caeles with plenum rated voice and data cables.
The cables being removed and replaced are non-safety related.
Based on a review of the UFSAR, the Fire Hazard Analysis Tabuletion tables in Appendix 9A are required to be revised as a result of this modification to reflect the cable insulation quantity added to the rooms.
The net increase in fire loading due to the addition of cables does not exceed the rating of fire barriers and is still within the design basis for the affected fire areas.
Therefore, this modification does not change the facility as described in the SAR.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
4HE-00067 Eks 11 This Design Change will build a new Hazardous Material Storage (HAZMAT) Facility inside the security fence east of the Central Warehouse to store hazardous material on site.
This facility will be controlled with existing procedures governing chemical and environmental controls.
This HAZMAT facility is not a primary plant system and does not directly interface with any safety related or radiological system.
It presents no increase in design basis threat to safety related systems or design features, and is therefore classified as none safety related and non-seismic.
As such the facilities are not required to mitigate the conseguences of a design basis transient or accident condition.
The primary function of the new building is to provide a storage area for hazardous material within the protected area boundary.
Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question.
Temporary Modifications Summary of Safety Evaluation T-Mod 94-021:
This Temporary Modification installed an electrical jumper across the #2 Feedwater Heater Hi-Hi Level trip switches and installed a temporary keep fill line to the low side of the level transmitters.
This modification is performed due to spurious indications during power ascension and is removed at approximately 40 % Reactor Power.
This T-Mod does not increase the probability or the consequences of an accident listed in Table 15.0-2 of the UFSAR since the worst case would be for water induction into the turbine resulting in a turbine trip.
Therefore, this Temporary Modification does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
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T-Mod 94-022:
This Temporary Modification installs an electrical jumper on "A" service Water Traveling Water Screen (TWS) Spray Wash Booster Pump flow switch.
This jumper will remain in place until a replacement flow switch can be procured and installed.
The "B" Service Water Pump also has the same modification (T-Mod 94-019) installed while the remaining pumps are all operable using the original system design flow switch.
Should there be one active failure of the TWS due to insufficient spray flow any debris will be carried over into the strainer where it will be removed via the backwash strainer.
Control room indications will alert operators to any failure to sufficiently clean the travelling screen by CRIDS points which will indicate increased differential pressure across the service water strainer.
Therefore, this Temporary Modification does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
QB Summary of Safety Evaluation RB HIC 94-097:
This Deficiency Report dispositioned a Drywell Temperature sensor at 162' elevation. reading a localized hot spot (thermal) of 247'F to 250'F as Use-As-Is.
SAR section 9.4.5.1 states that during normal operation the average air temperature in the drpell shall be limited to 135'F with no location above 150*F (localized hot spots above 162' elevation with 194*F hot spots are permitted).
Investigation during a forced outage found a missing piece of insulation from the RPV Package in the area which was replaced.
Additionally another piece of insulation was found below the RTD prohibiting upward ventilation flow in the annulus area which was removed prior to startup.
When unit returned to service temperature readings returned to normal range.
The localized hot spot experienced was detected by a non safety related temperature element in the drywell.
This component is not needed for the safe shutdown of the plant nor is it required to be operable following a design basis accident.
Evaluation of drywell temperatures indicates that the hot spot was localized and the temperatures did not affect any safety related components or structures.
Therefore, this Deficiency Report does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
f-e Other Egmmary 21 Safety Evaluation H-1-ZZ-MSE-0837:
This is an editorial change to the Hope Creek SAR Table 7.1-2 page 2 of 3.
It will correct the revision date of for Reg Guide 1.29 and bring it in line with Sections 8.1.4.4 and 1.8.1.29 of the Hope Creek UFSAR.
Therefore, this UFSAR change does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
H-1-EA-MSE-0829:
This Safety Evaluation addresses the impact of increasing the maximum allowable ultimate heat sink (UHS) temperature from 85'F to 88.6*F and, increasing the Service Water System design temperature from 85'F to 95'F for supply piping and 110*F to 120*F for discharge piping.
The original design analysis for the Service Water System is based on a maximum inlet water temperature of 85'F at the suction of the Service Water Pumps.
Subsequent iterative analysis was performed to establish the new limit of 88.6*F (SEE LCR 93-05).
Detailed engineering evaluations demonstrate that the proposed revisions to the service water temperature will not significantly reduce the ability of the Service Water System to provide sufficient cooling for safety related equipment and coolers during normal and accident conditions.
As such, this does not reduce the margin of safety as defined in the basis for any Technical Specification.
Therefore, this Safety Evaluation change does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
EAR Chance Notice 93-22 Rev 1:
This revision to the SAR changes the implementation of the Mechanical Eguipment Qualification (MEQ)
Program.
Active safety related mechanical equipment will be analyzed under Reliability Centered Maintenance Program for the appropriate inspection, testing, or maintenance program instead of fixed frequency based material evaluation.
This change affects the method of implementing as described in SAR Section 3.11.2.6.
However there is no procedure in the SAR which describes implementation of the MEQ Program.
Therefore this changes does not change the procedure as described in the SAR.
Therefore, this SAR Change Notice does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.
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