ML20046D405
| ML20046D405 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 07/31/1993 |
| From: | Hover R, Zabielski V Public Service Enterprise Group |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9308190193 | |
| Download: ML20046D405 (12) | |
Text
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O PSEG i
Put*c Sennce Electic and Gas Company P.O. Box 236 Hancocks Bridge, fJew Jersey 08038 i
Hope Creek Generating Station 4
August 13,1993 4
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 i
In compliance with Section 6.9, Reporting Requirements for the j
Hope Creek Technical Specifications, the operating
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statistics for July are being forwarded to you with the f
summary of changes, tests, and experiments that were implemented during July 1993 pursuant to the requirements of 10CFR50.59(b).
l Sincerely yours, R. J. Hovey l
General Manager -
Hope Creek Operations DR:WS:JC Attachments
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C Distribution 1
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i The Energy People l-9308190193 930731 iT PDR ADOCK 05000354 jf g a
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INDEX NUMBER i
SECTIOE OF PAGES Average Daily Unit Power Level.
1 l
Operating Data Report 3
Refueling Information.
1 Monthly Operating Summary.
1 Summary of Changes, Tests, and Experiments.
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t OPERATING DATA REPORT DOCKET NO.
50-354 UNIT Hooe Creek DATE 8/13/93 COMPLETED BY V.
Zabielski TELEPHONE (609) 339-3506 OPERATING STATUS 1.
Reporting Period July 1993 Gross Hours in Report Period 211 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 l
4.
Reasons for restriction (if any)
Month Date Cumulative 5.
No. of hours reactor was critical 744.0 5006.0 49261.6 6.
Reactor reserve shutdown hours 02 212 0.0 1
7.
Hours generator on line 744.0 4986.9 48491.8 8.
Unit reserve shutdown hours 220 220 222 9.
Gross thermal energy generated 2448421 16203348 154416566 (MWH)
- 10. Gross electrical energy 794440 5398800 51146854 generated (MWH)
- 11. Net electrical energy generated 759331 5167119 48869503 (MWH)
- 12. Reactor service factor 100.0 98.4 85.0
- 13. Reactor availability factor 100.0 98.4 85.0
- 14. Unit service factor 100.0 98.0 83.6
- 15. Unit availability factor 100.0 98.0 83.6
- 16. Unit capacity factor (using MDC) 2222 98.5 81.7
- 17. Unit capacity factor 1122 95.2 79.0 (Using Design MWe)
- 18. Unit forced outage rate 922 2x2
_111
- 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 1
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i OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO.
50-354 UNIT Hope Creek DATE 8/13/93 COMPLETED BY V.
Zabielski TELEPHONE (609) 339-3506 MONTH July 1993 METHOD OF SHUTTING I
DOWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE NO.
DATE S= SCHEDULED (HOURS)
(1)
POWER (2)
ACTION / COMMENTS 0
0 none t
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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
50-354 UNIT HoDe Creek i
DATE 8/13/93 COMPLETED BY V.
Zabielski TELEPHONE (609) 339-3506 MONTH July 1993 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
-(MWe-Net) 1.
1006 17.
1033 l
2.
1018 18.
1019 3.
1030 19.
1008 4.
1016 20.
1021 5.
1041 21.
1043 6.
1009 22.
1034 7.
221 23.
1032 8.
1007 24.
1041 9.
1027 25.
1024 10.
1016 26.
1011 11.
1006 27.
1023 12.
1016 28.
1013 13.
1018 29.
1020 l
14.
1009 30.
1024 i
.15.
1017 31.
1033 16.
1032 4
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REFUELING INFORMATION DOCKET NO.
50-354 UNIT Hope Creek 1 DATE Auaust
.3.1993 COMPLETED BY S.
Holl:.nasworth TELEPHONE (609) 339-1051 MONTH July 1993 1.
Refueling information has changed from last month:
Yes No X
.cheduled date for next refueling:
3/5/94 3.
Scheduled date for restart following refueling:
4/23/94 4.
A.
Will Technical Specification changes or other license amendments be required?
i Yes No X
j 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?
2/18/94 5.
Scheduled date(s) for submitting proposed licensing action:
Hgt ggheduled yet.
6.
Important licensing considerations associated with refueling:
N/A 1
7.
Number of Fuel Assemblies:
l A.
Incore 764 B.
In Spent Fuel Storage (pricr to refueling) 1008 i
C.
In Spent Fuel Storage (after refueling) 1240 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:
(Does allow for full-core offload)
(Assumes 244 bundle reloads every 18 months until then)
(Does D2h allow for smaller reloads due to improved fuel) i i
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HOPE CREEK GENERATING STATION MONTHLY OPERATING
SUMMARY
July 1993 Hope Creek entered the month of July at approximately 100% power.
i The unit operated throughout the month without experiencing any t
shutdowns or reportable power reductions.
As of July 31, the plant has been on line for 73 consecutive days.
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SUMMARY
OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION July 1993 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 way be increased; or 2.
If a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis
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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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a QCE Descriotion gi Safety Evaluation 4EC-3348 Presently seal water is provided to the radwaste pumps PKG-15 and instruments from the condensate system.
- However, pressure varies due to demand on the condensate system.
This sometimes creates excessive flow and adds load to the Liquid Waste System and increases the condensate makeup requirements.
This DCP installs an integral regulator and controller to the seal water lines to the following Radwaste Pumps; A/B Waste Evaporator Concentrate Waste Transfer Pump, A/B Waste Evaporator Recycle Pump, A/B Concentrated Waste Pump, A/B Waste Neutralizer Pump and the Decontamination Concentrata Waste Pump.
This will supply constant seal water flow to the affected equipment.
This modification will not increase the probability of an accident previously evaluated in the SAR because it is being designed and installed to the current design standards.
Indicators / controllers are designed to fail open, maintaining flow to the seals.
Therefore, this DCP does not involve any Unreviewed Safety Questions.
4EC-3407 This DCP is being performed to satisfy NRC Bulletin PKG 1 93-03, resolutions of issues related to reactor vessel level instrumentation in BWR's.
This DCP will install supports, route tubing trays, r
install valves, route tubing and drill and repair penetrations in the Reactor Building elev. 102', 132' and 145'.
This DCP Pkg does not make any connections to the operating plant systems.
The intent of this Package is to perform the non-outage work to lessen the amount of work required during a plant shutdown.
The tie-in to existing plant systems will occur in Pkg 2.
This DCP does not impact any accident analysis previously evaluated at Hope Creek Station.
It involves the installation of seismic 1 supports, installation of tube tray and valves to seismic category 1 criteria, and the routing of non-Q tubing to Seismic category 1 criteria.
Therefore this DCP does not involve any Unreviewed Safety Questions.
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Procedure Description gf Safety Evaluation HC.RE-FR.ZZ-021(Q)
This procedure provides an outline of the Rev. 1 activities associated with fuel rod retrieval at Hope Creek.
This revision eliminates the use of the spent fuel pool racks as an intermediate storage location for the canister and changes the method of moving the canister from the auxiliary hoists on the refuel platform to the auxiliary hook on the reactor building polar crane.
As required in the UFSAR 9.1.4.6 this procedure revision requires a single failure proof handling system for the movement of the canister over the spent fuel pool.
Therefore, This procedure revision does not involve any Unreviewed Safety Questions.
HC.OP-GP.HB-0001(R) This new procedure provides a method for Rev. O release of Radioactive Liquids to the Hope Creek Discharge Canal from sources and pathways other than those normally used as described in FSAR Sect. 11.2.3 (Rad Releases).
The normal pathway as described in the FSAR is through the CST or to the Cooling Tower Blowdown line via existing flow paths.
The procedure will allow the use of pathways not previously described in the FSAR and therefore constitutes a change to the facility.
FSAR Sections 15.7 Rad Release Accident Analysis, 6.0 Engineered Safety Functions and 7.5 Info Systems and Instrumentation Important to Safety were reviewed and found that this procedure revision does not increase the probability or consequences of a previously analyzed accident.
Therefore, this procedure revision does not involve any r
Unreviewed Safety Questions.
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Other Descriotion ni thg Safety Evaluation Safety Eval FA This Safety Evaluation (SE) provides justification IAW requirements established in NRC IE Bulletin 80-10 Action step 3 for the continued operation of the normally non-i radioactive Auxiliary Boiler (FA) system.
On July 12, 1993, the "A" and "B" Auxiliary Boilers were found to be contaminated with total activities of approximately 3.8E-06 pCi/ml and 1.3E-06 pCi/ml respectively.
The constituent nuclides were low levels of Cr-51, Mn-54, Fe-59, and Zn-65.
IE 80-10 states that "If these nonradioactive systems are or become contaminated, further use of the system shall be restricted until the cause of the contamination is identified and corrected and the system has been decontaminated.
Decontamination should be performed as soon as possible.
However, if it is considered necessary to continue operation of the system as contaminate, an immediate safety evaluation of the operation of the system as a radioactive system must be performed IAW the requirements of 10CFR50.59".
This SE evaluates the continued use of the Auxiliary Boiler System by establishing criterion for allowable contamination levels for the Auxiliary Boiler.
Continued operation of the Auxiliary Boiler as a contaminated system at the current contamination levels (23.8E-06 pCi/ml) is judged to be prudent course of action for the following reasons:
- 1) Sampling and controls will be established for the effluent pathways to assure all releases are maintained ALARA as addressed in Appendix I to 10CFR50 with the corresponding environmental dose limits of 40CFR190.
- 2) Auxiliary Boiler blowdown has been and will be minimized.
- 3) Radioactively contaminated equipment, sludges and other materials will not be disposed of until all state environmental and NRC regulations have been satisfied.
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- 4) At the low levels of activity observed to date (3.8E-06 pCi/ml), verification that acceptable contamination levels are not exceeded are based on Chemistry sampling results.
Smears and air samples of the Auxiliary Boiler systems were obtained by Radiation Protection Department.
The
l Radiation Protection smears were analyzed with negative results.
Chemistry sampling analyses and Radiation Protection smears will continue to be used to evaluate contamination levels.
No physical changes to the Auxiliary Boiler i
system are being proposed.
There are no i
accidents for the Auxiliary Boiler evaluated in the UFSAR.
The Auxiliary Boiler remains identical as described in the UFSAR.
l Therefore, there is no increase in the consequences of a malfunction of equipment important to safety previously evaluated in 4
the SAR and does not involve any Unresolved Safety Questions, i
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