ML20126A496

Monthly Operating Rept for Nov 1992 for Hope Creek Generation Station Unit 1

ML20126A496
Person / Time
Site:	Hope Creek
Issue date:	11/30/1992
From:	Hoolingsworth, Zabielski V Public Service Enterprise Group
To:
Shared Package
ML20126A494	List: ML20126A492 ML20126A496
References
NUDOCS 9212210088
Download: ML20126A496 (12)

Text

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INDEX NUMBER SECTION OF PAGES Average Daily Unit Power Level. . . . . . . . . . 1 Operating Data Report . . . . . . . . . . . . . . . 2 i

Refueling Information . . . . . . . . . . . . . . . 1 Monthly Operating Summary . . . . . . . . . . . . . 1 Summary of changes, Tests, and Experiments. . . . . 6 g.

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AVEPAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-354 UNIT Hone creek __

DATE- 12/15/92 -

COMPLETED BY V. Zabielski ' l/C s ..

-TELEPHONE (609) 339-3506 MONTH November 1992 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) _(MMe-Net)

1. A 17. 1069

2. A 18. 1Q12

3. A 19. 1068 t

4. A 20. 1066

5. A 21. 1063 '

6. A 22, 1911

'7. A 22 . 1043

8. A 24, 1051

9. A 2 5. . 1063

10. 122' 26. 1059.

11. 212 27. 1251

12. ARji 28. 1068 l

13, 221 29. 1059

14. 1089 30. 1073 15 . - 1062 31. HZA ,

16. 1072 i

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l OPERATING DATA REPORT DOCKET NO. 50-354 UNIT Hone Creek DATE 12/15/92 COMPLETED BY V. Zabielski -

TELEPHONE (609) 339-3506 OPERATING STATUS

1. Reporting Period November 1991 Gross Hours in Report Period 720

2. Currently Aut'horized 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)

Th '_6 Yr To

'.onth Date Cumulative

5. No. of hours reactor was critical 584.0 6388.5 43.549.8

6. Reactor reserve shutdown hours 22R 0.0 0.0 4

7. Hours generator on line 497;l 6239.8 42.814.3

8. Unit reserve shutdown hours 0R 1 0.0 0.0

9. Gross thermal energy generated 1.485.775 19.994.138 135.991.281 (MWH)

10. Gross electrical energy 498.850 6.644.440 44,996.934 generated (MWH)

11. Net electrical energy generated 472.323 6.332.581 42.984,130

12. Reactor service factor 83.1 79.5 83.5

13. Reactor availability factor 6 ~, .1 79.5 83.5

14. Unit service factor 69.1 77.6 82.1

15. Unit availability factor 69.1 77.6 82.1

16. Unit capacity factor (using MDC) 63.6 76.4 79.9

17. Unit capacity factor 61.5 73.8 77.i (Using Design MWe)

18. Unit forced outage rate 0.0 2.1 4.7

19. Shutdowns scheduled o"er 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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4 OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-354 UNIT Hope Creek DATE .12/15/92 COMPLETED BY TELEPHONE V.

Zabielski'U h (609) 339-3506 MONTH November 1992 METHOD OF SHUTTING DOWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE NC. DATE S= SCHEDULED (HOURS) (1) POWER (2) ACTION / COMMENTS 9 11/1 S 222.2 C 4 Continuation of 4th. Refueling Outage l

Summary l'

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REFUELING INFORMATION DOCKET NO. 50-354 UNIT HoDe Creek DATE 12/15/92 COMPLETED BY S. Hollinosworth TELEPHONE (609) 339-1051

-HONTH Noverker 1992

1. Refueling information has changed from last month:

Yes X No

2. Scheduled 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?

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? 2/18/94

5. Scheduled date(s) for: submitting proposed licensing action: HlA

6. Important licensing considerations associated with refueling:

- Highly likely that will use same or similar fresh fuel as current cycle: no new considerations.

7. Number of Fuel Assemblies:

A. Incore 764 B. In Spent Fuel Storage (prior to refueling)- 1008 C. In Spent Fuel Storage-(after-refueling) 1232 to 1264

8. Present licensed spent fuci storage capacity: 4006 Future spcnt fuel storage capacity: 4006

9. Date of last refueling that can be discharged 11/4/ 201Q to spent fuel pool as3uming the present (EOC16)

-licensed capacity:

(does not allow for full-core offload)

(this item not expected to be updated before January 1993)

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HOPE CREEK GENERATING STATION MONTHLY OPERATING

SUMMARY

NOVEMBER 1992 The 4th Refueling Outage began on September 12 and continued throughout the month of October. On November 6 the reactor was taken critical. On November 10 at 1334, the un[t was put on line, ending the refueling outage. The unit operated for the remainder of the month without experiencing any shutdowns or reportable power reductions. As of November 30, the plant had been on line for 20 consecutive days.

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SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION NOVEMBER 1992

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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IME _Qggcription of Safety Evaluation 92-027 This TMR. installed electrical jumpers across the #2-Feedwater Heater High High Level Trip Switches.

These switches cause spurious high level trip

-signals during low power levels. The jumpers were removed after the level signals stabflized.

The Feedwater system is not cafety related and is not required to be operable following a LOCA, other than for containment isolation. Failure of the Feedwater system does not compromise any safety related system or components. This TMR has no impact on the containment isolation function of the Feedwater system. Therefore, this TMR does not involve any Unreviewed Safety Questions.92-029 This TMR installed instrumentation on four Reactor Vessel Level Reference Leg Condensate Chambers to provide temperature-monitoring. This TMR was initiated in response to NRC Generic Letter 92-04, concerning-possible non-condensable gas accumulation with the transmitter reference legs.

The objective of the TMR is to measure the external surface temperatures of the top and bottom walls of the condensate chambers.

An insulation pad was used to prevent heat transfer between the two thermocouples. It was also determined that this TMR has no impact on loading.

The addition of external thermocouples does not have any affect on the internal operation of the condensate chambers. Therefore, this TMR does not involve any Unreviewed Safety Questions.92-030 This TMR removed the. overload heaters from the breakers for the Reactor Water Cleanup Discharge to Condenser Valve and the Reactor Water Cleanup Discharge to Equipment Drain Valve.- Removing,the-overload heaters from the breakers will prevent the-valves from inadvertently opening during an Appendix R fire.

Disabling these valves, along with the overhead annunciator, does not prevent their associated systems from performing their designed functions.

Also, the UFSAR discusses the Appendix R requirement that the valves be disabled.

Therefore, this TMR does not involve any Unreviewed Safety Questions.

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DE DescriDtion of Deficiency Reoort HMT 92-216 This DR addresses the~"use-as-is" disposition of an-elbow on a drain line back to the 'C' Condenser.-

Portions of the elbow have a~ wall thickness of less than the-allowable thickness:plus a conservative safety _ factor.

, There-are no equipment or components important to safety located in the area that could be affected-by a failure of the line.- The High Pressure Coolant Injection and Reactor Core Isolation-Cooling Drain Pots-that this line-serves isolate on an initiation signal. ~The failure-of this-elbow will not-prevent the High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems from operating. Therefore, the "use-as-is"=

disposition of this DR does not involve any Unreviewed Safety Questions.

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Procedure Revision Description of Safety Evaluation HC.OP-IS.KL-0101(Q) These procedure revisions change the Rev. 7 position of the isolation valves for the Suppression Chamber / Vacuum Breaker Gas HC.OP-IS.KL-0102(Q) Supply Line from open to closed when the Rev. 7 gas line is not being used to test the vacuum breakers. This change will help to eliminate the possibility of water intrusion into Primary Containment Instrument Gas affecting the vacuum breakers.

The realignment of these valves does not in any way affect the operation of the system or its response to accident conditions.

Control of these valves is still available from the Control Room. The closed position is conservative because that is the accident position for these valves.

Therefore, these procedure revisions did not involve Unreviewed Safety Questions.

NC.NA-AP.ZZ-0048(Q) This procedure revision affects the Rev. 1 administrative controls for the portion of the performance monitoring program that relates to component monitoring. This revision was initiated in responsa to the 1991 INPO assessment of Salem Nuclear Generating Station.

This procedure revision enhances the process and does not change the intent of the administrative controls for performance monitoring as described in the UFSAR.

Therefore, this procedure revision does not involve an Unreviewed Safety Question.

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Q UFSAR Section Description of Eafety Evaluation 12.3 The symbol used to designate resin fill cover plates on a Shielding and Radiation Zoning drawing is similar to the symbol used to designate Area Radiation Monitor locations on Shielding and Radiation Zoning drawings. This UFSAR Change Notice deletes the Area Radiation Monitor location symbols from figures and revises to identify reference instrument location drawings.

No changes were made to Radiation Monitoring System equipment. The change does not affect the operation of the Radiation Monitoring System. Therefore, this change does not involve any Unreviewed Safety Questions.

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