ML20097F583

Monthly Operating Rept for May 1992 for Hope Creek Generating Station,Unit 1

ML20097F583
Person / Time
Site:	Hope Creek
Issue date:	05/31/1992
From:	Hagan J, Hollingsworth, Zabielski V Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9206150379
Download: ML20097F583 (11)

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. 'ubhe' Service ElecPic and Gas Company P O. Box 236 Hancocks Bridge New Jersey 08038 Hope Creek Generating Station June 10, 1992 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 May are being forwarded to you along with the summary of changes, tests, and experiments for May 1992 persuant to the requirements of 10CFR50.59(b).

Sincere y yours,

(

one al Ha aager -

Hope Cree Operations R:ld Attachments C Distribution I

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9206150379 920531 op i PDR ADOCK 05000354 ,

PDR R

The Energy People f'

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l INDEX NUMBER SECTIQR gy pAGES Average Daily Unit Power Level. .. . . . ., . . . 1 Operating Data Report . . . . . . . . .. . . .. . 2 Refueling Information . . . . . . . . . . . . . . . 1 Monthly Operating Summary . . . . . . . . . . . . . 1 Summary of Changes, Tests, and Experiments. . . . . 4 B

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, AVERAGE DAILY UNIT POWER LEVEL 4

DOCKET NO'. 50-354 UNIT lione Creek DATE _ji.fj0/92

COMPLETED BY y. Zabielski TELEPHONE (609) 339-3505 1 MONTH May 1992 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) i

1. 1054 17. 1050 2, 1042 18. 1049*

3, 1043 19, 1049*

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4. 1044 20. 1057.

5. 1912 21. LQig.

. 6. 1049 22. 121Q

7. 1053 23. 1036

8. 1072 24. 1036

9. 1052 25. 1056 l 10. 242 26. 111 11, 1055 27. 0

[ 12. 1054 28. 2

13. 1043 29. a 14, 1911 30. 2

15. 1053 31. 111 1

16. 1048 e Due to an error'in recording the meter readings, the exact average daily power levels for May 18 and:19 are unknown. The listed averages represent the average of the two-day total.

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. OPERATING DATA REPORT DOCKET-NO. 50-354

-UNIT ~ Hooe Crsek DATE 6/10/92 e r- '

COMPLETED BY V. Zakiglg& W .

TELEPHONE" -(6091 339-3506 OPERATING STATUS

1. Reporting Period May 1992 Gross Hours in Report Period 111

2. Currently Authorized Power Level (MWt) 1112 Max. Depeno.. Capacity (MWe-Net)- 1921 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 1[LCl 3329.5 40.490.8 6.- Reactor reserve shutdown hours ALE D2A- -2xQ-

7. Hours-generator on line 632si- 3286.3 22 860.9-

8. Unit reserve shutdown hours - ad 2x2- 192Q .

9. Gross; thermal energy generated 2.043.069 -10.536.961 -126.534.104 (MWH)

10. Cross electrical energy.

generated-(MWH)

678.860 2 521.'8%Q 41.874.374 h

11. Net electrical energy generated 647.860 -3.365.9511 '40.017.60Q

12. Reactor service factor 87,1 91,3 84,8

13. Reactor availability factor - Ahl ' :21x1- fL4J 14.' Unit service factor 15x2 90.1 83.5

.15. Unit availability factor 85.0 90.1 122L'

16. Unit-capacity factor (using MDC) . gM ' 31.5, 81.3

17. Unit capacity factor- ELui ~ 80.5 78.5

.(Using Design MWe)'

18.-. Unit forced outage rate 15.0 .221 5.xa- 4

-19. Shutdowns. scheduled over-next 6 months-(type,Ldate, &Dduration):'

Refueling = outage, 9/12/92,f60 days-

20. If shutdown at'end of report period, estimated date of start-up:L N/A

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a OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER-REOUCTIONS DOCKET NO,12-354 UNIT Hone CIpok DATE 6/10/92 COMPLETED BY V. Zabiqlski TELFPHONE f609) 339-3191 MONTH May 1992 1[ETHOD OF liHUTTING

)OWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE NO. DATE S= SCHEDULED (HOURS) (1) POWER (2) ACTION / COMMENTS 4 S/26 F 111.9 A 1&2 Failed Drywell to Suppression Chamber Decay test: power was reduced to 21%

and the Reactor was manually scrammed LER 354/92-006 s

P Summary

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REFUELING INFORMATION DOCKET NO. 50-354 _

UNIT 11o pe Cr e e k . .._.____

DATE 6/10/92 COMPLETED BY L__llgiljngsworth TELEPHONE .1109) 339-1051 MO) Yu May 1S32

1. Refueling information has changed from last month:

Yes No X

2. Scheduled date for next refueling: 9/12/92

3. Scheduled date for restart following refueling: 13/11/92

4. A. Will Technical Specification changes or other license amendments be required?

Yes No X B. Has the reload fuel design been reviewed by the Station Operating Review Committee?

Yes No X If no, when is it scheduled? Dpt scheduled (on or prior to 7/24/92)

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

6. Important licensing considerations associated with refueling:

- Same fresh fuel as current cycle: no new considerations

7. Number of Fuel Assemblies:

A. Incore 764 B. In Spent Fuel Storage (prior to refueling) 760 C. In Spent Fuel Storage-(after refueling) lang

8. Prcsent licensed spent fuel storage capacity: 4006 Future spent fuel storage capacity: 4006

9. Date of last refueling that can be discharged 11/4, 2010 to spent fuel pool assuming the present (EOC16) licensed capacity:

(does not allow for full-core offload)

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. HOPE CREEK GENERATING STATION MONTHLY: OPERATING'

SUMMARY

- May 1992 Hope Creek entered the nonth of May:at approximately 100%~ power.  ;

A reactor shutdown was commenced at 1503-on May 26-per the requirements of. Tech Spec 3.6.1.1 because the Drywell to Suppression Chamber Decay Test-failed to meet its-acceptance!-

criteria. To comply with the Action Statement, a manual scram was initiated at 2213 with reactor power.at 21%..- The unit:was brought -

back on line on May 31.

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SUMMARY

OF CliANGES, TESTS, AND EXPERIMENTS FOR Tile llOPE CREEK GENERATING STATION MAY 1992 J

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9 The following items have been evaluated to determine:

1. If the probability of occurrence or the consequences of an acc.ident or mal.' unction of equipment important to safety previously evaluated in the safety ane. lysis 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 specitication 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 chance 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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l IMd Dss.criution of Safety Ev.gluation 92-012 This TMR installed electricel jumpers across the iiigh Bearing 011 Temperature Trip Switch. This jumper permits the 'A' Control Room Chiller to run with a defective module-and/or thermistor until a replacement part can be installed.

The Control Acca Chilled Water System is comprised of thetwo lossluo% capacity of any single redundant componentloops; cannottherefore[n result a loss of cooling. A3so, jumpering the trip circuit and providing for increased operator attention to oil temperature does not place the equipnent in any additional jeopardy. Therefore, this TMR does not involve any Unroviewed safety Questions. )92-013 This TMR installed electrical jumpers across the Feedwater Heater's High High Level Trip Switches.

These switches cause spurious high level trip signals during Icw power levels due to inloakage in the reference leg. The jumpers are only required until. the level signals stabilize.

The Feedwater system is not safety 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 Unroviewed Safety Questions.92-014 This TMn removed the overload heaters fiam the breakers for the Reactor Water Cleanup Discharge to Condenser Valve Tnd 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 valvee, along with the. overhead annunciator, does not prevent'their associated systems from performing their designed functions.

Also[rement requ that-the valves-be disabled.the UFSAR discusses the Appendix R Therefore, this TMR does not involve any Unreviewed Gafety Questions.

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j Procedure  !

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Rivision Descrinticq of SAlaty Evaluation l

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l HC.OP-GP.ZZ-0001(Q) This new procedure eliminates the

Rev 0 possibility of a Residual Heat Removal ,

i Shutdown Cooling Isolation due to a loss of Reactor Protection System-power by

defeating the automatic isolation signals-to the Shutdown-Cooling Suction Isolation. ,

l Valves and the Shutdown-Cooling Return to

' Reactor Pressure Vessel Valves. This procedure will be-used only dur.ing refueling with the Reactor cavity-flooded, the Fuel Pool to Reactor cavity _gaten 3

-removed, a.id- with management approval.

l' The Shutdown Cocling mode:of the Residual Heat Removal System is. designed to-be controlled by the operator from the Control '

, Room. :The design basis -for- the most -

! limiting single' failure is that Shutdown Cooling can.be establishedtby manual action. This procedure retains the ability to manually isolate the Shutdown Cooling Suction Isolation-Valves. Elimination of the' automatic isolation capabilit jeopardize the: functional-design y=does not basis of-the Shutdown' Cooling mode of the Residual Heat Removal-System; therefore,-there are no-Unreviewed Safety. Questions associated.

with this new procedure, e

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