ML20078Q937

Monthly Operating Rept for Jan 1995 for Hope Creek Generating Station Unit 1

ML20078Q937
Person / Time
Site:	Hope Creek
Issue date:	01/31/1995
From:	Hovey R, Lyons D Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9502220137
Download: ML20078Q937 (10)

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'O PSEG ,

Public Service Electric and Gae Cornpany P.O. Box 236 Hancocks Bridge, New Jersey 08038 Hope Creek Generating Station February 15, 1995 i

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 January are being forwarded to you with the summary of changes, tests, and experiments that were implemented during January 1995 pursuant to the requirements of 10CFR50.59(b).

Si cerely yours, Wf R J. Hovey General Manager -

g Hope Creek Operations MT DR:WS:JC i

Attachments C Distribution l

t hbOb(b b T he F ' ~ ~ D o ^ n'a ft 9502220137 950131 PDR ADOCK 05000354 \ 2"3 *' '2 "

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1 NUMBER SECTION OF PAGES Average Daily Unit Power Level. . . . . . . . . . . 1 Operating Data Report . . . . . . . . . . . . . . . 3 Refueling Information . . . . . . . . . . . . . . . 1 Monthly operatir.g Summary . . . . . . . . . . . . . 1 Summary of Changes, Tests, and Experiments. . . . . 2 I

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OPERATING DATA REPORT i l

DOCKET NO. 50-354  !

UNIT Hope Creek  ;

DATE 02/10/95 l COMPLETED BY D. W. Lyons l

TELEPHONE (609) 339-3517 OPERATING STATUS f

1. Reporting Period January 1995 Gross Hours in Report Period 744

2. Currently Authorized Power Level (MWt) 3293 I Max. Depend. Capacity (MWe-Net) 1031 i Design Electrical Rating (MWe-Net) 1067 l

3. Power Level to which restricted (if any) (MWe-Net) None i

4. Reasons for restriction (if any) l This Yr To  !

Month Date Cumulative  !

5. No. of hours reactor was critical 744.0 744.0 60679.9  :

6. Reactor reserve shutdown hours 0.0 0.0 222 l l

7. Hours generator on line 744.0 744.0 59747.4 l

8. Unit reserve shutdown hours og n 0.0 0.0 l I

9. Gross thermal energy generated 2410230 2410230 190824575 }

(MWH) l

10. Gross electrical energy 811347 811347 63239013 l generated (MWH) {

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11. Net electrical energy generated 778188 778188 60433504 i (MWH) i

12. Reactor service factor 100.0 100.0 85.3 I

13. Reactor availability factor 100.0 100.0 85.3 1

14. Unit service factor 100.0 100.0 84.0 J

15. Unit availability factor 100.0 100.0 84.0

16. Unit capacity factor (using MDC) 103.5 101.5 82.4

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17. Unit capacity factor 98.0 98.0 79.6 (Using Design MWe)

18. Unit forced outage rate 0.0 0.0 4.7

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

OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-354 UNIT HoDe Creek DATE 02/10/95 COMPLETED BY D. W. Lyons TELEPHONE (609) 339-3517 MONTH January 1995 METHOD OF SHUTTING DOWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE NO. DATE S= SCHEDULED (HOURS) (1) POWER (2) ACTION / COMMENTS

1. 1/14 S 0 H 5 - POWER UNIT POWER WAS REDUCTION REDUCED TO PERFORM

>20%. - ROD SWAPS

- SCRAM TIMING

- HCU MAINTENANCE

- CIV TEST I

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AVERAGE DAILY UNIT POWER LEVEL  !

i DOCKET NO. 50-354  ;

UNIT HoDe Creek  !

DATE 02/10/95 i COMPLETED BY D. W. Lyons l TELEPHONE f609) 339-3517 MONTH January 1995 s l

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL l (MWe-Net) (MWe-Net)  !

1. 1052 17. 1060 I

3. 1060 18. 1058

3. 1056 19. 1061 1

4. 1071 20. 1058

5. 1064 21. 91 5,

6. 1060 22. J051

7. 1060 23. 1075 i I

8. 1052 24. 1061

9. 1063 25. 1054 1

1074

10. 1061 26.

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11. 1059 27. 1064 >

13. 1056 28. 1065

13. 1052 29. 1059

14. Spl 30. 1062 j 1

15. 1044 31. 1032

16. 1051 i

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REFUELING INFORMATION j l

l DOCKET NO. 50-354  ;

UNIT HoDe Creek 1 l DATE 02/10/95 COMPLETED BY R. Schmidt t TELEPHONE (609) 339-3740 j i

MONTH January 1995 l

1. Refueling information has changed from last month:

Yes No X  !

2. Scheduled date for next refueling: 10/14/95

3. Scheduled date for restart following refueling: 11/13/95 i

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

Yes No X l B. Has the Safety Evaluation covering the COLR been reviewed by the  !

Station Operating Review Committee?  ;

Yes No 5 X

If no, when is it scheduled? Auaust 28. 1995

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

Hgt recuired.  !

6. Important licensing considerations associated with refueling: t HlA

7. Number of Fuel Assemblies:

A. Incore 764 I B. In Spent Fuel Storage (prior to refueling) 1240 >

C. In Spent Fuel Storage (after refueling) 1472  ;

8. Present licensed spent fuel storage capacity: 4QQG 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: l (Does allow for full-core offload)

(Assumes 244 buncIIe reloads every 18 months until then)  ;

(Does Dgt allow for smaller reloads due to improved fuel)

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HOPE CREEK GENERATING STATION MONTHLY OPERATING

SUMMARY

January 1995 Hope Creek entered the month of January operating at 100% power.

From 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> on January 14 until 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> on January 15, 1995, power was reduced to perform control rod swaps, scram timing, hydraulic control unit maintenance and the weekly CIV Test. On January 21 from 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> until 2359 hours0.0273 days <br />0.655 hours <br />0.0039 weeks <br />8.975995e-4 months <br />, unit power was reduced to complete the control rod swaps and do the weekly CIV Test. On January 31, 1995, power was reduced to 95% because of a transient in the "C" Feedwater Heater Train. There were no other major power reductions or scrams. As of January 31 1995, the unit has been on line for 112 consecutive days.

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SUMMARY

OF CH7.NGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION January 1995 The following items have been evaluated to determine:

1. If the probability of occurrence or the consequences of an i accident or malfunction of equipment important to safety {

previously evaluated in the safety analysis report may be i 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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Temporary Modification Summary of Safety Evaluation T-Mod 95-02: This Temporary Modification lifts leads on the signal resistor unit (SRU) that feeds the cooling tower average basin water temperature loop signal summator. These instruments feed CRIDS point (A2769) which is used by the shift to determine when to enter and exit the cooling tower de-icing fill bypass ,

modes. This data summator averages four Temperature Elements '

(TE's) inside the cooling tower basin. One of the TE's has a failed temperature transmitter which is sending an erroneous (down scale) signal to the summator affecting the average temperature read at A2769. This T-Mod isolates the "B" input at the SRU and combines the output with that of the "A" to provide a 50% weighted average 4 input summer. A new transmitter has been ordered and is expected to be received by 3/31/95.

Failure of the entire Circulating Water System does not compromise any safety related system or component or prevent a safe shutdown of the plant. The only postulated accident evaluated in the FSAR involves the complete rupture of one of the expansions joints which is located in the turbine building. The SAR states that this does not adversely affect any safety related system. The complete failure of the basin temperature averaging circuit can ,

not have any affect on the expansion joint failure described in the FSAR Section 10.4.5.3.

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.

T-Mod 95-004: This Temporary Modification installs steam heat tracing on the Station Auxillary Boilers for the purposes of preventing instrument line freeze-up as climate conditions cause outside temperatures to fall below 32*f.

Failure of the installed tubing for the heat tracing presents no credible failure mode associated with this change. The auxiliary boiler steam system has no safety related function. The system is designed so that the failure of the system or a system component does not compromise any safety related systems or components or prevents a safe reactor shutdown.

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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6ther Summary 2f Safety Evaluation UFSAR Section 2.3 Meteoroloav Chance: This UFSAR change includes f the removal of detailed information provided in Table 2.3-29 and t text reference to this information contained on page 2.3-36 of the UFSAR. Specifically the change involves removal of meteorological instrumentation and strip chart manufacturer, model specifics from the UFSAR. This change will make this section more consistent with other sections of the HC UFSAR which do not include this level of detail. There is no analysis of accidents evaluated in the HC UFSAR which include the Artificial Island Monitoring Program or meteorological instrumentation in the evaluation.

Therefore, this UFSAR change does not increase the probability or consequences of an accident previously described in the SAR and g does not involve an Unreviewed Safety Question. '

Revision 21 Previous ReDort:

A previous report submitted to the Commission dated November 15, 1994 reported that a station operating procedure HC.OP-SO.AB-001(Q) would be revised and required a 50.59 safety evaluation. A 50.59 was prepared as a contingency action in the event the procedure revision was needed however, the procedure revision was not required or implemented.

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