ML20086A312

Monthly Operating Rept for Hope Creek Generating Station, Unit 1,for Oct 1991

ML20086A312
Person / Time
Site:	Hope Creek
Issue date:	10/31/1991
From:	Hagan J, Zabielski V Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9111190039
Download: ML20086A312 (12)

Text

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Q.PSEG Pubhc Service Electric and Gas Company P.0, Box 23G Hancocks Bridge, New Jersey 080'28 i Hope Creek Generating Station -

November 14, 1991 l

-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 October are being forwarded to you with the Lsummary of changes, tests, and experiments for October 1991 pursuant to the requirements of 10CFR50.59(b).

Sincerely yours, on QL J. a an Generd1 nager -

Hope (r,gdk Operations RAR:ld

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9111190039 911031- mumny, u e3 PDR ADOCK 00000354

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INDEX l 1

NUMBER SECTION QF PAGES .;

l Average Daily Unit Power Level. . . . . . . . . . . 1  !

Operating Data Report . . . . . . . . . . . . . . . 2 Refueling Information . . . . . . . . . . . . . , , 1 Monthly Operating Summary . . . . . . . . . . . . . 1 Summary of Changes, Tests, and Experiments. . . . . 5 i

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-354 UNIT Hope Creek DATE 11/14/91 COMPLETED BY V. Zabielski TELEPHONE (609) 339-3506 MONTH October 1991 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)

1. 1044 17. 1056

2. 1044* 18. 1052

3. 1044* 19. 1051

4. 1052 20. 1036

5. 111 '21. 1061

6. 1952 22, 1052

7. 1054 23. 1049

8. 1054 24.- 191;}.

9. (Qld 25. 1041

-10. 3048 26. lail

11. 1044 27. 1070

12. 1060 28, 1935 1041 1951

-13. 29.

-14. 1062 30, 1062 15, 1046 31. 1045

.16 . 1050 CDue to an error in recording the meter readings, the exact average daily.

power level for October 2 and 3 is unknown. The listed averages for those two days reprasent the average of the two-day total.

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OPERATING DATA REPORT l

DOCKET NO. 50-354 UNIT Hooe Creek j DATE 11/14/91 i COMPLETED BY V. Zabielski TELEPHONE (fi2 .539-3001 OPERATING STATUS

1. Reporting Period October 1991 Gross Hours in Report Period 715 1

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 745.0 5915.8 35.697.3

6. Reactor reserve shutdown hours 222 0.0 Daa

7. Hours generator on line 2452R 5817.5 35,110.6

8. Unit reserve shutdown hours 922 229 0.0

9. Gross thermal energy generated 2,459.202 18.846.500 111,388.907 (MWH)

10. Gross electrical energy 841.760 6.200,361 36,822.034 generated'(MWH) 11.. Net electrical energy generated 780,535 5.926,488 35.183.172 (MWH)

13. Reactor service factor 100.0 81.1 83.7

13. Reactor availability factor 100.0 81x1 83.7-

14. Unit service factor 100.0 79.7 32.3

15. Unit availability' factor 100.0 79.7 g223,

16. Unit capacity factor (using MDC) 101.6 78.8 80.0

17. Unit capacity factor 98.2 76.1 77.3 (Using Design MWe) r- 18. Unit forced outage rate Azq 522 114

19. Shutdowns scheduled over next 6 months (type, date, & duration):

j- None l 20. If shutdown at end of report period, estimated date of start-up:

N/A l

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OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-354 UNIT Eppe Creek DATE 11/14/91 COMPLETED BY V. Zabielski i TELEPHONE (609) 339-3506 MONTH October 1991 i METHOD OF SHUTTING DOWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE ,

NO. DATE S= SCHEDULED (HOURS) (1) POWER (2) ACTION / COMMENTS None. j Summary-

4 REFUELING INFORMATION DOCKET NO. 50-354 UNIT lip +_ Creek DATE 11.14/9:

COMPLETED BY S. HgliJ1gnyprth TELEPHONE .f609) 339-1051 MONT' Qctober 1991

1. nefueling information has changed from last month:

Yes No X

3. Scheduled date for next refueling: 9/5/92

3. Scheduled date for restart following refueling: 11/4/92

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

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

Yes No X If no, when is it scheduled? not scheduled (on or prior to 7/24/92)

5. Scheduled date(s) for submitting proposed licensing action: 11/A

6. Important licensing considerations associated with refueling:

- Sano fresh fuel as current cycle no new considerations

7. Number o2 i'uel Assemblies:

-2 A. Incore 211 B. In Spent Fuel Storage (prior to refueling) 760 C. In Spent Fuel Stcrage (after refueling) 1R25

8. Present lic.ensed spent fuel storage capacity: 4006 Fature spent fuel storage capacity: ADR1

9. Date of last refueling that can be discharged 11/ 4 .__2fdGL to spent fuel pool assuming the present (EOC16) licensed capacity:

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IlOPE CREEK GENERATING STATION MONTHLY OPERATING

SUMMARY

October 1991 ilope Creek ontered tho month of October at approximately 100%

power and operated for the entire month without experiencing any shutdowns or reportable power reductions. On October 31 th plant completed its 172nd day of continuous power operation.e

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44 4

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FOR TiiE Il0PE CREEK GEliERATIliG STATIOli OCTOBER 1991 I

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The following items have been eveluated 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 Descriotion of Safety EvaluatiQD 91-045 This TMR replaced fiberglass insulation with ,

cellular glass insulation and pittwrap on the High '

Pressure Coolant Injection, Reactor Core Isolation Cooling, and Core Spray Suction Lines from the Condensate Storage Tank, and on the instrument line ,

used to automatically transfer High Pressure Coolant Injection Suction from the condensate Storage Tank to the Torus.

No Unroviewed Safety Questions were involved because there was no significant increase in piping stresses or su of insulation.pport The loads as aglass cellular resultinsulation of the change has also been evaluated for material compatibility with stainless steel pipe and is in the acceptable range for leachable chlorides and fluorides.91-052 This TMR lifted the relay contact output wires to eliminate a falso alarm input. A failed lamp in the Electro-Hydraulic Control System was causing an overhead annunciator to alarm an EHC Cabinet Problem. The relay contact that was jumpered out provides an alarm only. The contact provides no turbine trip function. Therefore, the turbine overspeed trip function is not affected by this TMR and there is no reduction in tha prevention of missiles resulting from a turbins overspeed.

The Electro-Hydraulic Control Cabinet has no safety-related function, is non-seismic, non-1E, and not Environmentally Qualified. The failure of this alarm is bounded by Pressure Regulator Open/ Closed Failures and Turbine Trip, both discussed in Chapter 15 of the UFSAR.

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Procedure -

Revision Description of safety Evaluatisn HC.MD-GP.ZZ-0100(Q) This procedure controls the use of the Rev. O HOVATS Checkmate II Test System. It is a non-intrusive valve analysis system that utilizes ultrasonics and acoustics to detect valve degradation. The system is used to acquire data for stable flow tests, stroke tests, and leak tests. The resultant data, acquired for stroxo time, will be used to satisfy Technical Specification requiremente.

No Unreviewed Safoty Questions were inrolved because the test method is non-intrusive and there is no affect on the artiity of the valve to perform as desi'ned. Technical Specification 4.0.5 estal dab .4 surveillance requirements for inserv ice .esting and inspection of ASME osta Cla ? 4 1, 2, and 3 components. Section 4.i Eg requires testing to be performed on those components. This p' .,cedure catablishes a method to comply with the Technical Specification requirements and meets those requirements.

NC.NA-AP.ZZ-0005(Q) The majority of the changes that were made Rev. 1 as a result of the revision to this administrative procedure are either editorial or have been added to incorporate recommendations from INPO SOER 91-01, Conduct of Infrequently Performed Tests or Evolutions.

No Unreviewed Safety Questions were involved because this procedure is in compliance with the SAR, and as stated above, the majority of the changes are either editorial or have been added to incorporate recommendations from INPO SOER 91-01.

MESAR SeckiDD Descriotion of Safety Evaluation 1.10.2 This UFSAR Change Notice deletes references to 9.5.1.5.1 specific training procedures by procedure 13.2.1 number. It also removes the responsibility App. 13A - 13K for conducting General Employee Training from the responsibilities for the Manager - Nuclear Training. Additionally, it rewrites the section on the Plant Personnel Training Program to state that the training programs are accredited and the details documented in the appropriate Training Department procedures.

No Unreviewed Safety Questions were involved because the training commitments based on 10CFR50, 10CFR55, Regulatory Guide 1.8, and ANSI /ANS 3.1 are not being changed. They are being relocated to the appropriate Training Department procedures.

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