ML20058M594

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Monthly Operating Rept for Nov 1993 for Hope Creek Generating Station Unit 1
ML20058M594
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 11/30/1993
From: Hovey R, Zabielski V
Public Service Enterprise Group
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9312210038
Download: ML20058M594 (11)
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Text

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     'O PSIRG Put;hc Service Electr'c anc Ga:, Company P O. Box 236 Han rocks Bridge, New Jersey 08038 Hope Creek Generating Station December 15, 1993 U. S. Nuc2 ear Regulatory Commission Document Control Desk Washington, DC i*O555

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 November are being forwarded to you with the summary of changes, tests, and experiments that were implemented during October and November 1993 pursuant to the requirements of 10CFR50.59(b). Sincerely yours, R. J. HoveyI General Manager - Hope Creek Operations 1.2 J; R:WS:JC Attachments C Distribution l i 9312210038 931130 ' 

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PDR ADOCK 05000354 ~ 

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

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4 t ' l OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-354 UNIT. Hope Creek DATE 12/08/93 COMPLETED BY V. Zabielski,V(,, J F TELEPHONE (609) 339-3506-MONTH November 1993 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 t t 

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_ ~_ _ . . . . _ _ . _ _ i AVERAGE DAILY UNIT POWER LEVEL l DOCKET NO. 1.0-354 UNIT . Hone Croek i DATE 12/08/9'd i COMPLETED BY V. Za biel sk i 'III- - ] TELEPHONE (609) 339-3506 i 

                                                                                                                  'l MONTH      Hovember               1993 l

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY P"WER LEVEL l (MWe-Net) (MWe-Net) 1 ~

1. 1064 17. 1053
2. 106G 18. 1061
3. 1065 19. 1056 l
4. 1062 20. 1060 1
5. 1053 21. 1058 >
6. 1053 22. 1061
7. 1048 23. 1060
8. 1064 24. .1059 i
9. 1063 25. 1066 )
10. 1_012. 26. 1068 ,

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11. 1060 27. 1047 .

12, 1056 2 

                                                             .8.                 1049-                               i
13. 1057 29. 1065 )
14. 1038 30, 1067 'I
15. 1036 L31. HIA -:

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16. 1057- l
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t J l OPERATING DATA REPORT DOCKET NO. 50-354: UNIT Hope Creek DATE 12/08/93 g - COMPLETED BY V. Zabielskiv TELEPHONE (609) 339-3506 i OPERATING STATUS

1. Reporting Period November 1993 Gross Hours in Report Period.720
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)

I This Yr.To Month Date Cumulative

5. No. of hours reactor was critical 720.0 7935.0 52190.6
6. Reactor reserve shutdown hours 0.0 0.0 0.0
  .7.      Hours generator on line                 720.0              7916.9         51470.8
8. Unit reserve shutdown hours 0.0 0.0 0.0-
9. Gross thermal energy generated 2368952 25774329 163987547 (MWH) .i
10. Gross electrical energy 788530- 8550800 -54298854 generated (MWH)
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11. Net electrical energy generated. 756988 8189325 51891709 i (MWH) l l
12. Reactor service factor 100.0' 99.0 85.7- )
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13. Reactor availability factor 100.0 99'.0 85.7.
14. Unit service factor 100.0 98.8 84.4.

15.-Unit availability. factor 100.0 98.8 84.'4 

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16. Unit capacity factor (using MDC) 102.0 99.1 87 6-
  -il7.. Unit,capacityifactor                       98.5-               95.7-            79.8 (Using Design MWe)                                                                                    ,
18. . Unit forced' outage rate 0.0- 112. 4 .' 3 '
  -19. Shutdowns'scheduledfover-'next 6 months-(type, date, & duration):

Refueling ~ Outage 6, March 5,:1993, 49-days.

20. If shutdown at end of report period, estimated date of start-up:

N/A j i 1 

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o REFUELING INFORMATION DOCKET NO. 50-354 UNIT Hope Creek 1 DATE December 13, 1993 COMPLETED BY S. Hollinasworth TELEPHONE (609) 339-1051 MONTH December 1993

1. Refueling information has changed from last month:

Yes No X

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 3 amendmants be required? i 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:  ;
               'Not scheduled yet.
6. Important licensing considerations associated with refueling:

ULh i P 7.- Number of Fuel Assemblies: A. Incore. . 764 B. In Spent Fuel Storage (prior to refueling) 1008 C. In Spent Fuel Storage (after refueling) 1240 l

8. Present licensed spent fuel storage capacity:. 4006 Future' spent-fuel storage. capacity: 4000  ;
     - 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) (Doesinot allow for-_ smaller reloads due.to-improved fuel)' i 

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   ~e HOPE CREEK GENERATING STATION MONTHLY OPERATING

SUMMARY

1 November 1993 i Hope Creek entered the month of November at approximately 100%_ 1 power. The unit operated at full power through the end of the ' month without any major power reductions or plant trips. As of November 30, the plant has been on line for 195 consecutive days. I i M . n n . - - y e w.+, -uss & 

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SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION October and November 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 may be increased; or
2. II a prasibility 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 i 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. , i 5 b

Desian Chances Eummary of Safety Evaluation 4EC-03254 This DCP installs two small ventilation fans to each FRVS Ventilation System unit heater control panels. This panel is not described in the UFSAR however it does modify the procedure (OP-ST.GU-001(Q) which will verify operability of the fans during the monthly surveillance. Therefore, this DCP does not increase the probability or consequences of an accident previously described in the SAR and does not involve any Unreviewed Safety Question. Procedure Summary of Safety Evaluation NC.NA-AP.ZZ-007(Q) The procedure revision describes administrative changes to the AAARA program and as'such does not affect the operabililty or reliability of any safety system. Changes included were workorder processing and pre-planning, pre-job briefings,-ALARA checklists, rehearsals, and mock-up training. Therefore, this procedure revision does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question. Temporary Summary of Safety Evaluation Modifications 93-024 This Temporary Modification installs a temporary hose between Turbine Building 77 '- elevation to a vendor truck on turbine building 102' elevation. The purpose of-this is to allow - transfer of spent' condensate resins from the Condensate Demineralizer system.to the vendor truck for disposal. This T-Mod affects drawing M-16-1 included in UFSAR as 10.4-4 due to the removal of a 5 foot section of pipe to hook up1 the temporary hose. LThe Condensate Demineralizer system does not impact any safety related systems. Any spill would have been directed to the Rad Waste systems via the Turbine Building floor drains.  ! 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, e 

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4 Deficiency Summary of Safat.y Evaluation Reports HTE 93-044 This Deficiency Report describes the failure of Hiller actuators on the SACS cooling lines valves to the Diesel Generator Room Coolers. The valves were failed open to allow cooling to - the Diesel rooms. The normal line up position , is closed. The valves are failed open ' a support both the normal and abnormal operation of the SACS Diesel Generator Room coolers. Therefore, this Deficiency Report does not increase the probability or consequences of an accident previously described in the SAR and ., does not involve an Unreviewed Safety Question. > HMD-93-045 This Deficiency Report describes the Use As Is disposition of a leaking packing gland on valve , 1ABHV-F033 (Inboard Drain Valve) until such time the leak can be repaired by a temporary leak repair or the next available outage. By closing the Primary Containment Isolation Valves 1ABHV-F016 and F019 the leakage was minimized at the F033 valve. This placed the valves in the positions required for a Group 1 NSSSS PCI signal. The F016 and F019 valves remained operable. Therefore, this Deficiency. Report does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question. Other Summary of Safety Evaluation H03.5-11 This Safety Evaluation states that-during Cycle-4, segments in nine control rods will ' exceed 34% boron-10 depletion, which is their normal design life and: describes the basis that-General Electric used for cycle-4 licensing analysis. The ef fects of: this additional depletion requires deeper insertion of the-blades to achieve the desired effect. The rods will be able to fulfill the reactivity control function for shutdown margin. Therefore, this Safety Evaluation 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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DEH-93-00158 This Safety Evaluation demonstrates that a 4 minute time delay of an automatic Standby Liquid Control System (SLCS) from the beginning of an Anticipated Transient Without SCRAM (ATWS) event will satisfy the ATWS criteria even if the event commences.from the Extended Load Line Limit Analysis Region. Therefore, this Safety Evaluation does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question. l 

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