ML20149G269

Monthly Operating Rept for Sept 1994 for Hope Creek Generating Station Unit 1

ML20149G269
Person / Time
Site:	Hope Creek
Issue date:	09/30/1994
From:	Hovey R, Lyons D Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9410200112
Download: ML20149G269 (12)

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. Public Service Electric and Gas Company P,0. B6x 236 Hancocks Bridge. New Jersey 08038 1

Hope Cree'k Generating Station

.l October 14, 1994 l U. S. Nuclear Regulatory Commission

' Document Control Desk Washington, DC -20555

Dear' Sir:

MONTHLY OPERATING REPORT HOPE CREEK GENERATION STATION UNIT 1 i DOCKET NO. 50-354 In compliance with Section 6.9, Reporting Requirements for the Hope Creek Technical Specifications, the operating statistics for September are being forwarded to you with the summary of changes, -

tests, and experiments that were implemented during September-1994 pursuant to the requirements of 10CFR50.59(b).

Sincerely yours, R. J..Hoveh General Manager -

N Hope Creek Operations 9tX 4tf f . ,

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INDEX NUMBER SECTION OF PAGES ,

Average Daily Unit Power Level. . . . . . . . . . . 1 Operating Data Report . . . . . . . . . . . . . . . 3 i Refueling Information . . . . . . . . . . . . . . . 1 .

Monthly Operating Summary . . . . . . . . . . . . . 1 Summary of Changes, Tests, and Experiments. . . . . 4 {

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OPERATING DATA REPORT DOCKET NO. 50-354 UNIT Hope Creek DATE 10/11/94 COMPLETED BY D. W. Lyons TELEPHONE (609) 339-3517 OPERATING STATUS i

1. Reporting Period September 1994 Gross Hours in Report Period 712

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

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 719.6 5109.9 57932.9

6. Reactor reserve shutdown hours 22Q 0.0 222

7. Hours generator on line 713.0 4999.9 57032.4

8. Unit reserve shutdown hours pig 0.0 0.0 j

9. Gross thermal energy generated 2312673 16015088 181978458  !

(MWH)

10. Gross electrical energy 766071 5298761 60262715 <

generated (MWH) l

11. Net electrical energy generated 733441 5055097 57582781 j e

(MWH)

12. Reactor service factor 99.9 78.0 84,9 '

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13. Reactor availability factor 99.9 78oO 84 9 l I

14. Unit service factor 99.0 76.3 83.6

15. Unit availability factor 99.0 76.3 83.6

16. Unit capacity factor (using MDC) 98.8 74.8 81.9  !

17. Unit capacity factor 95.5 72.3 79.1 [

(Using Design MWe) l

18. Unit forced outage rate 112 522 4.5

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

None

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

N/A )

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l OPERATING DATA REPORT UNIT SHUTDOWNS-AND POWER REDUCTIONS -i l

DOCKET NO. 50-354 '

l UNIT Hooe Creek DATE 10/11/94 COMPLETED BY D.W. Lyons TELEPHONE (609) 339-3517 MONTH Seotember 1994 METHOD OF i' SHUTTING DOWN THE i TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE i' NO. DATE S= SCHEDULED (HOURS) (1) POWER (2) ACTION / COMMENTS 1 9/1 F 7 A 4 CONTIN- UNIT SYNCHRONIZED ON (TOTAL UATION 09/01/94 AFTER OF 45) COMPLETION OF  ;

REPAIRS. l l

3 9/3 S 0 B 5 POWER POWER REDUCED MORE i REDUCTION THAN 20% FOR ROD  ;

ONLY ADJUSTMENT, & CIV SURVEILLANCE TESTS. i l

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AVERAGE DAILY UNIT POWER LEVEL ,

DOCKET NO. 50-354 UNIT Hope Creek DATE 10/11/94  ;

COMPLETED BY D.W. Lyons TELEPHONE (609) 339-3517 MONTH September 1994 '

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

1. 112 17. 1035

2. 1012 18. 1037

3. Ela 19. 1057

4. 1060 20. 1051

5. 1056 21. 1049

6. 1067 22. 1039 i

7. 1030 23. 1034 1050 J044

8. 24.

9. 1039 25. 1032

10. 1047 26. 1041

11. 1.0J1 27. 1045

12. 1041 28. 1032 1

13. 1057 29. 1064 I

14. 1211 30. 1067 l l

15. 1042 31. nLa '

16. 1040 l

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REFUELING INFORMATION DOCKET NO. 50-354 UNIT HoDe Creek 1 DATE Oct. 11. 1994 COMPLETED BY R. Schmidt TELEPHONE (609) 339-3740 MONTH Sect 1994

1. Refueling information has changed from last month:

Yes No X

2. Scheduled date for next refueling: 9/16/95

3. Scheduled date for restart following refueling: 10/31/95

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? Aucust 28, 1995

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

Hgt reauired.

6. Important licensing considerations associated with refueling:

HlA

7. Number of Fuel Assemblies:

A. Incore 2.51 B. In Spent Fuel Storage (prior to refueling) 1240 C. In Spent Fuel Storage (after refueling) 1472

8. Present licensed spent fuel storage capacity: 4006 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:

(Does alloy for full-core offload)

(Assumes 244 bundle reloads every 18 months until then)

(Does n2t allow for smaller reloads due to improved fuel)

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l HOPE CREEK GENERATING STATION  !

MONTHLY OPERATING

SUMMARY

September 1994 Hope Creek entered the month of September shut down due to a. plant trip on August 30th caused by a loss of stator water cooling. The plant was restarted on September 1, 1994. After attaining'100%

power on September 2, 1994, power was reduced greater than 20% on  !

September 3, 1994 to allow for rod adjustment and Combined Intermediate Valve testing. The plant was operated at 100% for  ;

the remainder of the month. As of September 30, 1994 the plant ,~

has been on line for 30 consecutive days.

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SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS '

FOR THE HOPE CREEK GENERATING STATION September 1994 i 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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l TemoorarV Modification Summary of Safety Evaluation T-Mod 94-024: This Temporary Modification installed an electrical jumper across the #2 Feedwater Heater Hi-Hi Level trip switches and installed a temporary keep fill line to the low side of the level transmitters. This modification is performed due to spurious indications during power ascension and is removed at approximately 40 % Reactor Power. This T-Mod does not increase the probability or the consequences of an accident listed in Table 15.0-2 of the UFSAR since the worst case would be for water induction into the turbine resulting in a turbine trip.

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 94-026: This temporary modification is a change to the normal line-up of the test line for the MSIV Seal System return to PSIG. It includes the addition of a drain hose to the drain leg of the test return line valve. The system is designed to utilize this point as a drain. The only change is from the normal system valve line-up and the addition of the drain hose.

This modification changes the plant as described in the UFSAR in that the P&ID describes the normal alignment of the system without a drain hose as shown in the UFSAR.

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 94-027: This Temporary Modification raises the reset point for the flow switch in the SACS system. The flow switch is  !

causing spurious trips when the "C" pump is switched from auto to I manual when the "B" loop of SACS is supplying TACS. The trips are being caused by momentary pressure oscillations in the SACS system which, when combined with a suspected leakage past the 1EG-HV-2522A & C valves, is causing a pressure spike, which resets the flow switch and then when the spike disappears, causes the flow switch to trip.

This T-Mod will move the reset point high enough to prevent the switch from improperly resetting and will allow the system to function properly until the suspected valve leakage can be l corrected. This modification has sufficient margin to the normal flow level and will not impact normal system operation.

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.

I I-H2d 94'-028: This Temporary Modification creates an automatic control loop for operating the Auxiliary Boiler Steam Dump Valve.

The purpose for this modification is to limit the amount of steam exhausted to the atmosphere as this will reduce fuel consumption and increase boiler efficiency. The present configuration does not automatically manipulate the dump valve open or closed based on load demand since the air signal is input manually in order to set valve position. As a result, the valve is opened continuously exhausting steam to the atmosphere.

This modification attempts to automate the operation of the dump valve by utilizing steam header pressure as a process input signal to control the dump valve open or closed. Automatic operation of the dump valve is achieved by placing the Selector Switch in VENT Position to control the dump valve position utilizing the Auxiliary Steam Header Pressure.

The auxiliary steam system has no safety related function. The system is designed so that a failure of the system or component does not compromise any safety related system or component or prevent a safe reactor shutdown.

Therefore, this Temporary Modiff ' tion does not increase the probability or consequences of accident previously described in the SAR and does not involve an areviewed Safety Question.

T-Mod 24-029: This Temporary Modification installed an electrical bypass jumper on "D" Service Water Traveling Water Screen (TWS)

Spray Wash Booster Pump flow switch. This jumper will simulate a flow signal to allow the "D" TWS to operate until a replacement flow switch can be procured and installed. When the TWS is in AUTO, the spray pump will start upon the start of it's associated Service Water Pump. Should there be a failure of this T-Mod the spray wash booster pump will trip which will cause the TWS to stop. Debris will enter the self-cleaning strainer which will remove this debris before reaching the Station Service Water System (SSWS). An alarm will also be generated by CRIDS in the main control room for differential water level on the TWS speed status.

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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H-1-ZZ-N'SE-0811: This evaluation addresses on a generic basis a subset of proposed temporary modifications that may be implemented for the purpose of bypassing control room overhead annunciator input signals which have alarmed and cannot be immediately j corrected by maintenance. The alarm represents a nuisance distraction to control room operators, may mask other contributions to the =larm window and provide no useful information to the operator since the initiating condition has been recognized and appropriate action to correct the condition would be initiated in accordance with applicable alarm response procedures and the work control program. Additionally, Operations procedures provide for marking of alarm windows which have inoperable or modified inputs.

This subset of T-Mods will be controlled by the Temporary Modification Procedure which requires Licensed Operator approval, independent verification, System Engineering review, and appropriate periodic management review to ensure that the T-Mods are adequately administered and controlled.

Therefore, this generic Safety Evaluation for this subset of Temporary Modifications does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.

Procedure Summary 2f Safety Evaluation NC.NA-AP.ZZ-0047(Z) Rev 11 This procedure titled " Nuclear Support Facility Management Program", has been deleted and replaced by a Nuclear Department Business Procedure NC.NA-BP.ZZ-003(Z) to cover this information. The operational Quality Assurance program does not apply to this procedure. Although this procedure is briefly described in the UFSAR (Section 13.5.1 Hope Creek Plant Procedures) it performs no function relative to the safe operation of Hope Creek.

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.

Other . ' Summary gf Safety Evaluation UFSNR CH #94-36: This UFSAR Change Notice addresses the Structural Steel Fireproofing Tables (9A-103 and 9A-104) which are being deleted in the UFSAR in favor of direct reference to the Structural Steel Fireproofing Drawings (Architectural Drawings A-0241 through A-248). The information contained within the Tables is redundant to the information contained on these Drawings.

Elimination of these tables will simplify the revision process, and reduce to one source for room fireproofing status. Text changes also included are mainly editorial in nature to make the information contained in the fireproofing program consistent with the information described on the design drawing. There are no parameters or systems affected by this change.

Therefore, this UFSAR Change Notice 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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