ML20070F196

Monthly Operating Rept for June 1994 for Hope Creek Generating Station,Unit 1

ML20070F196
Person / Time
Site:	Hope Creek
Issue date:	06/30/1994
From:	Hovey R, Zabielski V Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9407180216
Download: ML20070F196 (11)

Text

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O PSLEG Pubhc Service Electnc and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038 ,

Hope Creek Generating Station I

l July 14, 1994 i

U. S. Nuclear Regulatory Commission i Document Control Desk (

Washington, DC 20555 '

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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 June are being forwarded to you with the summary of changes, tests, and experiments that were implemented during June 1994 pursuant to the requirements of -[

i 10CFR50.59(b). I i

Sincerely yours, (

R. J. Hovef 'l General Manager -  !

Hope Creek Operations D :WS:JC >

Attachments  !

C Distribution  ;

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The Energy People 9407180216 940630 a.gn ggg, PDR ADOCK 05000354 i

.R PDR

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. INDEX l NUMBER SECTION OF PAGES Average Daily Unit Power Level. . . . . . . . . . . 1 l Operating Data Report . . . . . . . . . . . . . . . 3 j Refueling Information . . . . . . . . . . . . . . . 1 ;

Monthly Operating Summary . . . . . . . . . . . . . 1  !

Summary of Changes, Tests, and Experiments. . . . . 3 l i

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, OPERATING DATA REPORT DOCKET No. 50-354 UNIT Hope Creek DATE 07/11/94 l COMPLETED BY V. Zabielski i TELEPHONE (609) 339-3506 OPERATING STATUS

1. Reporting Period June 1994 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)

This Yr To Month Date Cumulative

5. No. of hours reactor was critical 720.0 2991.3 55814.3

6. Reactor reserve shutdown hours 0.0 0.0 0.0

7. Hours generator on line 704.4 2899.9 54932.4

8. Unit reserve shutdown hours 0.0 0.0 0.0 l

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9. Gross thermal energy generated 2256004 9202960 175166330 (MWH)

10. Gross electrical energy 746960 3061670 58025624 generated (MWH)

11. Net electrical energy generated 714236 2917643 55445327 l (MWH) 1

12. Reactor service factor 100.0 68.9 84.6

13. Reactor availability factor 100.0 68.9 84.6

14. Unit service factor 97.8 66.8 83.2 l

l 15. Unit availability factor 97.8 66.8 83.2

16. Unit capacity factor (using MDC) 96.2 65.2 81.5

17. Unit capacity factor 93.0 63.0 78.7 (Using Design MWe)

18. Unit forced outage rate 222 115 4.5

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 1

l OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-354 UNIT Hope Creek DATE 07/11/94 -t 7 COMPLETED BY V. Zabielski TELEPHONE (609) 339-3506 MONTH June 1994 i

METHOD OF SHUTTING DOWN THE I TYPE REACTOR OR l F= FORCED DURATION REASON REDUCING CORRECTIVE- )

NO. DATE S= SCHEDULED (HOURS) (1) POWER (2) ACTION / COMMENTS 1 06/21 F 15.6 A 1 EHC LEAK ON #2 i BYPASS VALVE REPAIRED.

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AVERAGE DAILY UNIT POWER LEVEL 1

DOCKET NO. 50-354 UNIT Hope Creek , l DATE 07/11/94 4*

COMPLETED BY V. Zabielski V I TELEPHONE (609) 339-3506 l

i MONTH June IEli DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWo-Net) (MWe-Net)

1. 103S 17. 1046

2. 1050 18. 1033

3. 1050 19. 1016

4. 1051 20. 1033 l

5. 1040 21, 344

6. 1031 22. 415

7. 1031 23. 1021

8. 1050 24. 1028

9. 1049 25. 1028

10. 1042 26. 1030

11. 1042 27. 1028

12. 1035 28. 1035

13. 1034 29. 1030

14. 1032 30. 1036

15. 1024 31. n/a

16. 1026 i

REFUELING INFORMATION DOCKET NO. 50-354 UNIT Hope Creek 1 DATE July 11. 1994 COMPLETED BY V. Zabielski TELEPHONE (609) 339-3506 MONTH June 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? Not scheduled yet i

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

H21 scheduled vet.  ;

6. Important licensing considerations associated with refueling:

Ulh  ;

7. Number of Fuel Assemblies:

1 A. Incore 764 i B. In Spent Fuel Storage (prior to refueling) J240 C. In Spent Fuel Storage (after refueling) 1472

8. Present licensed spent fuel storage capacity: 4006 l 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) 1 licensed capacity:

(Does allow for full-core offload)

(Assumes 244 bundle reloads every 18 months until then)

(Does D21 allow for smaller reloads due to improved fuel) 1.

l MONTHLY OPERATING

SUMMARY

June 1994 Hope Creek entered the month of June at approximately 100% power.

The unit operated at full power through June 21 when the station experienced a hydraulic control oil leak on turbine bypass valve No. 2. The plant was placed in operating condition 2 Cstartup) and the bypass valve was removed from service for repairs. After repairs were completed, the plant was returned to operation condition 1 (Power Operation). On June 22, the main generator was synchronized to the grid. The plant operated at full power through the remainder of the month. As of June 30 the unit has been on line for 8 consecutive days.

SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION June 1994 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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Procedure Summary of Safety Evaluation NC.NA-AP.ZZ-0024(O): Nuclear Administrative Procedure " Radiation Protection Program" has been revised to incorporate specific changes made to several policies and procedures. This revision incorporates the requirements of the revision to 10CFR20

" Standards for Protection Against Radiation" and related policies.

This revision constitutes a change to the SAR because certain specifications of the revision disagree with certain specifications in the SAR.

This revision establishes Controlled Area outside of the Protected Area, and it institutes the posting of Very High Radiation Areas with associated required access controls. These changes constitute a change to the facility as described in the SAR.

10CFR20.1008 allows implementation of the revised 10CFR20 requirements without Technical Specification Revision. Applicable sections of the revised 10CFR20 are to be used in lieu of any part of the old 10CFR20 cited in license conditions or Technical Specification.

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.

EPIP 205H: This is a revision to the Emergency Plan Implementing  !

Procedure (EPIP) "TSC-Post Accident Core Damage Assessment". It l is intended to supersede Hope Creek Chemistry implementing i procedure HC.CH-T1.ZZ-0011(Q) " Estimation of Reactor Core Damage l Under Accident Conditions".

The superseding procedure does not change the intent or methodology relative to the assessment activities as described in the deleted chemistry, procedure. There were no new requirements introduced for radiation / chemistry analysis or response.

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.

i Temoorary Modifications Summary 21 Safety Evaluation T-Mod 94-015: This Temporary Modification installs a test fixture in front of the service water pipe that supplies make-up water to the circulating water system. Installation of this test fixture will not alter the operation of the system, component or structure. The specific location of the fixture prevents the fixture or its failure to have any affect or come in contact with any other component, system, or structure. If any piece should part from the fixture, there are physical barriers installed that would inhibit the pieces from entering the suction of the circulating water pumps.

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 94-016: "B" Reactor Recirculating Pump vibration probe cables were inadvertently switched upstream of their respective proximiters. This occurred during reinstallation of instrumentation originally removed to support seal work on the pump. Due to the inaccessibility of the probes during normal operation, the proposed Temporary Modification will roll the power, signal and shield leads of the associated vibration transmitters. Vibration instruments for both circuits are identical. The modification is functionally equivalent to the original wiring configuration and will provide true information at the Vibration Monitors.

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 94-017: 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 j 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.

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i T-Mod 94-018: This Temporary Modification installs flow transmitters for the purpose of measuring the reactor feedwater i flowrates as compared to the monitored parameters used to j determine core thermal power. The purpose of this comparison is 1 to ascertain if a lack of flow to the vessel could be contributing to a decrease in Plant electrical output.

An instrument tubing failure could potentially cause an increase ~

I in flow and differential pressure across the flow element. This would result in higher indicated flow. However, because the tubing is 3/8" as compared to the flow element which is 24", the .

indicated flow increase would have little impact. '

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 Other Summary of Safety Evaluation  ;

UFSAR Section 6.2: The UFSAR change provides clarification pertaining to primary containment leakage rate testing. The need for this clarification became evident during the performance of the Containment Integrated Leak Rate Test (ILRT) in the fifth refueling outage. The changes will clarify the description of containment leak rate testing contained in section 6.2 by making it technically more precise. i This revision will not result in any changes to plant systems or procedures. The method of performing containment leak rate testing and evaluating the test results will remain unchanged as do the applicable acceptance criteria. There are no credible failures associated with the change.

Therefore, this UFSAR change 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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