ML20072C301

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

ML20072C301
Person / Time
Site:	Hope Creek
Issue date:	07/31/1994
From:	Hovey R, Zabielski V Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9408170175
Download: ML20072C301 (11)

Text

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2PSEG Pubhc Service Electric and Gas Cornpany P.O. Box 236 Har, cocks Brid e, New Jersey 08038 L

Hope Creek Generating Station August 12, 1994 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 July are being forwarded to you with the summary of changes, tests, and experiments that were implemented during July 1994 pursuant to the requirements of 10CFR50.59(b).

Sincerely yours, R.

J.

Hove General Manager -

Hope Creek Operations V h*

3+DR:WS:JT Attachments C-Distribution i

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PDR ADOCK 05000354 R

PDR

INDEX l

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 i

l OPERATING DATA REPORT DOCKET NO.

50-354 UNIT Hope Creek DATE 8/09/94 COMPLETED BY V. Zabielskid j

TELEPHONE (609) 339-3506 OPERATING STATUS M

1.

Reporting Period July 1994 Gross Hours in Report Period 714 2.

Currently Authorized Power Level (MWt) 3293 j

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 744.0 3735.3 56558.3 6.

Reactor reserve shutdown hours 91Q 0.0 0.0 7.

Hours generator on line 744.0 3643.9 55676.4 8.

Unit reserve shutdown hours 0.0 0.0 0.0 9.

Gross thermal energy generated 2442151 11645112 177608481 (MWH)

10. Gross electrical energy 794120 3855790 58819744 generated (MWH)

11. Net electrical energy generated 759408 3677051 56204735 (MWH)

12. Reactor service factor 100.0 73.4 84.7

13. Reactor availability factor 100.0 73.4 84.7

14. Unit service factor 100.0 71.6 83.4

15. Unit availability factor 100.0 212 6 83.4

16. Unit capacity factor (using MDC) 99.0 70.1 81.7

17. Unit capacity factor 95.7 67.7 78.9 (Using Design MWe)

18. Unit forced outage rate 0.0 4.4 4.4

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

B OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO.

50-354 UNIT Hope Creek DATE 08/09/94 V b'[j d

COMPLETED BY V.

Zabielski TELEPHONE (609) 339-3506 1

MONTH July 1994

]

METHOD OF SHUTTING DOWN THE TYPE REACf0R OR F= FORCED DURATION REASON REDUCING CORRECTIVE NO.

DATE S= SCHEDULED (HOURS)

(1)

POWER (2)

ACTION / COMMENTS 1

7/10 F

0 A

5 Power reduced to 65% to repair leak on steam seal evaporator shell drain.

2 7/14 F

0 H

5 Lightning strike caused transient which resulted in multiple FWH trips.

Power reduced to 80%, tripped FWHs were restored, and unit was returned to 100%.

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AVERAGE DAILY UNIT POWER LEVEL i

DOCKET NO.

50-354 UNIT Hope Creek DATE 8/09/94 COMPLETED BY V.

Zabielski IU TELEPHONE (609) 339-3506 MONTH July 1994 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 1.

J034 17.

1011 2.

1029 18.

1028 3.

LQll 19.

1024 4.

1A22 20.

1017 5.

1033 21.

1021 6.

1021 22.

1022 7.

1021 23.

1016 8.

1035 24, 1025 9.

1211 25.

1025 10.

930 26.

1026 11.

1922 27.

101*.

12.

1222 28.

1030 13.

1Q30 2 5'.

1030 14.

1006 30.

1020 15.

1021 31.

1007 26.

1033

~

REFUELING INFORMATION DOCKET FO.

50-354 UlIT Hone Creek 1 DATE Julv 09. 1994 COMPLETED BY V.

Zabielski Y

TELEPHONE (609) 339-3506 MONTH July 1994 1.

Refueling information has changed from last month:

Yen 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 5.

Scheduled date(s) for submitting proposed licensing action:

HIA 6.

Important licensing considerations associated with refueling:

U91M 7.

Number of Fuel Assemblies:

A.

Incore 26A 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 4/28/2006 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)

(Does not allow for smaller reloads due to improved fuel) 1 5

)

l HOPE CREEK GENERATING STATION MONTHLY OPERATING

SUMMARY

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

The unit operated at full power through the 10th of July when power was reduced to 65% to repair a leak on the Steam Seal Evaporator shell. It returned to 100% power that same day.

on the 14th of July a lightning strike caused a transient which resulted in multiple trips of Feedwater Heaters.

Power was reduced to 80%

until the tripped FWH's were restored later that day.

The Unit returned to 100% for the remainder of the month without any further power reductions or plant trips.

As of July 31, the plant had been on line for 40 consecutive days.

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SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION July 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.

Desian Chance Packace Summar_y of Safety Eyaluation LEC-3348 Pka Li The purpose of this DCP is to verify operation of the floor drain waste processing system.

Additionally, it identified / procured the materials necessary to produce solidified (bitumenized) samples of the simulated floor drain concentrates.

It utilizes the existing extruder / evaporator and auxiliary systems for the tests.

The original plant design included provisions for processing of floor drain waste products by evaporation and solidification.

The resultant floor drains concentrates will be solidified in bitumen using the extruder / evaporator.

The equipment for doing this was installed during plant construction, but except for the extruder / evaporator, has not been fully tested and turned over to operations.

The proposed tests do not alter the present Solid Rad Waste System design configuration or operating procedures.

The test uses non-radioactive waste solutions (which may have an insignificant amount of contamination from residual activity in the bottoms tank used to store the sample solution) for its test medium.

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

AllE-0108 : This change will install a set of normally closed contacts, from the local keylock control switch for the RHR pump 1AP202 suction valve, in series with the control room open/close control circuit.

This change modifies the control circuit for the RHR suction valve as shown on figure 7.3.7 of HC UFSAR.

The change does not reduce the margin of safety as described in the basis for any Tech Spec.

This is based on the fact that the new manual keylock switch functions the same as the existing keylock switch.

The new switch has an additional feature which will prevent the valve opening or closing from the main control room when an operator at the MCC is operating the valve.

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

ADE-0133: This Design Change will replace a 3 foot wide 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated fire door on 77'elv. in the Turbine building common area.

It will be changed out with an 8 foot wide double door with a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rating.

There is no effect of this modification on plant systems and components other than to allow larger items to be moved into a temporary storage in the Unit 2 area.

4 The installation of this larger door is considered a change to the facility as it will require the appropriate UFSAR figures to be updated.

Therefore, this DCP 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 of Safety Evaluation NC.NA-AP.ZZ-0034(Z) Rev 11 This Nuclear Administrative Procedure

" Performance Indicator Program" is being deleted. The performance Indicator Program has been incorporated into the Business Plan.

The Operational Quality Assurance Program does not apply to this procedure.

Although this procedure is briefly described in the relativ(Section 13.5.1),

UFSAR Plant Procedures, it performs no function e to the safe operation of Hope Creek Station.

Therefore deleting this procedure will have no affect on the safe operation of Hope Creek.

Therefore, this Procedure deletion does not increase the probability or consequences of an accident previously described in the SAR and does not involve an Unreviewed Safety Question.

HC.OP-SO.AE-0001(O):

This procedure revision is for the operation of the Hope Creek Generating Station using a Two Feed Pump Configuration.

Although two feed pump operation is already identified in this Operating Procedure, the maximum power level allowable is not identified.

The procedure indicates that a reactor feed pump may be removed from service if all three pumps are in service and one is no longer required for Reactor Vessel level control and Reactor Power <95%.

It does not indicate if Reactor Power can then be raised to 100%.

Analysis of the entire Condensate /Feedwater train has proven the acceptability of 100% operation with only two of the three feedwater pumps.

Maximum feed rate possible in this configuration is approximately 116%.

This feed rate is bounded by current analysis for Feedwater Controller Failure Maximum Demand event which assumes three feedwater pump operation.

In two feed pump configuration, feed pump turbine speed for 100% power is approximately 5100 RPM which still provides over 380 RPM until the overspeed clamp is reached at 5480 RPM.

This gives a margin to compensate for any transient which may be seen.

Therefore, this Procedure revision does not increase the probability or consequences of an accident previously deccribed in the SAR and does not involve an Unreviewed Safety Question.

Other Summary of Safety EvaluatiqD UFSAR Chance for Sectiom 17.2:

This UFSAR change revises chapter 17.2 " Quality Assurance During the Operation Phase".

The change reflects current organization structure, editorial enhancements and clarifications and changes selected QA oversight practices for monitoring implementation of the Nuclear Department Procedure System.

There are no credible failure modes 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.

H-1-BG-MSE-0835:

This is to correct the minimum available NPSH for the Main Cleanup Recirculation Pump in table 5.4-3 of the Hope Creek UFSAR.

This will change the stated minimum available NPSH from 14 feet to 16.5 feet.

Additionally this editorial change will bring table 5.4-3 in line with the Design, Installation, and Test Specification for Reactor Water Cleanup System, and the Heat exchanger Instruction Manual.

This editorial Change will not change any of the credible failure modes.

This is due to the fact that this change will bring table 5.4-3 in accordance with as built and design conditions.

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.

}{-1-BG-MSE-0 8 3 6 :

This is a correction of the stated shell side design pressure for the non-regenerative heat exchanger in table 5.4-3 of the Hope Creek UFSAR.

This editorial change will only change the stated shell side design pressure from 1475 psig to 150 psig on table 5.4-3 in line with the Design, Installation, and Test Specification for Reactor Water Cleanup System, and the Heat exchanger Instruction Manual.

This editorial Change will not change any of the credible failure modes.

This is due to the fact that this change will bring table 5.4-3 in accordance with as built and design conditions.

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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