ML20237J926

Monthly Operating Rept for Jul 1987

ML20237J926
Person / Time
Site:	Hope Creek
Issue date:	07/31/1987
From:	Jensen H Public Service Enterprise Group
To:
Shared Package
ML20237J580	List: ML20237J578 ML20237J926 ML20237J928
References
NUDOCS 8708190002
Download: ML20237J926 (11)

Text

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. AVERAGE DAILY. UNIT POWER LEVEL DOCKET NO.86-354 UNIT Hope Creek Gen. Station DATE 99/97/97' COMPLETED BY H. J en gy4 TELEPHONE (609) 339-5261 uly, 1987

.- MONTH DA1. AVERAGE-DAILY POWER LEVEL DAY- AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1023 997 3 37 906 18 988 .

2 943 3 19 582 20 967 4

977 5 21 984 863 6 22 1015 984 7 23 99 999 8 24 981 753 25 9

987 26 799 10 _

' 920 11 - . . _ _ _ _

27 ___n 992 12 28 968 991 13 29 947 325 14 30 9

15 31

.16 g B1 jog B7 J

p.

OPERATING DATA REPORT DOCKET NO. R6-354' UNIT 3r Hnpn Creek Gen. Sta. ]'

DATE O R /07 / 87 COMPLETED BY H_ aonnen OPERATING ~ STATUS

1. REPORTING PERIOD July 1987 GROSS HOURS IN REPORTING PERIOD 744 2.. CURRENTLY AUTHORIZED POWER LEVEL (MWt) 3291--

MAX. DEPEND.. CAPACITY (MWe-Net) 1061*

DESIGN ELECTRICAL RATING (MWe-Net) '1061-

3. POWER LEVEL TO WHICH RESTRICTED'(IF ANY) ( MWe-Ne t ) __ Nong_

4.- REASONS FOR RESTRICTION (IF ANY)

THIS YR TO MONTH DATE CUMULATIVE

5. NO. OF HOURS REACTOR WAS CRITICAL 7as 6 AA77.6 5165.6

6. REACTOR RESERVE SHUTDOWN HOURS n a a

7. HOURS GENERATOR ON LINE , 7n5 ? __4852 1 5140 3

8. UNIT: RESERVE SHUTDOWN HOURS _

n n n ]

9. GROSS THERMAL ENERGY GENERATED (MWH) 7 17n;soa la_mno ans y 1 s . 514. R13 19 ~ GROSS ELECTRICAL ENERGY GENERATED (MWH) 7mn;1n1 a ; AKR 10; __G 169.745

'11. . NET ELECTRICAL ENERGY GENERATED (MWH) A77,7a7 a . 6 6 0 . 9 41 4.946.415

12. REACTOR SERVICE FACTOR oa n 45 4 96.1

13. REACTOR AVAILABILITY FACTOR oa n on o 46_1

14. UNIT SERVICE FACTOR oA_n on a 49 6 c15. UNIT AVAILABILITY FACTOR oa n os a 45 6

16. UNIT CAPACITY FACTOR (Using Design MDC) 85.3_ _. RR o 86.2

17. . UNIT CAPACITY FACTOR (Using Design MWe) n5.1 _ nn o n6 7

18. ' UNIT FORCED OUTAGE RATE R 7 A A aa

19. SHUTDOWNS SCHEDULED OVER NEXT 6 MONTHS (TYPE, DATE, & DURATION) :

Surveillance, 9/18/87, 16 days Refueling, 1/31/88, 55 days

20. IF SHUT DOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP:

N/A

• Estimated Value. Actual value to be determined in August 1987.

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OPERATING DATA REPORT UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO.86-354 UNIT NAME Hope Creek Gen. Sta.

DATE 08/07/87 COMPLETED BY H. Jensen July 1987 (609)339-5261 REPORT MONTH TELEPHCNE

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METHOD OF SHUTTING ,

I DOWN THE TYPE REACTOR OR F FORCED DURATION REASON REDUCING CORRECTIVE ACTION /

NO. DATE S SCHEDULED (HOURS) (1) POWER (2) COMMENTS I 8 7/0 3 F 0 B 5 Condensete Demineralized Bypass Operation i Test  ;

i 9 7/2 1 F 0 B 5 Reactor Feed Pump Repairs. And Condensate i Demineralized

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Bypass Operation Test I 10 7/30 F 38.8 H 3 Automatic Scram Incorrect Operation Of an Inverter By an equipment operator in ,

preparation for preventive f maintenance 4

SUMMARY

MONTELY OPERATING

SUMMARY

JULY 1987 l

Hope Creek entered the month operating at 100% reactor power. The' unit operated at essentially full load until July 3 when load was reduced to approximately 60% in order to perform a Condensate .

Demineralized Bypass Operation test. After test completion, power I was restored to 100%. On July 24 power was again reduced for Condensate Demineralized Bypass Operation testing. Reactor Feedwater Pump repairs also took place during this' reduction in power. Full power was restored on July.27.

After completing 150 days consecutive power operation, .an automatic reactor scram occurred on July 30. The' automatic scram was initiated by the incorrect operation of an inverter by an equipment operator in preparation for preventive maintenance on i the inverter. The reactor was started up on July 31 and went critical at 2136 of the same day.

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SUMMARY

OF CHANGES, TESTS, AND EXPERIMENTS l l

FOR THE IIOPE CREEK GENERATING STATION l 1

JULY 1987 1

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. i DfE DescriEt12D 91 DB21sn Chanqg Package HC-KE-002 This DCP added a steel Upending Stand for new fuel containers. The stand allows the i container to be placed in the vertical l position, enabling the fuel bundle to be raised ,

.out of the container. This DCP did not create  !

a new safety hazard to the plant nor did it j affect the safe shutdown of the reactor. It did not change the plant effluent releases and ,

did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

4-HMM-86-0704 This DCP was a document-only change to update design drawings to reflect the actual as-built condition of Radwaste Heat Tracing. Updating the design documents did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor. This DCP did not change the plant effluent releases and did not alter the existing environmental impa.t.

Therefore, no unreviewed safety or environmental questions are involved.

4-HME-86-0759 This DCP reterminated wires at the DC Drive Panels so the Centrifuge Metering Pumps "A" and "B" in the Solid Radwaste system would operate as-designed. Reterminating the wires did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.

This DCP did not change the plant effluent releases and did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

4-BMM-86-0904 This document-only change provided a mechanism for updating design documents to reflect the acceptance Class 2, SA312 Type 304/304H pipe i for use as Reactor Vessel Instrumentation i l

piping. Updating the design documents did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.

This DCP did not change the plant effluent releases and did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

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DfE Descripi;1on of pagign Change PackAga 4-EMJ-87-0948 A design drawing incorrectly labeled manifolds ,

in the Residual Beat Removal system. This i document-only c)ange corrected the drawing error and did not create a new safety hazard to 1 the plant nor did it affect the safe shutdown l of the reactor. This DCP did not change the plant effluent releases and did not alter the existing environmental impact. Thereforef no unreviewed safety or environmental questions '

are involved.

4-EMJ-86-0951 This document-only change clarified notes on design drawings. The notes were misleading and could have led to misinterpretation of the operating characteristics of the Feedwater system. Clarifying the notes did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor. This DCP did not change the plant effluent releases and did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

4-EMJ-86-0969A This DCP corrected various legends, mimics, and nameplates in the Main Control Room to conform to design documents and to meet NUREG 0700 requirements. These deficiencies were discovered during an as-built walkdown.

Correcting these deficiencies reduced the probability of operator error and did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.

This DCP did not change the plant effluent releases and did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

4-EME-86-0972 This DCP is a document-only change to identify installed thermocouple as duel element rather than single element. The thermocouple are located in the Residual Heat Removal Heat Exchanger loops. Upgrading the design drawings did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor. This DCP did not change the plant effluent releases and did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

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D.CE. Dggcription gf Design ChaDSR EasliaSR- .j 4-11CJ-86-133 0 This document-only change corrected discrepancies between Nuclear Boiler logic drawings. Upgrading the drawings did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.

This DCP did not change the' plant effluent  !

releases and did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

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Descript12D pf Temporary Modification Saf21Y EYAluat1OD Request 1111B.)_

87-0059 The Temporary Modification Request (TMR) associated with this Safety Evaluation added an Ion Chromatograph to the Balance of Plant Process Sample Panel to assist ,

Chemistry personnel in determining the  !

source of the high conductivity and high sulfates in the Reactor Pressure Vessel.

The sample panel has no control functions or automatic actions associated with it and is used solely for chemical monitoring ,

of the Condensate, Feedwater, and Main  !

Steam Systems. The addition of the Iron Chromatograph did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor. This TMR did not change the plant effluent releases and did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

87-0061 The Temporary Modification Request (TMR) associated with this Safety Evaluation added a temporary valve and a hose connection to the Solid Redwaste system.

This connection will supply flush water and seal water to the pumps from the Demineralized Water system rather than the Condensate Water s" stem. The modification will prever/ the Solid ,

Radwaste system .;. becoming '

contaminated during t. , test period.

Adding the valve and hose did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor. This TMR did not change the plant effluent releases and did not alter the existing environmental impact.

Therefore, no unreviewed safety or environmental questions are involved.

Safety Eyalpation Description of Temporary Modification Request (TMR) 87-0062 The Temporary Modification Request (TMR) associated with this Safety Evaluation replaced the Centrifuge Discharge Chutes in the Solid Radwaste system with metal flashing chutes with sight glasses. This modification allows observation of centrifuge discharge products during system testing and did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.

This TMR did not change the plant effluent releases and did not alter the existing environmental impact.

Therefore, no unreviewed safety or environmental questions are involved.

87-0064 The Temporary Modifications Request (TMR) associated with this Safety Evaluation rewired a field terminal box so that it monitors a single tempera'ure c element in the Cooling Tower Basin instead of averaging all 4 temperature monitors.

Prior to the installation of this TMR, a wiring problem and 2 fr:lty elements led to an error of approximately 7' in the Control Room Integrated Display System computer. This TMR eliminated this error until a permanent design change can be implemented and did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor.

It did not change the plant effluent releases and did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

87-0065 The Temporary Modification Request (TMR) associated with this Safety Evaluation installed a temporary compressor to be uced as a backup to a Station Air Compressor during compressor maintenance.

The temporary installation of a backup compressor did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor. This TMR did not change the plant effluent releases and did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

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I Safety Evaluation Description af Deficiency Report (DR) 87-0057 This Deficiency Report identified a pinhole leak in a section- of Service Water piping. The Safety Evaluation authorized the repair of this pipe by welding a plate over the leak area. This plate will remain in place until the next available outage when either a basemetal weld repair will'be performed or the pipe will be replaced. The . pieces of equipment located closest to the leak are j the Safety Auxiliaries cooling System  :

Pumps. The pumps have drip-proof motors f and. are mounted on pedestals to protect them during flooding. Therefore this 1 deficiency did not create a new safety hazard to the plant nor did it affect the safe shutdown of the~ reactor. 'It did not change the plant effluent releases and j did not alter the existing environmental impact. Therefore, no unreviewed safety or environmental questions are involved.

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87-0058 This Deficiency Report addresses the degradation of pipe wall thickness in the  ;

Service Water piping elbows. immediately downstream of the Safety Auxiliaries Cooling System Heat Exchanger Balancing Valves. The Safety Evaluation authorized the' repair of this pipe by welding a plate over the affected areas. This plate i will remain in place until the next ,

available outage when either a basemetal weld repair will be performed or the pipe will be replaced. The pieces of equipment located closest to the affected areas are the Safety Auxiliaries Cooling System Pumps.

The pumps have drip-proof motors and are mounted on pedestals to protect them during flooding. Therefore this deficiency did not create a new safety hazard to the plant nor did it affect the safe shutdown of the reactor. The deficiency did not change the plant effluent releases and did not alter the existing environmental impact. Therefore, no unreviewe6 safety or environmental questions are involved.

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