ML20094E257

Monthly Operating Rept for Dec 1991 for Hope Creek,Unit 1

ML20094E257
Person / Time
Site:	Hope Creek
Issue date:	12/31/1991
From:	Zabielski V Public Service Enterprise Group
To:
Shared Package
ML20094E260	List: ML20094E257 ML20094E279 ML20094E283
References
NUDOCS 9201220306
Download: ML20094E257 (13)

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. AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-354 UNIT Hone Creek ,

DATE 1/15/92 COMPLETED BY V. Zabielski TELEPHONE (609) 339-3506 MONTH- December 1991 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER-LEVEL (MWe-Net) (MWe-Net) ,

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1. 1QS.Q 17. 1070

2. 1063 18. 1055

3. 1052 19. 1pJl

.4, 1058 20. B.02

5. 1058 21. 321

6. 1058 22. 253.

'7. ,93 1 23. 12,5

'8. . 113. 24. 1061

9. 211 25. 1Q51

.10 . 125 26. 1Q11

11. 114 27. 1060

12. 52.Q 28. J'jj,1

13. S3l 29. 1056

14. 1Q14 30. 1058

15. 1057 31. 1061

16. 1052 i:

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.. 1 OPERATING DATA REPORT DOCKET No. 50-354 UNIT Hope Creek DATE 1/15/92 COMPLETED BY V. Zabielski TELEPHONE (609) 339-3506 OPERATING STATUS

-1. Reporting Period December 1991 Gross Hours in Report Period 211

-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) Norse

4. Reasons for restriction (if any)

This Yr To Month Qatn Q.umulative

• 5.- No. of hours reactor was critical 744.0 7379.8 37.161.3

6. - Reactor reserve shutdown hours 929 229 222

7. Hours generator on line 21122 7281.5 36 574.6

8. Unit reserve shutdown hours Q2A 229 atQ

9. Gross thermal energy generated 2.233.960 23.454.735 115.997.142 (MWH)

10. Gross electrical energy 742.460 7.730.821 23 352.494 generated (MWH)

11. Net electrical energy generated 711.146 7.394.865 36.651.549 (MWH) la. Reactor service factor 100.0- R121 84,2 13.-Reactor availability factor 100.0 84.2 84.2

14. Unit service f ctor. 100.0 83.1 82.9 15.' Unit availability factor 100.0 83.1 82.9

16. Unit capacity factor (using MDC) 2222 Bitt 'aQth

17. Unit capacity factor 89.6 79.1 77.9 (Using Design MWe)

18. Unit forced outage rate AxQ 121 Ezl

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

. :one

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

N/A-

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.- REFUELING INFORMATION DOCKET NO. 50-354 UNIT Hope Creek DATE 1/15/92 COMPLETED BY S. ~ HollinaswpI,th TELEPHONE (609) 339-1051 MONTH December 1991

1. Refueling information has changed from last month:

Yes No X

3. Scheduled date for next refueling: 9/12/92

3. Scheduled date for restart following refueling; 11/11/92

4. A. Will Technical Specification changes or other license amendments be required?

Yes No  %

B. Has the reload fuel design bcen reviewed by the Station Operating Review Committee?

Yes No X If no,-when is it scheduled? not scheduled (on or prior to 7/24/92)

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

6. Important licensing considerations associated with refueling:

- Same fresh fuel as current cycle: no-new considerations

.7. ---Number of! Fuel Assemblies:

A. Incore- 764-B. In Spent Fuel. Storage (prior t.o refueling) 760 C. In Spent' Fuel Storage.(after refueling) 1008

8. Present licensed spent fuel storage capacity: 4QQ5 Future spent fuel storage capacity: 4006

9. Date of last refueling that can be discharged 11/4. 2010 to spent fuel pool assuming the present. (EOC16) licensed capacity:

OPERATING DATA REPORT

-UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO. 50-354 UNIT Hope Creek DATE 1/15/92 COMPLETED BY V. Zabielski TELEPHONE (509) 339-3506 MONTH December 1991 METHOD OF SHUTTING DOWN THE TYPE REACTOR OR F= FORCED DURATION REASON REDUCING CORRECTIVE NO. DATE S= SCHEDULED (HOURS) (1) POWET, (2) ACTION / COMMENTS 9 12/7 F 0 A 4 Condenser Air Leak due to crack in Steam Ueal Evaporator Inlet Relief Piping.

10 12/20 F 0 A 4 Full Recirc Runback caused by failure the 'C' Primary condensate Pump Lube 011 Supply Line.

11 12/23 F 0 A 4 Moisture Separator Leak.

Summary l

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HOPE CREEK GENERATING STATION MONTHLY OPERATING

SUMMARY

-December 1991 i

Hope Creek entered the month of December at approximately 100%

power. On December 7, power was reduced because of a Condenser- ,

Leak caused by a crack in the Steam Seal Evaporator Inlet Relief Piping. Power was. restored to approximately 100%. On December-20, there was a full Recirc Runback caused by a failure of the 'C' Primary Condensate Pump Lube Oil Supply Line, which reduced power.

Thc unit remained at reduced power until a Moisture Separator Manway leak was repaired on December 23. On December 23, power was restored to approximately 100%. The unit operated for the remainder of the month without experiencing any -hutdowns or any other reportable power redections. On Decembec 3, th completed its 234th day of continuous power operatlon.e plant

- This surpasses the station's previous record of 221 days. ,

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SUMMARY

OF-CHANGES, TESTS, AND--EXPERIMENTS FOR THE HOPE CREEK GENERATING STATION DECEMBER 1991 P

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. . 4 The'following items have been evaluated to determines

- 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 nargin of safety as defined in the basis for any i 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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e a QCE Descriotion of Safety Evaluation 4EC-3112/10 This DCP added conduit, power distribution 1 equipment, and valve pit sump pump discharge piping to the yard area. This DCP also installs lighting, power distribution, and nump pumps with discharge piping in Service Water System Valve Pits. Heat Tv=.cing with pipe insulation is being added to a fire protection line that was originally buried.

The cathodic protection system components in a valve pit are being relocated to preclude damage during construction and maintenance of the valve pit.

No Unreviewed Safety Questions were involved because it j, all non-safety related equipment that l does not interface with the Station Service Water l System or any electrical sysc.im T3chnical i Specification.

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4EC-3115 This DCP replaced existing carbon steel piping in the Service Water System with now material of I piping class HZD, 6% molybdenum stainless steel.

The new piping is more resistant to corrosion and erosion. This DCP also added larger capacity Service Water Dewatering Drain Headers and a larger capacity Service Water Drain Tank to shorten the time required for dowatering and to preclude the

, manual cycling of the dewatering pump.

The new Service Water piping has been designed to the criteria as the existing piping. The new piping is more rusistant to corrosion and erosion.

This DCP does not affect the safety-related function of the Service Water piping. Therefore, no Unreviewed Safety Questions were involved.

4EC-3204 This DCP installed a spring support in place of a vertical rigid pipe support in the High Pressure Coolant Injection System. Substituting the spring support in place of the rigid pipe support will allow pipe movement caused by thermal bowing during system warmup.

No Unreviewed Safety Questions are involved with this DCP because the new analy'.ed pipe stresses are lower than the previous pipe stresses. Also, all piping, supports, and componento are within the allowable tolerances.

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. IM31 Description of Safety Evaluation 4EC-3211 This DCP modified the start logic to the Deep Well Pumps so the Nuc1 car Department Administrative Building will not run out of fresh water during normal conditions. A new low level switch contact was added to provide a signal to the Deep Well Pump Start Control circuitry to start the lead Deep Well Pump and send a si nal to open the Domestic Water Storage Tanks Supp y Valves. This will reduce the number of Deep Wel Pump starts and ensure that both the Domestic Water Storage Tanks and the Nuclear Department Administrative Building Fresh Water Storage Tank are being filled.

This DCP does not have an adverse effect on the operation of any of the affected systems, neither does it involve any safety-related systems or equipment. Therefore, no Unreviewed Safety Questions were involved.

4HM-0G27 This DCP installed a 1" drain line and associated valves in the ' B' Primary Containment Instrument Gas Compressor Test Return Suction Line. This test return line is only used during an 18 month surveillance test. The drain valves will be closed during normal operation.

The drain line and associated valves are passive during the use of the test return line. Being in the closed position during the test will not affect the Loss of Coolant Accident isolation requirements of the test return line. This DCP does not impact the normal operation of the system; therefore, no Unreviewed Safety Questions were involved.

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IMB. Descrintion of Safety EvaluatiQD 91-055 This TMR disabled a high temperature switch input to the Offgas Panel Trouble Annunciator in the Main Control Room. At periods of low flow, heat from the reheater raised the temperature above the setpoint even though the exit temperature from the cooler condenser was at design conditions. This resulted in a nuisance alarm that has been eliminated by this TMR.

Moisture and temperature are monitored downstream of this temperature element at the reheater. If any moisture er temperature problems exist, they will be detected and alarmed at that point; therefore, no Unreviewed Safety Questions were involved.91-056 This TMR authorizes the use of a 10 amp fuse in the 125VDC control power supply feed to the 250VDC High Pressure Coolant Injection System Motor Control Center. The vendor document indicates that a 30 amp fuse should be used. It is not apparent if the 10 amp fuse was supplied by the vendor or inadvertently installed after the Motor Control Center-was placed in operation.

An analysis-was performed that showed the 10 amp fuse to be of an adequate size for the High-Pressure Coolant Injection System tt perform its intended safety functions. Therefore, no Unreviewed Safety Questions were involved.91-057 This TMR authorizes the use of a 10 amp fuse in the 125VDC control power supply feed to the.250VDC Reactor Core Isolation Cooling System Motor Control Center. The vendor document indicates that a 30 amp fuse should be used. It is not apparent if the 10 amp fuse was supplied by the vendor.or.

inadvertently installed after the Motor Control Center was placed in operation.

An analysis was performed that showed the 10 amp fuse to be of an adequate size for the Reactor Core Isolation Cooling System to perform its intended safety functions. Therefore, no Unreviewed Safetu Questions were involved.

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DR Descriotion of Deficiency Report HTE 91-196 -The Reactor Building. Lightning Mast toppled from itJ mounting on top of the Secondary Reactor Containment Dome. It remained lying horia ntally on top of the building, supported by the surrounding railing until it was removed.

Various roof areas that could be-impacted by the lightning mast were analyzed and found to be able to sustain impact from the lightning mast. Probability analysis indicates that only one direct lightning strike to a critical building is expected to occur every 5 years.

It has also been qualitatively concluded that there are-no increased risks associated with the lack of the lightning mast for the duration required for replacement. Therefore, no Unreviewed Safety Questions are involved if the new lightning mast is installed prior to 5/31/92.

a Procedure Revision Description of Safety Evaluation VilC. MD-GP. Z Z-02 2 4 (Q) This procedure is a vendor procedure that Rev. O controls the use of the Valve operation Test and Evaluation System for the Residual Heat Removal Discharge to Radwaste Isolation Outboard Valve. This 1s-a non-intrusive analysis system used to evaluate the performance of motor-operated valves.

The Valvo Operation Test and Evaluation System is a non-intrusive analysis system that does not require any alterations to the valve. There is no safety impact caused by attaching the force sensor to the yoke of the valve. Because no alterations are made to either the characteristics or the internals of the valve, this procedure toes not involve an Unreviewed Safety Question.

f Tech Spec

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UFSAR Section Descrintion of Safety Evaluation 9.2.2.4 This UFSAR Change Notice addresses the use of the 1983 edition through the summer of 1983 addenda of ASME Section XI for the Hope Creek InservicenTesting Program. Previously, the UFSAR referenced the 1977 edition through the summer of 1978 addenda.

No Unreviewed Safety Questions were involved because Hope Creek has complied with the 1983 edition and summer addenda as previously submitted to the NRC in the Safety Evaluation Reoort. The two codes are the same concerning th'e performance of inservice testing.

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