ML20070Q917

Rev 1 to Monthly Operating Rept for Nov 1982

ML20070Q917
Person / Time
Site:	Davis Besse
Issue date:	12/15/1982
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML20070Q906	List: ML20070Q903 ML20070Q917
References
NUDOCS 8301270200
Download: ML20070Q917 (11)

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AVERAGE DAILY UNIT POWER LEVEL .

- DOCKETNO. 50-346 W T Davis-Besse Unit 1 DATE December 15, 1982 COMPLETED BY Erdal Caba b TELEPHONE (419) 259-5000 Ext.

196

, MONDI November, 1982 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (gw Net)

F 852 I 37 882 2 F70 gg 881 870 875 f 3 g9 ,

4 . 873 20 875

-5 870 21 873 6 871 874 22 7 871 - 23 877

-3 323 24 878 9 439 25 878  ;

10 . 801 26 878

-II 863 27 877 12 866 28 878 13 872 -

29 878 34 868 30 877 15 877 3; ___

16 879 INSTRUCTIONS On this format, list the average daily unit power level in MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.

(9/77) kkIhohkO O R

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OPERATING DATA REPORT DOCKET NO. 50-346 DATE December 15, 1982 COMPLETED BY Erdal Caba TELEPHONE m 9) 259-5000, OPERATING STATUS Ext. 196

' l. Unit Name: Davis-Besse Unit 1 Notes

. 2. Reporting Period: November, 1982

3. Licensed Thermal Power (MWt): 2772

4. Nameplate Rating (Gross MWe): 925

5. Design Electrical Rating (Net MWe): 906

6. Maximum Dependable. Capacity (Gross MWe): 918 *

7. Maximum Dependable Capacity (Net MWe): 874 ,

8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report. Give Reasons:

9. Power Level To Which Restricted. If Any (Net MWe):

10. Reasons For Restrictions.If Ar.y:

This Month Yr to.Date Cumulative

- !!. Hours In Reporting Period 720 8,016 38,017

12. Number Of Hours Reactor Was Critical 71 L 6 1.921.5 20.151.5

13. Reactor Reserve Shutdown Hours 6.4 29.4 3,364.1

14. Hours Generator On.Line 701.1 3,780.0 19,L30.2

15. Unit Reserve Shutdown Hours 0.0 1.1 1,732.5

16. Gross Thermal Energy Generated (MWH) 1.889.111 8.324.585 43.446.110 .

17. Gross Electrical Energy Generated (MWH) 634.838 ,_, 2.776.805 14,459,006

18. Net Electrical Energy Generated (MWH) 603,233 2,604,208 13,501,493

-19. Unit Se vice Factor 97.3 47.2 50.1

20. Unit Availability Factor 97.3 47.2 54.6

21. Unit Capacity Factor (Using MDC Net) 95.9 37.2 40.6 l 22. Unit Capacity Factor (Using DER Net) 92.5 35.9 39.2

l. 23. Unit Forced Outage Rate 2.6 1.3 21.2

24. Shutdowns Scheduled Over Next 6 Months (Type. Date.and Duration of Each):

l l 25. If Shut Down At End Of Report Period, Estimated Date of Startup:

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26. Units In Test Status (Prior to Commercial Operation): Forecast Achiesed l- INITIAL CRITICALITY l INITIAL ELECTRICITY COMMERCIAL OPERATION (9/771

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UNIT SIIUTDOWNS AND POWut REDUCTIONS DOCKET NO. 50-346 . ', , *

• UNIT NAME Davis-Besse Unit 1-

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DATE December 15, 1982 COMPLETED BY Erdal Caba REPORT MONTil November. 1982 TELErilONE (419) 259-5000. Ext. 196 l .

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- ,$ g 3* y .!! 5 Licensee Ev, g*, Cause & Corrective .

t No. Date . E 3 ij 2s& Event - g,7 Action to E fE $ jigg Report a mu g}

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Prevent Recurrence- -

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8. 82 11 08 F 18.9 A 3 NA HJ INSTRU Reactor tripped due to the Anticipa-tory Reactor Trip System (ARTS) caused by an erroneous moisture separaf.or reheater high water level

. signal.

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1 I 2 3 4 F: Forced Reason: Method: Exhibit G . !nstructions l S: Schedu!cd A Equipment Failure (Explain) 1-Manual for Preparation of Data 1 B Mainienance of Test 2 Manual Scrani. Entry Sheets for Licensee .

C Refueling 3 Automatic Scram. Event Report (LERI File (NUREG-D-Regulatory Restriction ' 4-Continuation from Previous Month 0161)

E Operator Training & License Exan-Jussion l 5-Load Reduction F Administrative i 9-Other (Explain) 5

• G-Operatiinial Error (Explain) Exfiibit I . Sanw Source (9/77) ll Other (Explain)

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OPERATIONAL

SUMMARY

l November, 1982 l 11/1/82 - 11/7/82 Reactor power was -increased to approximately 98% i of full power and maintained at this level.

11/8/82 At 0938 hours0.0109 days <br />0.261 hours <br />0.00155 weeks <br />3.56909e-4 months <br /> on November 8, 1982, an erroneous moisture separator reheater (MSR) high water j level signal caused a turbine trip. The Antici- '

patory Reactor Trip System (ARTS) then tripped i'

the reactor from a power level of approximately 100% full power. The MSR #1 high level turbine trip switch was replaced, and the reactor was-critical again at 1603 hours0.0186 days <br />0.445 hours <br />0.00265 weeks <br />6.099415e-4 months <br /> the same day.

l tl 11/9/82 - 11/30/82 The turbine-generator was synchronized on line at 0435 hours0.00503 days <br />0.121 hours <br />7.19246e-4 weeks <br />1.655175e-4 months <br /> on November 9, 1982. Reactor power was increased to approximately full power and maintained for the rest of the month. 1 i

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) .j COMPLETED FACILITY CHANGE REQUEST

. FCR NO: 77-006 SYSTEM: Miscellaneous

- COMPONENT: Control Room Vertical Panels CHANGE, TEXT OR EXPERIMENT: FCR 77-006 proposed the fabrication and installation of covers for the aast and south ends of-the vertical panels in the Control Room.

REASON FOR CHANGE: On July 19, 1977, these panels were installed to close the east and south ends of these panels.

SAFETY EVALUATION: The added vertical panel covers will in no way affect

- the safety function of the Control Room panels, nor were the seismic qualifications affected.

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• 7 s, COMPLETED FACILITY CHANGE REQUEST FCR NO: 78-315

- SYSTEM: Sufety Features Actuation System (SFAS)

- COMPONENT: SFAS wiring for CC 1407B d

. CHANGE, TEXT OR EXPERIMENT: On June 26, 1978, the internal wire connected to pin "K" of J208 and J408 connectors was rerouted through the spare wire linking pin "c" of the same connectors.

REASON FOR CHANGE: Wires connected to pin "K" of J208 and J408 connectors showed discontinuity along Channels 2 and 4 of SFAS. This discontinuity prevented actuation of L422B and L424B logic modules, which in turn

prevented the Component Cooling Return Header Containment Ouisr Isola-tion,_ Motor Actuated Butterfly Valve, CC1407B from closing. The problem was caused by the defective wire attached to pin "K".

SAFETY EVALUATION: Facility Chante Request 78-315 provided for the change of an internal cabinet connection wire in Channels 2 and 4 of SFAS. The wire would be replaced by a good wire which would enable actuation of CC1407B as required by SFAS.

Replacing the wires does not constitute -an unreviewed safety question. -

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i COMPLETED FACILITY CHANGE REQUEST FCR No: =79-213 SYSTEM: Emergency Diesel Generators COMPONENT: Cabinets C-3617 and C-3618

. _ CHANGE, TEXT OR EXPERIMENT: On May 1, 1980, new relays, relay sockots, _

and diodes were installed and tested to enable proper annunciator action for the SCR Diode Failure Exciter Regulator Alarm.

. REASON FOR CHANGE: When running the Diesel Generator, the SCR Diode Failure Alarm is on, however, there are no red lights (which indicate an alarm) on the SCR diodes. Several resistors and relays for this alarm have failed, and it is believed this happens because of a design problem in the circuit.

SAFETY EVALUATION: This FCR censiste'of installing new relays, relay-sockets, and diodes associated with the SCR Diode Failure Exciter Regula-tor Alarm.

This change will not affect the safety function of the Eaergency Diesel Generators. .It will improve-the reliability of the annunciator alarm circuit.

The audifications are' internal to Cabinets C-3617 and C-3618 and will not prevent a safe shutdown of the plant.

- All work involved with this package bac been done under the supervision of the vendor. Installation in accordance with the vendor's instructions is to insure no adverse environment is created. An unreviewed safety quest. ion

, does not exist.

Rev. 1 - 12/15/82 o

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. .J' COMPLETED FACILITY CHANGE REQUEST FCR NO: 80-01G SYSTEM: Auxiliary Feedwater System COMPONENT: Rooms 237 and 238

.i: CHANGE, TEXT Oh EXPERIMENT: FCR 80-018 proposed the installation of a 6" curb around the penetrations through the Auxiliary Feed Pump Room ceiling.

REASON FOR CHANGE: Installation of the curb prevented excessive water on the heater bay floor from draining to the Auxiliary Feed Pump Rooms direce,1y below and causing damage to equipment.

Work was completed June 6, 1982.

SAFETY EVALUATION: FCR 80-018 involved the installation of curbing around three (3) existing core drills on the floor of the heater bay area. The curbing would prevent possible water damage to equipment in Auxiliary Feed Pump Rooms located below the heater bay area.

No new adverse environments were created and an unreviewed safety question does not exist.

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l COMPLETED FACILITY CHANGE REQUEST FCR NO: 80-275 SYSTEM: -Emergency Diesel Generators COMPONENT: 'Turbochargers CHANGE, TEXT OR EXPERIMENT: New turbochargers with a high contact d-ive gear were installed on the Emergency Diesel Generators (EDG) per FCR 80-275. The work was completed June 30, 1982.

REASON FOR CHANGE: Farmer turbochargers were being replaced every 200 operating hours. The new high capacity turbocharger ,would increase the life span to approximately 3000 operating hours.

SAFETY EVALUATION: FCR 80-275 provided for changes to the turbocharger of each EDG. The new turbocharger has thicker gears to reduce contact stress level and increase contact ratio to better distribute torque. This turbocharger was intended to increase EDG reliability without affecting its safety function. No new adverse environment was created.

An unreviewed safety question does not exist- .

' COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-060 SYSTEM: . Emergency Diesel Generator-COMPONENT: Lube Oil Coolere CHANGE, TEXT OR' EXPERIMENT: FCR 81-060 proposed the installation of rolled type lube oil cooler cores to replace presently installed soldered Cores.

REASON FOR CHANGE: FCR 81-060 was implemented following reports of lube oil cooler leaks at other nuclear generating stations. Leaks were caused by a conosive attack of the soldered joints on the cooler core.

Work was completed and inspected on Emergency Diesel Generator.1-2 on-April 7, 1982 and on Emergency Diesel Generator 1-1 on April 17, 1982.

SAFETY EVALUATION: FCR 81-060 involved the Emergency Diesel Generators which are required for safe plant shutdown in tha event of loss of offsite power. This aforesaid change dcas not affect the performance of the coolers, nor does it affect the capability of the Emergency Diesel Genera-tors to perform their safety function. This change improved the reliabi-lity of the lube oil coolers and thus ' improved the reliability of the Emergency. Diesel Generators.

No new adverse' environments were created.

An unreviewed safety question does not exist.

'Rev. 1 - 12/15/82 1

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 82-025 SYSTEM: 125/250 Volt D.C.

, COMPONENT: Battery Charger DBC-2N

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CHANGE, TEXT OR EXPERIMENT: FCR 82-025, proposed the temporary operation ,

of Battery Charger DBC-2N with only 42 of the total 43 filter capacitors in. service.

REASON FOR CHANGE: The aforesaid Battery Charger DBC-2N was needed in order to facilitate the outage testing of the batteries as well as act as a replacement charger in the event of failure on DC MCC2. Since the replacement filter capacitor was.on order and not available, the one filter capacitor that had failed was unable to be repaired.

Work was-completed March 5, 1982 and represented a corrective action for Non-Conformance Report 159-82.

SAFETY EVALUATION: FCR 82-025 provided for the temporary operation of Battery Charger DBC-2N without its full complement of filter capacitors.

Use continued until a replacement for the faulty capacitor was installed August 5, 1982.

The safety ft.netion of the filter capacitor banka are to reduce the output ripple ~ voltage of the battery charger. The loss of one filter capacitor

'from the total of 43 did not adversely affect the function of DBC-2N. The modification is internal to Battery Charger DBC-2N and did not prevent the safe shutdown of the' station. S, FCR 82-025 does not constitute an unreviewed safety question and, further, created no new adverse environments.

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