ML19274D179

Monthly Operating Rept for Dec 1978

ML19274D179
Person / Time
Site:	Davis Besse
Issue date:	01/08/1979
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML19274D178	List: ML19274D177 ML19274D179
References
NUDOCS 7901150129
Download: ML19274D179 (9)

Text

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 UNIT Davis-Besse Unit 1 DATE

.Tanuary 8, 1979 Erdal Caba COMPLETED SY 419-259-5000, Ext.

TELEP110NE 236 MONTil December, 1970 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LL*'EL (MWe-Net) (MWe-Net) j 832 0 17 2 569 18 0 3

677 0 19 4

745 0 20 692 0 5 21 6 716 0 22 _

7 100 23 0 0 0 8 24 9 265 25 0

10 383 0 26 716 0.

33 27 70 12 28 0 13 371 0 29 ,

14 373 0 30  :

373 0  ;

35 31 i

16 10 i INSTRUCTIONS On this farmat.'ist the average daily unit power leselin MWe-Net for each day in the reporting month. Compute to the nearest whole rnegawatt.

(9/77) 790115 0 jacj .

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OPERATING DATA REPORT DOCKET NO. 50-346 DATE January s, 1979 COh!PLETED BY Erdal Caba TELEPHONE 419- 259-5000, Ext.

236 OPERATING STATUS Davis-Besse Unit 1 Notes

1. Unit Name:

December, 1978

2. Reporting Period:

2772

3. Licensed Thermal Power (51Wt):

925

4. Nameplate Rating (Gross 51We):

5. Design Electrical Rating (Net htWe):

906

6. Sfaximum Dependable Capacity (Gross 5tWe): to be determined

7. Sf aximum Dependable Capacity (Net 51We):

to be determined

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 h!We): None

10. Reasons For Restrictions. If Any:

This hionth Yr.-to-Date Cumulative

11. Hours in Reporting Period 744 8760 11765

12. Number Of Hours Reactor Was Critical -328.8 4839.7 7 331.6

13. Reactor Resene Shutdown Hours 0 38.9 422.6

14. Hours Generator On-Line 314.1 4266.Z 5/33.2 0 0 0

15. Unit Resene Shutdown Hours

16. Gross thermal Energy Generated (A1WH) 626,864 8,523,538 10,187,570 .

17. Gross flectrical Energy Generated (51WH) 196,972 , , , ,

2,859,306 3,383,755

18. Net Elect.;uil Energy Generated (htWi!) 173,312 2,611,642 3,041,460 42.2% 48.7% 51.4%

19. Unit Service Factor

20. Unit Asailability Factor 42.2% 48.7% 15.4%

21. Unit Capacity Factor (Using b!DC Net) .to .a determined to be determined

22. Unit Capacity Factor (Using DER Net) 25.7% 32.9% 33.3%

23. Unit Fuced Outage Rate 57.8% 25.4% 26.7%

24. Shuraowns Scheduled Over Next 6 51onths (Type. Date.and Duration of Each):

None

25. If Shut Down At End Of Report Period Estimate <* Date of Startup:

26. Units in Test Status (Prior to Commercial Operat m): Forecast Achiesed INITIA L CRITICALITY INITIAL ELECTRICITY N/A CON 151ERCIA L OPER ATION (9/77) i i

i

- OPERATIONAL SUFDIARY FOR DECEMBER, 1978 12/1/78 Reactor p er was maintained at 100 percent with the turbine generator gross load at 876 + 10 MWe. The High Pressure Feedwater Heater Train 1 was still isolated from service.

12/2/78 A reactor power reduction to 50 percent was initiated at 0540 hours0.00625 days <br />0.15 hours <br />8.928571e-4 weeks <br />2.0547e-4 months <br /> to enable the scheduled withdrawal of Group 7 control rods at 145 EFFD. Group 7 control rods were pulled (maintaining at 85 to 95% withdrawn) and a reactor power in-crease was initiated ut 1420 hours0.0164 days <br />0.394 hours <br />0.00235 weeks <br />5.4031e-4 months <br />. The reactor power increase was delayo. from 1955 hours0.0226 days <br />0.543 hours <br />0.00323 weeks <br />7.438775e-4 months <br /> to 2350 hours0.0272 days <br />0.653 hours <br />0.00389 weeks <br />8.94175e-4 months <br /> to replace the sole-noid of NRV 491 which is the discharge valve on the Main Feed-water Pump Turbine 1-2.

12/3/78 Reactor power attained 91 percent at 1530 hours0.0177 days <br />0.425 hours <br />0.00253 weeks <br />5.82165e-4 months <br /> and was then increased to 93 percent. This power level was maintained with the turbine generator gross load at 830 + 10 MWe, 12/4/78 Reactor power was decreased to 83 percent at 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br /> to prepare for the Moderator Temperature Coefficient Measurement Test. The turbine generator gross load at this power level was 735 + 10 MWe.

12/5/78 The measurement of the moderator temperature coefficient was delayed at 1330 hours0.0154 days <br />0.369 hours <br />0.0022 weeks <br />5.06065e-4 months <br /> because of oscillations of the Turbine Control Valve Number 3. The oscillation was caused by a pivot pin missing in the valve's control linkage. A load reductior.

was initiated at 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br /> to replace the pivot pin.

12/6/78 At 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />, the unit load had been decreased to 595 MWe.

The repair work on the Turbine Control Valve Number 3 was com-pleted and reactor power was increased and attained 92 percent at 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />.

12/7/78 At 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />, the Post-Accident Radiation Monitor RE 5030 failed. The Post-Accident Radiation Monitor RE 5029 was started but failed at 0230 hours0.00266 days <br />0.0639 hours <br />3.80291e-4 weeks <br />8.7515e-5 months <br />. A unit shutdown was initia-ted per Technical Specification requirements. The turbine generator was ,ff line at 0435 hours0.00503 days <br />0.121 hours <br />7.19246e-4 weeks <br />1.655175e-4 months <br /> and Mode 3 was attained at 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />. Tc permit reactor startup, the Post-Accident Radiation Monitor RE 5030 or RE 5029, the Safety Features Actuation System Radiation Monitor RE 2005, and the Group 3 Rod 4 Absolute Position Indication (API) were repaired.

12/8/78 The RE 5030, RE 2005 and Group 3 Rod 4 API were operable and reactor startup was initiated. For details on the repair work, refer to Licensee Event Reports NP-33-78-143 for RE 5030; e

OPERATIONAL SUMS!ARY FOR DECEMBER, 1978 PAGE 2 12/8/78 Cont'd NP-33-78-140 for RE 2005; and NP-33-78-135 and NP-33-78-144 for Group 3, Rod 4 API.

Reactor criticality was attained at 2243 hours0.026 days <br />0.623 hours <br />0.00371 weeks <br />8.534615e-4 months <br />.

12/9/78 The turbine generator was synchronized on line at 0530 hours0.00613 days <br />0.147 hours <br />8.763227e-4 weeks <br />2.01665e-4 months <br />.

Reactor power was increased to 50 percent and maintained at this power level the remainder of the day.

12/10/78 The turbine generator gross load was maintained at 420 t 10 MWe, with reactor power at 50 percent. The unit power level was limited to 50 percent because the Main Feedwater Pump Turbine (MFPT) 1-2 was isolated to complete repair work on a drain line steam leak at the pump's casing. The high pressure feedwater heater train 1 was returned to service at 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br />.

12/11/78 The repair work on MFPT 1-2 was completed and a reactor power increase was initiated. At 0918 hours0.0106 days <br />0.255 hours <br />0.00152 weeks <br />3.49299e-4 months <br />, reactor power was at 92 percent with the turbine generator gross load at 840 + 10 MWe.

12/12/78 The Moderator Temperature Coefficient Measurement Test was completed. At 2205 hours0.0255 days <br />0.613 hours <br />0.00365 weeks <br />8.390025e-4 months <br />, reactor power was reduced to 50 per-cent because of problems experienced with high pressure feed-water heater 1-4 extraction steam pressure.

12/13/78 - 12/15/78 Reactor power was maintained at 50 percent with the turbine generator gross load at 420 1 10 MWe, while preparations were made for a unit outage. Th~e outage was necessary to repair the extraction steam line expansion joint failures which caused the feedwater heater 1-4 extraction steam pressure problens.

12/16/78 - 12/31/78 The turbihe generator was off line at 0303 hours0.00351 days <br />0.0842 hours <br />5.009921e-4 weeks <br />1.152915e-4 months <br /> and the reactor was suberitical at 0510 hours0.0059 days <br />0.142 hours <br />8.43254e-4 weeks <br />1.94055e-4 months <br /> on December 16, 1978.

During the outage. there were four extraction steam line bellows replaced, the MFPT 1-2 was disassembled and internals inspected to determine the cuase of seal leakage, the Main Fuel Handling Bridge was inspected and repairs implemented per Stearns-Rogers, the vendor, and various valves in containment were repacked. Also, the pressurizer relief valve RC 11 which was torquing out was repaired by installing a new torque switch.

The outage duration was increased on December 24, 1978, when.

16 out of 40 newly installed ccndenser manway studs broke when the condenser was being returned to service.

OPERATIONAL _SU> DIARY FOR DECEMBER, 1978 PAGE 3 12/16/78 - 12/31/78 Reactor criticality was attained at 1353 hours0.0157 days <br />0.376 hours <br />0.00224 weeks <br />5.148165e-4 months <br /> on December Cont'd 31, 1978. The turbine generator was placed on line at 0340 hours0.00394 days <br />0.0944 hours <br />5.621693e-4 weeks <br />1.2937e-4 months <br /> on January 1,1979 and a power increase was initiated.

' t I

DOCKET NO.

50-346 UNIT S!!UTDOWNS AND POWL!t REDUCTIONS Davis-Besse Unit 1, UNIT N AME

. January 8, 19/9 JATE COMPLETED IlY Charles N. Alm December, 1978 TELEPilONE 419-b9-5000, Ext. 251 REPORT MONTil l

- E E -

E'1, Cause & Corrective

, .$ ? 3g $ .Y 3 jg5 Licensee ,E-t, 3,7 $3 Action to  !

No. Date g 3g Event Prevent Recurrence i

H

$5- 5 Ji g F Report # cn U u

EO

<- g O

BB INSTRU The Po!>t-Accident Radiation Monito,rs

• F 48.9 A 1 NP-33-78-143 30 78 12 07 RE5029 and RE5030 both failed. Unit

• was forced to shutdown per Technical Specification requirements. Refer to the attached summary and Licensee e Event Report NP-33-78-143 for further details. ,

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IIH PIPEXX Unit loail was decreased to 50 percent 78 12 12 F 0 A 4 N/A because of low pressure in the high 31

. pressure feedwater heater 1-4. The low pressure was caused by failure of the extraction steam line. bellows.

HH PIPEXX The unit was shutdown to repair the 381 A 1 N/A 32 78 12 16 F extraction steam line bellows. Refer

'to the attached summary for further details. .

3 4 I 2 Exhibit G. Instructions Reason: Method:

F: Forced 1-Manual for Preparation of Data S: Schedu!ed. A Equipment Failure (Explain) Entry Sheets for Licensee B Maintenance of Test 2-Manual Scram.

3-Automatic Scram. Event ReporI(LER) File (NUREG-C Refueling 0161)

D Itegulatory Restriction 4 Other (Explain)

E Operator Training & License Examination 5

• F Admmistrative Exhibit 1. Same Source G-Operational Eirur (Explain)

(9/77) Il Other (Explain) f

FACILITY CHANGE REQUESTS COMPLETED DURING DECEMBER, 1978 FCR NO: 77-363 SYSTEM: Containment Spray System COMPONENT: Containment Spray Pumps 1-1 and 1-2 CHANGE, TEST, OR EXPERIMENT: On May 31, 1978, mechanical seals were installed on Containment Spray Pumps 1-1 and 1-2 under the direction of a representative of the pump vendor, Goulds Pumps. On September 14, 1978, work was completed which pro-vided.an easily accessible source of demineralized water for flushing the seals following operation of the pump. The unit architect-engineer, Bechtel Corporation, has revised all affected drawings.

REASON FOR,THE FCR: The performance of the former seals, which employed packing material, was not satisfactory. The Icakage was within Technical Specification limits, but the leakage was not in accordance with good operating practice. It was decided to replace the packing with mechanical seals.

SAFETY EVALUATION: The use of mechanical sea?u or packing f alls completely within the design of the containment spray pumps. Either may be used providing all design requirements, including Quality Assurance documentation, meet the specification.

The change will not adversely affect the safety function of the containment spray system.

FACILITY CHANGE REQUESTS COMPLETED DURING DECEMBER, 1978 FCR NO: 78-013 SYSTEM: Final Safety Analysis Report (FSAR)

COMPONENT: Chapter 14 CHANGE, TEST, OR EXPERIMENT: FCR 78-013 was written on January 6,1978, to nullify the erroneous requirement in the FSAR abstract of TP 800.25, " Shutdown From Outside the Control Room", that the reactor be tripped from outside the Control Room as a startup test. TP 800.25 will be performed by tripping the reactor from the Control Room.

REASON FOR THE FCR: All of the accident analyses in the FSAR and the Davis-Besse Unit 1 Fire Hazards Analysis Report assume the reactor to be tripped prior to evacuation of the Control Room.

SAFETY EVALUATION: Administative Procedures require that the reactor be tripped prior to evacuation of the Control Room. The FSAR abstract of TP 800.25 requires that the reactor be tripped from outside the Control Room as a startup test. This commitment was an error and is not consistent with the remainder of the FSAR.

Reference the Chapter 7 discussion of the Auxiliary Shutdown Panel.

This change in the Startup Test Program from that described in Chapter 12 of the FSAR does not therefore represent an unresolved or unreviewed safety question.

,s

• FACILITY CHANGE REQUESTS CO'iPLETED DURING DECEMBER, 1978 FCR NO: 78-295 SYSTEM: Reactor Protection System (RPS)

I COMPONENT: Grounding i l

l CHANCE, TEST, OR EXPERDIENT: Oa June 25, 1978, physical work and the associated j testing was completed to provide an Additional instrument ground cable from each '

RPS channel to the instrument ground bus which is located in the cable spreading room. {

i REASON FOR THE FCR: The change provides for a redundant ground on each channel  !

which allows the testing of the channel ground while maintaining ground continuity '

to the channel under test. The testing is necessary on a monthly basis to ensure proper RPS operation due to the safety hazard a loss of ground to an RPS channel could constitute. This concern was previously reported in a letter to NRC Region j III dated March 15, 1978, Serial Number 420. l SAFETY EVALUATION: The additional instrument ground cable from each RPS channel  ;

to the instrument ground bus will not affect the safety function of the RPS because the ground system will be mainta'.ned intact during testing of the NI/RPS grrind. I I

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