ML20071Q323

Monthly Operating Rept for Oct 1982

ML20071Q323
Person / Time
Site:	Davis Besse
Issue date:	11/09/1982
From:	Sarsour B TOLEDO EDISON CO.
To:
Shared Package
ML20071Q320	List: ML20071Q317 ML20071Q323
References
NUDOCS 8212290309
Download: ML20071Q323 (8)

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 UNIT -Davis-Besse Unit 1 DATE November 9, 1982 COMPLETED BY Bilal Sarsour TELEPHONE (419) 259-5000, t.x t. ;84 MONTH October, 1982 i

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 3 766 37 765 2

765 gg 765 1

3 766 l9 765 4

765 20 765 3

766 21 765 6

765 22 764 6

7 763 23 766 8

757 24 762 9

751 25 764 10 752 26 763 13 751 27 763 12 760 33 17 13 765 148 29 14 763 30 779 15 763 804 3,

16 745 INSTRUCTIONS On this format. list the average daily urut power leselin MWe. Net for each day in the reportmg month. Compute to the nearest whole megawatt.

(4/77 )

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8212290309 021109 i

PDR ADOCK 05000346 i

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OPERATING DATA REPORT

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DOCKET NO.

50-346 DATE Novemoer 9, 1982 COhlPLETED BY Rilal hrSour TELEPHONE (4191 259-5000, Ext. 384 OPERATING STATUS

1. Unit Name:

Davis-Besse Unit 1 Notes

2. Reporting Period:

October. 1982

3. Licensed Thermal Power (MWr):

2772 925

4. Nameplate Rating (Gross MWe):

5. Design Electrical Rating (Net MWe):

906

6. Maximum Dependable Capacity (Gross MWe):

918

7. Maximum Dependable Capacity (Net MWe):

R76

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

9. Power Lesel To Which Restricted,if Any (Net MWe):

10. Reasons For Restrictions.If Any:

This Month Yr.-to-Date Cumulative II. Hours in Reporting Period 745 7,296

-37,297

12. Number Of Hours Reactor Was Critical 722 3,207.9 19,437.9

13. Reactor Reserve Shutdown Hours 23 23 3,357.7

14. Hours Generator On-Line 712.3 3.078.9 18,329.1

15. Unit Reserve Shutdown Hours 0

1.1 1,732.5

16. Gross Thermal Energy Generated (MWH) 1.671.849 6.435.474 41.556.999

17. Gross Electrical Energy Generated IMhH) 567,832,_,

2,141,917 13,824,168

18. Net Electrical Energy Generated (MWH) 535,615 2,000,975 12,898,260

19. Unit Service Factor 95.6 42.2 49.1

20. Unit Availability Factor 95.6 42.2 53.8

21. Unit Capacity Factor (Using MDC Net) 82.3 31.4 39.6

22. Unit Capacity Factor (Using DER Net) 79.4 30.3 38.2

23. Unit Forced Outage Rate 4.39 1.05 21.8

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

25. If Shur Down At End Of Report Period. Estimated Date of Startup:

26. Units in Test Status (Prior to Commercial Operation):

Forecast Achiesed INITIA L CRITICALITY INITIAL ELECTRICITY COM\\f ERCI A L OPER ATION (4/77)

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UNIT S!!UTDOWNS AND POWMt REDUCllONS Ul A IE Day -Besse Unit 1 DATE November 9, 1982 COMPLETED BY Bilal Sarsour REPORT MON fil October, 1982 TELErilONE (419) 259-5000, Ext. 384 I

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Cause & Corrective

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Even 3,y 6.i Action to gO Prevent Recurrence-H fE E

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82 10 28 F

32.7 11 3

NA NA NA The reactor tripped by the Antici-patory Reactor Trip System (ARTS) during the performance of PT 5193.01 Main Turbine Steam' Valves Test.

(See Operational Summary for further details.)

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F: Forted Reason:

Method:

Exhibit C. Instructions S: Schedu!cd A Equipment Failure (Emplain) 1-Manual for Preparation of Data j

j B. Maintenance or Test 2 Manual Scram,.

Entry Sheets for Licensee 2

0 Refueling 3 Autunutic Scram.

Event Report (LERI File (NUREG-D Regulatory Restriction 4-Continuation from Previous Month 0161)

F Opeutor Training & License Examination l 5-Load Reduction F. Adminiuutive i 9-Other (Explain) 5 G Opesational 1 nor (lixplain)

Esfiibit I. Sanw Source 19/77) ll Othee (E mplaini l

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OPERATIONAL

SUMMARY

October, 1982 10/1 - 10/8 Reactor power was maintained at approxiantely 86% with the generator gross load at approximately 810 2 10 MWe.

(The station was limited to a_ power level of 86%

because of an inoperable main safety valve. At 0030 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> on October 8,1982, power was reduced to approxi-mately 80% for turbine valve testing.

10/9 - 10/28 Reactor power was slowly increased and attained 85% on October 16, 1982. Later on October 16, 1982, reactor power was reduced to approximately 83% to perform ST 5013.04, " Control Rod Drive Exercise". Reactor power was also reduced to approximately 80% on October 16, 1982 for turbine valve testing. After the completion of turbine valve testing, reactor power was increased steadily to approximately 85?..

The reactor power was maintained at this level until 0150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> on October 28, 1982 when the reactor tripped during the performance of PT 5193.01, " Main Turbine Steam Valves Test", when a momentary low hydraulic pressure was sensed at the turbine stop valves and the Anticipatory Reactor Trip System (ARTS) automatically tripped the reactor.

10/28 - 10/31 The reactor was critical at 2327 hours0.0269 days <br />0.646 hours <br />0.00385 weeks <br />8.854235e-4 months <br /> on October 28, 1982.

The turbine was synchronized on line at 1033 hours0.012 days <br />0.287 hours <br />0.00171 weeks <br />3.930565e-4 months <br /> on October 29, 1982.

Reactor power was slowly increased and attained 90%

full power on October 30, 1982. Reactor power was maintained at this level for the rest of the month.

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q REFUELING INFORMATION DATE:

October, 1982 1

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Name of facility:

Davis-Besse Unit 1 t

2.

Scheduled date for next refueling shutdown:

September 3. 1983 3.

Scheduled dato for rustart following refueling:

October 29. 1983 4.

Will refueling or resumption of operation thereaf ter require a technical specifi-cation change or other license amendment? If answer is yes, what in general, will these be? If answer is no, has the reload fuel design and core configura-tion been reviewed by your Plant Safety Review Committee to determine whether any unreviewed safety questions are associated with the core reload (Ref. 10 CFR Section 50.59)?

Expect the relaad report to require standard reload fuel desien Technical Socciff-cation (3/4.1 Reactivity Control Systems and 3/4.2 Power Distribution Limits) 5.

Scheduled date(s) for submitting proposed licensing action and supporting infor-mation.

Julv, 1983 6.

Important liccasing considerations associated with refueling, e.g., new or different fuci dusinn or supplier, unruviewed doulgn or performancu unulyuls muthods, significant changes in fuel design, new operating procedures.

Ncno identified to date 7.

The number of fuel assemblies (a) in the core and (b) in the spent fuel storage pool.

(a) 177 (b) 92 - Spent Fuel Assemblies 8.

The present licensed spent fuel pool storage capacity and the size of any in-crease in licensed storage capacity that has been requested or is planned, in number of fuel assemblies.

Present 735 Increase size by 0 (zern) 9.

The pru)cetud datu uf thu last refueling that can be discharged to the npent fuel pool assuming the present licensed capacity.

Datu 1992

- assuming ability to unload the entire core into the spent fuel pool is maintained.

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T COMPLETED FACILITY CHANGE REQUEST FCR NO.79-139 SYSTEM: Main Steam COMPONENT: SP-17A1, SP-17A2, SP-17A8, SP-17A9, SP-17B1, SP-17B2, SP-17B8, and SP-17B9 CHANGE. TEST, OR EXPERIMENT: Facility Change Request (FCR)79-139 encom-passed the relocation of 1050 psig safety valves SP-17A1, SP-17A2, SP-17B1 and SP17-B2 in positions which were at the time held by 1100 psig safety valves SP-17A8, SP-17A9, SP-17B8 and SP-17B9, respectively. The 1100 psig safety valves were then relocated to the former position of the 1050 psig safety valves.

REASON FOR CHANGE: FCR 79-139 was implemented to improve the reliability of the 1050 psig safety valves. The station has had difficulty in maintain-ing these valves within the 1% design setpoint tolerance. See Licensee Event Report NP-33-79-51.

SAFETY EVALUATION: The requested relocation of the four code safety valves on each of the main steam headers caused no adverse effect to the relieving capacity of the valves or of the entite header. No change was made to the setpoints of the valves.

Implementation was completed on September 24, 1982. No unreviewed safety question exists.

4' COMPLETED FACILITY CHANGE REQUEST FCR NO: 80-244 SYSTEM: Control Room Heating, Ventilation and Air Conditioning (HV&AC)

COMPONENT: HV5301A, HV5301B, HV5301E, HV5361A, HV5362A, HV5362B, HV5311A, HV5311B, HV5311E CHANGE, TEST, OR EXPERIMENT: FCR 80-244 provided for the addition of solenoid valves to the instrument air lines in the Control Room normal ventilation system.

These valves were added to the instrument air lines in order to allow for a faster bleed-off time which in turn shortens the response time of the isolation dampers.

REASON FOR CHANGE: Response times for the Control Room ventilation system isola-tion valves was not within time limits as required by Technical Specifications.

See Licensee Event Report NP-33-80-96 for further details.

SAFETY EVALUATION: FCR 80-244 was completed by an installation similar to that of the existing system which includes Class lE electrical equipment and seismic Class 1 qualification. The testing accomplished under ECR 80-244 has demonstrated that the response time required for these valves will be met.

This change does not constitute an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR No.81-138 SYSTEM: Decay Heat COMPONENT: DR-13A, DH-13B, DH-14A, and DH-14B CHANGE, TEST, OR EXPERIMENT: Facility Change Request-(FCR)81-138 was implemented in order to revise Note 3 on Drawing 7749-E52B Sheet 25A to read:

FOR DETAILS OF CONTROL SW "CS" SEE DWG E-30B SH. 7 FIG. 7A FOR HIS DH13A AND HIS DH13B AND FIG. 7B FOR HIS DH14A AND HIS DH14B.

REASON FOR CHANGE: Original Drawing 7749-E32B was incorrect under Revi-sion 5 in such that it referenced "CS" (control switch) for only two valves. The aforesaid revision referenced all four valves with the "CS" notation.

SAFETY EVALUATION: The change authorized by FCR 81-138 was completed April 7, 1982 and did not create any new adverse environments. No un-reviewed safety question was involved.

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