ML20076H343

Monthly Operating Rept for Apr 1983

ML20076H343
Person / Time
Site:	Davis Besse
Issue date:	05/09/1983
From:	Murray T, Sarsour B TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA)
References
K83-697, NUDOCS 8306160540
Download: ML20076H343 (9)

Text

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AVERAGE DAILY UNIT POWER LEVEL e

DOCKET NO.

50-346 Davis-Besse Unit 1 UNIT DATE May 9, 1983 COMPLETED BY Bilal Sarsour TELEPHONE 419-259-5000, Ext.

384 MONTH April, 1983 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 877 874 -

37 2

877 18 875 3

878 g'9 875 4

879 20 874 5

876 21 874 6

870 22 874 7

875 23 874 8

840 24 872 9

37 25 872 10 86 867 26 595 855 27 808 12 28 819 13 868 29 791 14 873 730 30 15 875 3g 16 876 l

INSTRUCTIONS l

On this format. list the average daily unit power levelin MWe Net for each day in the reporting month. Compute to the nearest whole megawatt.

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8306160540 830509 PDR ADOCK 05000346 R

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OPERATING DATA REPORT DOCKET NO.

50-346 DATE May 9, 1983 COMPLETED BY Bilal Sarsour TELEPHONE 419-259-5000, OPERATING STATUS

1. Unit Name:

Davis-Besse Unit 1 Notes

, 2. Reporting Period:

April, 1983

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 (Irems Number 3 Through 7) Since Last Report. Give Reasons:

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

10. Reasons For Restrictions.If Any:

This Month Yr.-to.Date Cumulative

11. Hours In Reporting Period 719.0 2,879.0 41,640.0

12. Number Of Hours Reactor Was Critical 710.3 2,543.6 23,439.I

13. Reactor Reserve Shutdown Hours 0.0 313.9 3,678.0

14. Hours Generator On.Line 687.8 2,496.4 22,256.0

15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5

16. Gross Thermal Energy Generated (MWH) 1.815.549 6,710,291 52,083,052

17. Gross Electrical Energy Generated (MWH) 604.660,_,

2.248.371-17.354.025

18. Net Electrical Energy Generated (MWH) 573.042 2.130.770 16.246.210

19. Unit Service Factor 95.7 86.7 53.4

20. Unit Availability Factor 95.7 86.7 57.6

21. Unit Capacity Factor (Using MDC Net) 91.2 84.7 44.6

22. Unit Capacity Factor (Using DER Net) 88.0 81.7 43.1

23. Unit Forced Outage Rate 4.3 13.3 19.8

24. Shutdowns Scheduled Over Next 6 Months (Type. Date.and Duration of Each1:

July 29. 1983 Refueling Outage Duration: Approximately 8 weeks

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

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

Forecast Achiesed INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION (9/77i

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>l0 DOCKET NO. 50-346 e

UNIT SHUTDOWNS AND POM.!! REDUCTIONS UNIT NAME Davis-Ronco ifnit 1 DATE May 9, 1983 COMPLETED BY Bilal Sarsour

,' !l TELEPilONE 419-259-5000. Ext. 384 REPORT MONTil April. 1983 i j3 I

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$w, go, Cause & Corrective No.

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83 04 09 F

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Reactor power was reduced to appr'oxi-mately 7% due t<t a failure of the OTSG startup Acvel indication used for level control df the Auxiliary I

Feedwater System.

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15.3 11 3

NA The reactor tripped on Reactor Pro-tection System 9/ A 9/ flow when Xenon burnout produced a condition of nega-tive imbalance that exceeded the Reactor Protection System trip setting.

See Operational Summary for further l

detdils.

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F: Forced Reason:

Method:

Exhibit G. Instructions

'.i S: Schedu!cd A Equipment Failure (Explain) 1-Manual for Preparation of Data

'l B-Maintenance of Test 2 Manual Scraid.

Entry Sheets for Licensee C-Refueling 3 Automatic Scram.

Event Repost (LERI File (NUREG-D. Regulatory Restsiction 4-Continuation from Previous Month 0161)

E Operatur Training & License Exsudnatitm 5 Load Reduction F-Administrative 9-Other (Explain) 5 G-Operational Ensur (Explain)

Exfiibit I. Same Source (9/77) 110:her (Explain) e

OPERATIONAL

SUMMARY

April, 1983 4/1/83 - 4/9/83:

Reactor power was maintained at approximately 99 percent full power until 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> on April 9, 1983 when power was slowly reduced to approximately 7 percent power. The power reduction was due to a failure of the once through steam generator (OTSG) startup level indication used for level control of the Auxiliary Feedwater System. The turbine-generator was taken off line at 0322 hours0.00373 days <br />0.0894 hours <br />5.324074e-4 weeks <br />1.22521e-4 months <br />.

Reactor power was maintained at approximately 7 percent until 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> on April 9, 1983. When the indicator was repaired, a reactor power increase was initiated. The turbine-generator was synchronized on line at 1915 hours0.0222 days <br />0.532 hours <br />0.00317 weeks <br />7.286575e-4 months <br />.

4/10/83:

On April 10, 1983, the reactor tripped from 98 percent power on Reactor Protection System (RPS) flux-delta flux-flow (9/AS/ flow) when xenon burnout produced a condition of negative imbalance that exceeded the RPS trip setting.

The reactor was critical at 1255 hours0.0145 days <br />0.349 hours <br />0.00208 weeks <br />4.775275e-4 months <br />. The turbine generator was synchronized on line at 1904 hours0.022 days <br />0.529 hours <br />0.00315 weeks <br />7.24472e-4 months <br />.

4/11/83 - 4/29/83: Reactor power was slowly increased and attained 99 percent full power on April 14, 1983. Reactor power was maintained at 99 percent power until 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> on April 28, 1983, when power was reduced to approximately 85 percent due to steam leaks on both moisture separator high level drain lines to the condenser.

4/30/83:

Reactor power was also reduced to approximately 55 percent power to remove the axial power shaping rods to extend core life. Reactor power was slowly increased to approximat,ely 90 percent power and maintained at this level for the rest of the month.

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REFUELING INFORMATION DATE: April, 1983 1.

Name of facility: Davis-Besse Unit 1 2.

Scheduled date for next refueling shutdown: July 29, 1983 3.

Scheduled date for restart following refueling: September 23, 1983 4.

Will refueling or resumption of operation thereafter require a technical specification 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 configuration 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)?

Ans: Expect the Reload Report to require standard reload fuel design Technical Specification changes (3/4.1 Reactivity Control Systems and 3/4.2 Power Distribution Limits).

5.

Scheduled date(s) for submitting proposed licensing action and supporting information: June, 1983 6.

Important licensing considerations associated with refueling, e.g.,

new or different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures.

Ans: None 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 increase in licensed storage capacity that has been requested or is planned, in number of fuel assemblies.

Present: 735 Increase size by: 0 (zero)

I 9.

The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity.

Date:

1993 - assuming ability to unload the entire core into the spent fuel pool is maintained.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-159 SYSTEM: Containment Isolation r,

COMPONENT: Vent and drain lines CHANCE. TEST OR EXPERIMENT: On June 27, 1979, work implemented by FCR 79-159 was completed. This involved capping of lines pertaining to containment penetrations. Vent lines, drain lines, test and flush connec-tions were capped with special precautions taken to insure the lines would not be overpressurized during heatup.

REASON FOR CHANGE: 'The capping of these lines will aid in the verifica-tion of containment isolation of lines associated with penetrations.

SAFETY EVALUATION: The associated caps were reviewed to insure that they would not adversely affect system operation given precautions to preclude overpressurization of lines between closed valves and valve caps.

This did not affect any event described in the Safety Analysis Report, nor did it affect the station Technical Specifications. An unreviewed safety question was not involved.

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1 COMPLETED FACILITY CHANGE REQUEST S

FCR NO:

79-166 i

SYSTEM:

Steam and Faedwater Eupture Control System (SFRCS)

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COMPONENT: Ca*ofnet C5721

':,, l CHANGE, TEST UR '#XPERIMENT: On May 17, 1980, work implemented by-FP O 79-166 was c3mpleted. This involved the installation of studs on ter-ninals 21T311-13,16, 211B13-13,10, 21TB25-13,16, and 21T327-13,15 on the field wiring side of cabinet C5721.

2.1'ASON FOR CHANCE: Lurin the performance of the S?RCS Moathly Test,

- tervisals 13 and 16 in c4b$. net 3721 an used to simclats runnica reactor coolant pumps by p10:ics a jumper between these two terminals. The studs fkat were $nstelled provide a place to securely attach the jur.pers. Prior i

to this modification, the jumpers tended to fall off and could havescausec personnel injury or equipment damage.

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SAFETY EVALUATION: This change did not adversely affect the fu@.th n of this system.

It will improve reliability vhen performing the S7ECS Monthly Test. All modifications were internal to the cabiaet and will not prevent the safe shutdown of the plant. This change does not creat'e any new adverse environments and does not constitute an unreviewed safety

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COMPLETED FACILITY CHANGE REQUEST A

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NErlM: 480 volt ar.d 2?O role moter control centers a

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COY 20NENTA 460 volt and 240 volt feeders CH:.NGF., TEST OR EXDEE1M.5T: N R ??-210 was inplemented to revise thermal overInd heater selections to verify that the heater selections for the 460 vcit and 240 velt f eeder breakers in the field are in agreement with the relay setting sacers. All descrepanciaa were corrected, and all work 1

wu coupleted Secarber 3,1981.

RFAMN 16R dift:6E: This.revirica was required to coordinate existing and revis:.d cc:diciens in the plant.

_ SAFE'C EVAINATIb: h safety f.urction of Class IE electric equipment and syctems 19 to provide estrgency reacter shutdovu, centainment isolation,

' k riedtor core r.colhg, el containmeat asd reactor heat removal or they are I

d caaentist in preventing significant rele.sse of radioactive material to the enwi otmen..

p The o'terload' heaters for the Class 1E motor control centers perform no W'

acti6n. Functionally, they have no affect on nuclear safety, and they do 3l not'pa.rfers any control or tripping function.

Circuit breaders in the motor control centers provide fault protection.

Since overload heaters do not provide any safety fuction, there was no adverse environde'nt created by this change. Therefore, no unreviewed safety que4 tion was involved.

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w TOLEDO

%ss EDISON May 9, 1983 Log No. K83-697 File: RR 2 (P-6-83-04)

Docket No. 50-346 Lic_anse Nc. NPE-3 Mt. Norman Haller, Director Office of banagerant and Progran, Analysis U. S. Nuclear Regulatory Commission Washinstou, D.C.

20555

Dear Mr. Ihller:

Menchly Operating Report, April,1983 Eavis-Besse huclear Power Station Unit 1 Enclosed are ten cepics of the Manthly Operating Report for Davis-Besse Nuclear Power Station Unit 1 for the month of April, 1983.

Yours truly,

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Terry D. Murray Station Superintendent Davis-Besse Nuclear Power Station TDM/BMS/ljk i

l Enclosures l

cc:

Mr. James G. Keppler Regional Administrator, Region III Encl:

1 copy l

Mr. Richard DeYoung, Director Office of Inspection and Enforcement Encl: 2 copies l

Mr. Tom Peebles NRC Resident Inspector

((l' Encl:

1 copy THE TOLEDO EOiSON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEOO. OHIO 43652 1