ML20235J953

Monthly Operating Rept for Jan 1989 for Davis-Besse Unit 1

ML20235J953
Person / Time
Site:	Davis Besse
Issue date:	01/31/1989
From:	Sarsour B, Storz L TOLEDO EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
KB89-00057, KB89-57, NUDOCS 8902240451
Download: ML20235J953 (6)

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.y AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 UNIT avis-Besse Unit #1 D

DATE 2-13-89 COMPLETED BY Bilal Sarsour TELEPHONE (419) 249-5000 ext. 7384

. MONTH. January 1989 DAY AVERAGE DAILY POWER LEVEL' DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) g 866 17 885' 2

859 gg 489 3

872 39 c.

4 885 20 0

5 884 21 0

6 884 22 0

7 884 23 0

8 885 24 433 3

9 888 25 875 10 885 26 879 gi 890 27 883 12 890 28 881 13 885 882 29 14 885 30 882 885 15 3:

881 16 886 INSTRUCTIONS On this format, list the average daily unit power level in MWe-Nei for each day in the reporting month. Compute to the nearest whole megawait.

(9/77I die i

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e OPERATING DATA REPORT DOCKET NO.

50-346 DATE 2-13-89 COMPLETED BY Bilal Sarsour TELEPflONE (419 M49-sn00 ext. 7384 OPERATING STATUS

1. Unit Name:

Davis-Besse Unit #1 Notes

2. Reporting Period:

January 1989

3. Licensed Thermal Power (MWt):

2772

4. Nameplate Rating (Gross MWe):

925 906

5. Design Electrical Rating (Net MWe):

6. Maximum Dependable Capacity (Gross MWe):

904

7. Maximum Dependable Capacity (Net MWe):

860

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 Any:

This Month Yr.-to-Date Cumulative

11. Hours in Reporting Period 744.0 744.0 92.113

12. Number Of Hours Reactor Was Critical 620.1 620.1 46.227.6

13. Reactor Reserve Shutdown Hours 0.0 0.0 5.304.7

14. Hours Generator On Line 611.1 611.1 44.304.9

15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5

16. Gross Thermal Energy Generated (MWH) 1.668.522 1,668,522 103.537.911-

17. Gross Electrical Energy Generated (MWH) 557,996 557,996 34,199,180

18. Net Electrical Energy Generated (MWH) 526,019 526,019 31,991,071

19. Unit Service Factor 82.1 82.1 48.1

20. Unit Availability Factor 82.I 82.1 50.0

21. Unit Capacity Factor (Using MDC Net) 82.2 82.2 40.4

22. Unit Capacity Factor (Using DER Net) 78.0 78.0 la.1

23. Unit Forced Outage Rate 17.9 17.9 32.2

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

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

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

Forecast Achieved INITIA L CRITICA LITY INITIAL ELECTRICITY l

COMMERCIAL OPERATION (9/77)

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OPERATIONAL

SUMMARY

JANUARY 1989 Reactor power was maintained at approximately 100% full power until 1700. hours on January 2, 1989 when power was reduced to approximately 96% to perform turbine control valve testing.

After the completion of the control valve testing, reactor power was slowly increased and attained approximately 100% on January 2, 1989.

Reactor power was maintained at this power level until 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> on January 7, 1989 when a manual power reduction to approximately 94% vas initiated to perform Control Rod drive exercise testing.

Reactor power was slowly increased following Control Rod Drive exercise testing to 100% full power which was achieved at 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> on January 7, 1989.

Reactor power was maintained at approximately 100% until 1345 hours0.0156 days <br />0.374 hours <br />0.00222 weeks <br />5.117725e-4 months <br /> on j

January 18, 1989, when a reactor trip occurred.

The reactor tripped on high Reactor Coolant System (RCS) pressure. The cause of the high RCS pressure was the spurious initiation of the Integrated Control System (ICS) Rapid Feedvater Reduction (RFR) circuitry. The RFR is initiated by the CRD Trip confirm signal. At the time of the event, CRD breaker testing was in progress with CRD breaker "B" tripped open. The confirmation of a reactor trip signal, CRD trip confirm, requires that ChD breakers "B" or "D" and breakers "A" or'"C" be open.

Due to a fault in one or several of the logic boards for'the CRD trip confirm signal circuitry, a spurious CRD trip confirm signal was generated with only breaker "B" tripped. RFR is initiated when the reactor trips to decrease feedvater to a level that vill match post-trip decay heat. The RFR closes the main feedvater block and control valves and closes startup feedvater control valves to approximately 20% open.

Since the effect of this action was to greatly decrease feedvater flow, the reactor, which was still at 100% power, tripped on high RCS pressure, approximately 16 seconds after RFR was initiated.

At 1650 hours0.0191 days <br />0.458 hours <br />0.00273 weeks <br />6.27825e-4 months <br /> on January 19, 1989, the pressurizer spray valve, RC2, was found to be cycling intermittently, causing the motor vindings to burn open.

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motor was replaced and an anti-rotation device was fabricated and installed.

After all repairs and modifications were completed, the actuator was set up and tested with MOVATS and found satisfactory.

The reactor was critical at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> on January 23, 1989, and the turbine l

vas synchronized on line at 0238 hours0.00275 days <br />0.0661 hours <br />3.935185e-4 weeks <br />9.0559e-5 months <br /> on January 24, 1989.

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The reactor was slowly increased to approximately 100% full power which was attained at 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> on January 25, 1989, and maintained at this power level for the rest of the month.

REFUELING INFORMATION Date:

January 1989 1.

Name of facility: Davis-Besse Unit 1 2.

Scheduled date for next refueling outage? February 1990 3.

Scheduled date for restart from current refueling: N/A 4.

Will refueling or resumption of operation thereafter require a technical specification change or other license amendment? If answer is yes, what in general vill 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: The Reload Report requires standard reload fuel design Technical Specifications changes (2. Safety Limits and Limiting Safety System Settings, 3/4.1 Reactivity Control Systems, 3/4.2 Power Distribution Limits and 3/4.4 Reactor Coolant System.)

5.

Scheduled date(s) for submitting proposed licensing action and supporting information: Received Amendment from NRC October 3, 1988 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.

a.

The highly absorbing silver-indium-cadmium axial power shaping rods were replaced with reduced absorbing inconel rods.

b.

The discrete neutron sources were removed from the core and not replaced.

c.

The physics testing has been reduced in scope.

7.

The number of fuel assemblies (a) in the core and (b) in the spent l

fuel storage pool, and (c) the new fuel storage areas.

(a) 177 (b) 268 (c) 0 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 Increased size by:

0 (zero) 9.

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

l Date:

1995 - assuming ability to unload the entire core into the spent fuel pool is maintained

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TOLEDO

% EDISON o

EDISON PLAZA.

300 MADISON AVENUE TOLEDO, OHIO 43652-0001 February 13, 1989-KB89-00057 Docket No. 50-346' License No. NPF-3 Document Control Desk' U. S. Nuclear Regulatory Commission

'7920 Norfolk Avenue.

. Bethesda, MD 20555 Gentlemen:

Monthly Operating Report, January 1989 Davis-Besse Nuclear Power Station Unit 1 Enclosed are ten copies of the Monthly Operating' Report for Davis-Besse

' Nuclear Power Station Unit No. 1 for the month of January, 1989.

If you have any questions, please contact Bilal Sarsour at (419) 249-5000, extension 7384.

Very truly yours, d

Louis F. Storz Plant Manager Davis-Besse Nucl:>ar Power Station BMS/mjb Enclosures cc:

Mr. A. Bert Davis Regional Administrator, Region III Mr. Paul Byron NRC Resident Inspector

]

Mr. A. V. DeAgazio NRC Project Manager

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