ML20084M383

Monthly Operating Rept for Mar 1983

ML20084M383
Person / Time
Site:	Davis Besse
Issue date:	04/08/1983
From:	Murray T, Sarsour B TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
K83-572, NUDOCS 8306020129
Download: ML20084M383 (9)

Text

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 UNIT Davis-Besse Unit 1 DATE April 8, 1983 COMPLETED BY Bilal Sarsour TELEPHONE 419-259-5000, Ext.

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-MONTH March, 1983 ,

DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe. Net) (ywe. Net) 1 876 872 37 2 875 18 859 3 879 g9 879 4 866 20 861 5 872 21 877 6 700 22 877 7 877 23 878 8 869 24 878 9 879 878 25 10 877 26 878 e 878 877 11 27 f 12 878 877 28 13 878 878 29 14 877 877 30 15 878 3; 877 16 878 INSTRUCTIONS On this format list the average daily unit power leselin MWe-Net for each day in the reportanc month. Compute to the nearest whole megawatt.

(9/77) 8306020129 8:30408 PDR ADOCK 05000346 R PDR I I

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OPERATLNG DATA REPORT DOCKET NO. 50-346 DATE April 8, 1983 COSIPLETED BY Bilal Sarsour TELEPHONE 419-259-5000 OPERATING STATUS

1. Unit Name: Davis-Besse Unit 1 Notes

. 2. Reporting Period: March, 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 (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 2,160.0 40.921.0

12. Number Of Hours Reactor Was Critical 744.0 1,833.3 22,728.8

13. Reactor Reserve Shutdown Hours 0.0 313.9 3,678.0

14. Hours Generator On-Line 744.0 1,808.6 21,568.2

15. Unit Reserve Shutdown Hours 0.0 0.0 1.732.5

16. Gross Thermal Energy Generated (MWH) 2.024.447 4.894.742 50.267.503

17. Gross Electrical Energy Generated (MWH) 679,685 ,_, 1,643,711 Iti,749,365

18. Net Electrical Energy Generated (MWH) 647,067 '1,557,728 15,673,168

19. Unit Service Factor 100.0 -83.7 52.7-

20. Unit Availability Factor 100.0 83.7 56.9

21. Unit Capacity Factor (Using MDC Net) 99.5 82.5 43.8

' 22. Unit Capacity Factor (Using DER Net) 96.0 79.6 42.3 l 23. Unit Forced Outage Rate 0.0 16.3 20.3

24. Shutdowns Scheduled Oser Next 6 Months (Type.Date.and Duration of Eacht:

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 INITIA L CRITICA LITY INITIAL ELECTRICITY COMMERCIA L OPER ATION (4/77 )

50-346 '.

. UNIT SHUTDOWNS AND POW:.;t REDUCliONS DOCKET NO.

UNIT NAME Davis-Besse Unit 1.

DATE April 8, 1983 COMPLETED llY Bilal Sarsour REPORT MONTil March, 1983 419-259-5000, Ext. 384 ELErilONE i

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• EU Licensee E* E Cause & Corrective

, No. Date E jl .3 3i$ Event

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mu E.g1 Action to p

F f5 $ j p, Reporisr y0 Prevent Recurrence 5

! 3 83 03 05 F 0 A 5 NP-33-83-18 >RB CKTBRK Reactor power was reduced to

. approximately 55% by a plant runback when control rod drive group 7 con-trol rod 12 ratcbet tripped.

See Operational Summary for ,

i further details.

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, 1 2 3 4 i* F: Forced Reason: Method: Exhibit G. Instructions S: Schedu!cd A Equipment Failure (Explain) 1-Manual ,

for Preparation of Data i B Maintenance of Test 2-Manual Scram. Entry Sheets for Licensee C Refueling 3 Automatic Scram. .

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

E-Operator Training & License Examination l 5-Load Reduction l F Administrative 9-Other (Explain) $ '

G-Operational I!:ror (Explain) Extiibit I - Sanw Source c1/77) II.Other (Explaini

OPERATIONAL SQiMARY MARCH. 1983 3/1/83 - 3/6/83: Reactor power was maintained at approximately 99% of full power until Control Rod Group 7 Rod 12 ratchet tripped into the core. A plant runback was initiated which reduced plant power to approximately 55%. The cause of the ratchet trip was determined to be a blown fuse on the "B" phase of a control rod drive motor power supply. The fuse was replaced, and Rod 7-12 repulled at 0145 hours0.00168 days <br />0.0403 hours <br />2.397487e-4 weeks <br />5.51725e-5 months <br />.

Reactor power was restored to 86% by 0715 hours0.00828 days <br />0.199 hours <br />0.00118 weeks <br />2.720575e-4 months <br /> for valves testing. Reactor power slowly increased and attained approximately 99% by 2200 hours0.0255 days <br />0.611 hours <br />0.00364 weeks <br />8.371e-4 months <br /> on March 6, 1983.

3/7/83 - 3/31/83: Reactor power was maintained at approximately 99%

until 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> on March 17, 1983, when power was reduced to approximately 96% to perform Surveillance Test ST 5010.02, " Moderator Temperature Coefficient Physics Testing". After the completion of ST 5010.02, reactor power increased steadily to approximately 99%

and maintained at this level for the rest of the month.

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REFUELING INFORMATION DATE: March, 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)

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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f COMPLETED FACILITY CHANGE REQUEST FCR NO: 80-181 SYSTEM: Safety Features Actuation System (SFAS)

COMPONENT: Logic modules CHANCE, TEST OR EXPERIMENT: On July 24, 1980, work implemented by FCR 80-181 was completed. This provided a delay in tripping of output logics, five per SFAS logic channel, that are on sequence steps other than sequence step 1. ,

REASON FOR CHANGE: This has improved sequencer operation by proper sequencing of loads on the diesel generator. This change will also allow the output logic to be blocked before tripping.

SAFETY EVALUATION: This FCR called for providing a time delay in tripping of the following logic modules in the SFAS:

SEQUENCE CHANNEL 1 CHANNEL 2 CHANNEL 3 CHANNEL 4 STEP L211 L212 L213 L214 2 L221 L222 L223 L224 5 L241 L242 L243 L244 4 L311 L312 L313 L314 -

3 L411 L412 L413 L414 5 L511 L512 L513 L514 5 These logic modules are on sequence steps other than sequence step 1 and have to be blocked by the sequencer in case of a SFAS trip with a loss of offsite power.

During the analysis of results from ST 5031.07, Integrated SFAS Test, conducted in June 1980, it was discovered that several of the above logic modules tripped before their respective sequence step. Soon after their tripping, the tripped logic modules went to the untripped state. These logic modules tripped again, later, at their respective sequence steps, as designed. From this, it was concluded that the trip signals from these modules were not being timely blocked by the sequencer.

This FCR modified the output modules listed above so that they will be timely blocked, after a SFAS trip with loss of offsite power, until tripped by the sequencer. The time delay in tripping of these logic modules enables proper blocking from the sequencer.

The Final Safety Analysis Report (FSAR) Table 7-7 states that the SFAS response time should be less than or equal to five seconds for all measured varibles. After adding the required equipment to provide the time delay, the field measurements have indicated that the delay ranges anywhere from 45 to 82 milliseconds. This time delay is negligible in comparison with the response time requirement provided in the FSAR. Moreover, response

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FCR 80-181 Page 2 time testing has indicated considerable margin implying that the FSAR and Technical Specification response times will still be met with the added delay.

Pursuant to the above, the change implemented by this FCR does not involve an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST t,

FCR NO: 82-103 SYSTEM: Containment Air Sample System COMPONENT: Valves HV5010E and HV50llE

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CHANGE, TEST OR EXPERIMENT: This FCR was implemented to reduce the torque switch settings for Limitorque actuators for valves HV5010E and HV5011E.

The new settings, both for opening and closing, are 1.0. Work was completed on August 11, 1982.

REASON FOR CHANCE: The original torque switch settings of these two valves caused the valve stems to be overtorqued and bent. The reduced settings will help prevent this.

SAFETY EVALUATION: This FCR called for decreasing the torque switch settings for containment air sample return valves HV5010E and HV50llE.

The safety function of these valves is to isolare containment on a Safety Features Actuation Incident Level 1.

The previous torque dial setting of 2.0 had resulted in the bending of,th"e valve stem during valve operation.

The calculated values for the dial settings for both valves are 0.99 for /

opening and 0.785 for closing. It is convenient in the field to set the dials at 1.0.

4 The new settings of 1.0, both for closing and opening, have enhanced the -

equipment operation and are still sufficient to perform their function.

Therefore, this change has not affected the safety function of HV5010E and HV5011E. Hence, no unresiewed safety question is involved.

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TOLEDO e .

%s EDISON

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April 8, 198,6 Log No. K83-572 File: RR 2 (P-6-83-03)

Docket No. 50-346 License No. NPF-3 Mr. Norman llaller, Director i Office of Management and Program Analysis U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Haller:

i Monthly Operating Report, March, 1983 Davis-Besse Nuclear Power Station Unit 1 Enclosed are ten copies of the Monthly Operating Report for Davis-3 esse Nuclear Power Station Unit 1 for the month of March, 1983.

Yours truly, fqD,Wy M Terry D. Murray Station Superintendent Davis-Besse Nuclear Power Station TDM/BMS/ljk i

Enclosures cc: ' Mr. James G. Keppler i Regional Administrator, Region III Encl: 1 copy i<

Mr. Richard DeYoung. Director Office of Inspection and Enforcement Enc 1: 2 copies =

Mr. Tom Peebles NRC Resident Inspector

- Encl: 'l copy

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THE TCLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652

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