ML20077A010

Monthly Operating Rept for May 1983

ML20077A010
Person / Time
Site:	Davis Besse
Issue date:	06/09/1983
From:	Murray T, Sarsour B TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
K83-852, NUDOCS 8307220092
Download: ML20077A010 (12)

Text

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 Davis-Besse Unit 1 WIT DATE COMPLETED BY Bilal Sarsour TELEPHONE 419-259-5000 Ext-384 MONTH May, 1983 DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (MWe Net)

(MWe-Net) 1 793 795-37 2

792 18 793 3

793 19 791 4

789 20 794 5

794 794 21 6

793 22 795 7

791 23 792 8

789 24 795 9

796 25 792 10 337 26 793 11 643 27 793 12 771 28 791 13 792 643 29 34 792 642 30 35 797 31 757 j

16 795 INSTRUCTIONS i

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

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9307220092 830609 PDR ADOCK 05000346 R

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OPERATING DATA REPORT DOCKET NO.50-34A DATE June 9, 1983 COMPLETED BY Bilal Sarsour TELEPHONE _.419-259-5000, Ext. 384 OPERATING STATUS

1. Unit Name:

Davis-Besse Unit 1 Notes

2. Reporting Period:

Flav. 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 e

11. Hours In Reporting Period 744.0 3.623.0

~42.384.0

12. Number Of Hours Reactor Was Critical 739.2 3,282.8 24,178.3

13. Reactor Reserve Shutdown Hours 0.0 313.9 3,678.0

14. Hours Generator On.Line 734.0 3,230.4 22,990.0

15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5

16. Gross Thermal Energy Generated (MWH) 1,801,578 8.511,869 53,884,630

17. Gross Electrical Energy Generated (MWH) 599,378 2,847,749 17,953,403

14. Net Electrical Energy Generated (MWH) 566,76T 2,697,539 16,812,979

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19. Unit Service Factor 98.7 89.2 54.2

20. Unit Av'ailability Factor 98.7 89.2 58.3

21. Unit Capacity Factor (Using MDC Net) 87.2 85.2 45.4

22. Unit Capacity Factor (Using DER Net) 84.1 82.2 43.8

23. Unit Forced Outage Rate 1.3 10.8 19.4

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

i July 29, 1983 Refueling Outage Duration: Anoroximately 8 weeks

' 25. If Shut Down At End Of Report Period. Estimated Date oiStartup:

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

fcrecast Achiesed INITIAL CRITICALITY INITIA L ELECTRICITY COMMERCIA L OPER ATION l

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50-346 DOCKET NO.

, 1 UNIT S!!UTDOWNS AND POW:.it REDUCTION $

UNIT NAME Davis-Besse Unit I

DATE June 9. 1983 COMPLETED BY Bilal Sarsour REPORT MONTil May. 1983 TELErilONE 419-259-5000. Ext. 384 E

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Cause & Corrective d l No.

Date f.

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Event 3,7 98 Action to y'

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Prevent Recurrence i

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3 NP-33-83-32 EB CKTBRK The reactor tripped on high Reactor Coolant System (RCS) pressure. The trip was a result of a loss of Y4

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due to construction work.

See Operational Summary for further i

details.,

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4 F: Forced Reason:

Method:

Exhibit G. Instructions S: Schedu!cd A. Equipment Failure (Explain) 1 Hanual for Preparation of Data

'l B Maintenance of Test 2-Manual Scrani.

Entry Sheets for Licensee C R fueling-3-Automatic Scrum.

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

E-Operator Training & License Examination j 5-Load Reduction F-Administrative

9-Other (Explain) 5 G Operational Enor (lixplain)

Exliibit I - Same Source 07/77)

Il-Other (Explain) l J

OPERATIONAL

SUMMARY

MAY, 1983 5/1/83 - 5/10/83 Reactor power was maintained at approximately 90 percent power (the station was limited to a power level of 90 percent due to steam leaks on both moisture separator high level drain lines to the condenser) until 1027 hours0.0119 days <br />0.285 hours <br />0.0017 weeks <br />3.907735e-4 months <br /> on May 10, 1983, when the reactor tripped on high Reactor Coolant System (RCS) pressure due to a loss of one of the four 120 VAC essential instrument buses (Y4). This was caused by a fuse.

failure due to water spraying onto the plastic covering a rectifier.

The reactor was critical at 1518 hours0.0176 days <br />0.422 hours <br />0.00251 weeks <br />5.77599e-4 months <br />, and the turbine was synchronized on line e,t 2025 hours0.0234 days <br />0.563 hours <br />0.00335 weeks <br />7.705125e-4 months <br /> on May 10, 1983.

5/11/83 - 5/31/83 Reactor power was increased to approximately 90 percent power and maintained until 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on May-29, 1983, when it was reduced to 74% power. This reduction was due to low load requirements over the holidays. Reactor power was increased to 90 percent of full power at 0800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> on May 31, 1983.

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' REFUELING INFORMATION DATEt May, 1983 i

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 s

-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 procedurcs.

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 Assemblics i

8.

The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or

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is planned, in number of fuel assemblies.

i Present: 735' Increase size by: 0 (zero)'

9.

.The projected date of the last refueling that can be discharged to j

the spent fuel pool assuming the present licensed capacity.

I Date:

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

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COMPLETED FA_CILITY CHANGE REQUEST FCR NO:

77._412 SYSTEM: Hydrogen Dilution System COMPONENT: Blower _1-1 Outlet Pressure Gauge, PI5147 s.

CHANGE, TEST OR EXPERIMEN_Tr FCR'77.-412 was completed September 29, 1980..

This involved the relocation of tubing.from_PI5147 to its source valve,'.

CV5147.

REASON FOR CHANGE:

In order to p rmit installation of a ladder in Penetra-

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tien Room #4, the relocation of.thlettybing was necessary.'

SAFETY EVALUATION: Thereroutingof*the'Erainlessstee'ltubingfrom PI5147 to the source valve for PI5147 has n'o't affect'ed the safety function of th'~e hydrogen dilution system. The quality status of the installation was maintained. Therefore, c unreviewed safety question exists.

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w COMPLETED FACILITY CHANGE REQUEST FCR NO: 78-282 SYSTEM: Essential 4.16 KV Permissive undervoltage reset COMPONENT: '

.A CHANGE, TEST OR EXPERIMENT: FCR 78-282 provided a means to allow an operator to manually close the tie breaker between essential and non-essential 4.16 KV. busses during the loss of offsite power. This would assure that safety loads would be sequcnced if a safety activation signal were to occur. Work was completed March 5, 1981.

REASON FOR CHANGE: Station Emergency Blackout Procedure EP J202.02 states that on loss of offsite power with a diesel generator supplying power to an essential 4.16 KV bus, the tie breaker to the associated non-essential bus could be manually closed. Closure of this tie breaker was sufficient to cause reset.

If a safety activation was to occur at this time the

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safety loads would not be sequenced.

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SAFEYY EVALUATION: The system,.as it was previously designed, did not account for the possibility that a safety signal might follow a loss of offsite power, where the operator had manually energized the non-essential

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4,16 KV bus from the essential 4.16 KV bus.

I Since there is some additional capacity.on the diesel generators above

,that.which is requiced for safety equipment, it is desirable to continue

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the use of feeding,C2 (D2) from C1 (D1) after loss of offsite power;

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-reference Station Emergency Blackout Procedure EP 1202.02. The addition of these Joads does3not negate the previous safety analysis on the emer-gency diesel generators. This FCR provided proper means to insure safety equipment sequencing under the condition of loss of offsite power followed 2

,by safety features activation.

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'FCR NO: 79-168 SYSTEM: Borated Water Storage Tank (BWST)

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. COMPONENT: Pipe HCC-91, Circuits 17, 21, 20, and 24 j

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CHANGE, TEST OR EXPERIMENT: Work; implemented by FCR 79-168 was completed lt March 27, 1981. Redundant circuits 17 and 21 were shortened to extend only for the portion of the BWST minimum circulation pipe within the pipe tunnel. The thermocouples for these circuits were then relocated as-appropriate. Redundant circuits 20 and 24 were then added to protect the c

portion of,the pipe outside of the pipe tunnel from freezjng, and the thermocoupled for these circuits were located near the top of the BWST.

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REASON FOR CHANGE: As originally installed, with t'he'thermoc'ouples inside the pipe tunnel, the system was inoperable and the portion of pipe above ground level froze. SeeLicenseeEventReportNP-32-7j-03(79-034).

4 SAFETY EVALUATION: This FCR= involved redesigning the freeze protection for three inch pipe HCC-91. 'The new freeze protection design consists of two circuits for the portion outside the pipe tunnel and two circuits for the portion inside the pipe tunnel'with separate controlling thermostats and thermocouples.

This change did not adversely affect the function of the high pressure injection pump to the BWST minirum circulation pipe, HCC-91.

It enhances the operation of the minimum recirculation line by adding more reliability to. insure that the line does not-freeze during cold weather.

This is not an unreviewed safety question.

o COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-209 SYSTEM: Emergency Diesel Generators (EDG)

COMPONENT: Visual warning lights CHANGE, TEST OR EXPERINENT: This FCR provided for a blue strobe light in each EDG room. These lights are illuminated when the EDG receives an emergency start signal and are automatically extinguished when the emergency signal is cleared. When the EDG test is begun, the blue lights are illuminated momentarily to establish the operability of the warning lights. The lights are then ready to alarm if an emergency start signal occurs during the test. This change was completed September 10, 1982.

REASON FOR CHANGE: _This will notify the operator at the diesel generators that an emergency start signal is present.

SAFETY EVALUATION: Portions of the work required for FCR 79-209 were safety related in that there were PICA requirements, and "Q" core drills were involved.

Installation in accordance with PICA requirements assured that no new adverse environments were created for safety equipment.

Proper control of the "Q" core drill reports assured that the penetration of "Q" barriers did not result in an unreviewed safety concern. No unreviewed safety question exists.

COMPLETED FACILITY CHANGE REQUEST FCR NO: 81-146 SYSTEM: 240 Volt Motor Control Centers COMPONENT: Transformers YE2 and YF2 CHANGE, TEST OR EXPERIMENT: Work implemented by FCR 81-146 was completed on September 27, 1982. This involved increasing the capacity of the 3 KVA transformers associate d with motor control centers YE2 and YF2 to 30 KVA.

REASON FOR CHANGE: Increasing the capacity of these transformers will decrease the voltage drop through the motor starts. This will ensure the containment air sampling isolation valves to close within the time require-ments of Technical Specification 3.6.3.1.

SAFETY EVALUATION: This FCR consisted of replacing an existing 480/240 volt motor control center 3 KVA transformers, numbers YE2 and YF2, pre-viously located inside motor control centers YE2 and YF2, respectively with 30 KVA transformers mounted outside of the motor control centers, The safety function of these transformers is to supply power to 240 and 120 volt motors actuated by safety signals.

This change insures availability of more than minimum starting voltage on motor control centers YE2 and YF2 to feed the Class 1E motors during a simultaneous start of safety actuation loads on all Class 11 busses. This is the desired result of this change and, therefore, does not create an adverse environment.

'this is not an unreviewed safety question.

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m COMPLETED FACILITY CHANGE REQUEST FCR'NO: 82-031 SYSTEM: Non-Radioactive Heating and Ventilation COMPONENT: Temperature switches TS5315 and TS5318 CHANGE, TEST OR EXPERIMENT: FCR 82-031 was implemented to change the setpoint of TS5315 and TS5318 to 60 F and 84'F i 5*F from their previous setpoints of 80*F and 104*F i 5'F, respectively. Work was completed June 29, 1981.

REASON FOR CHANGE: These new setpoints will provide an additional margin in activating the ventilation fans in the Low Voltage Switchgear Room prior to reaching a temperature which may potentially damage equipment.

SAFETY EVALUATION: The hes, ting and ventilating systems in the acn-radioactive area are designed to provide a suitable environment for equipment and personnel. Airborne radioactivity levels inside these systems are not significant, requiring no special treatment of exhaust air.

1 Lowering the setpoint of the temperature switches has not compromised the function of the ventilation fans. The safety of the unit _has been enhanced by providing equipment reliability in the Low Voltage Switchgear Room.

This is not an unreviewed safety question.

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TOLEDO

%m EDISON June 9, 1983 Log No. K83-852 File: RR 2 (P-6-83-05)

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

20555

Dear Mr. Haller:

Monthly Operating Report, May, 1983 pavis-sesse Nuclear Power Station Unit 1 Enclosed are ten copies of the Monthly Operating Report for Davis-Besse Nuclear Power Station Unit 1 for the month of May, 1983.

Yours truly, b

M Terry D. Murray Station Superintendent Davis-Besse Nuclear Power Station TDM/BMS/ljk Enclosures cc:

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

1 copy Mr. Richard DeYoung, Director j

Office of Inspection and Enforcement Encl: 2 copies Mr. Tom Peebles NRC Resident Inspector Encl:

1 copy h

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

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