ML20112D898

Monthly Operating Rept for Feb 1985

ML20112D898
Person / Time
Site:	Davis Besse
Issue date:	02/28/1985
From:	Quennoz S, Sarsour B TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
K85-488, NUDOCS 8503220456
Download: ML20112D898 (12)

Text

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s AVER AGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 UNIT Davis-Besse Unit I DATE March 8, 1985 COMPLETED BY Bilal Sarsour TELEPHONE (419) 259-5000 Ext. 384 February 1985 MONTH DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(Mwe. Net) 1 802 17 819 2

818 gg 819 3

818 39 695 4

619 20 691 5

819 21 814 6

799 22 822 7

820 23 805 g

817 24 820 9

817 25 827 10 818 26 826 13 818

- 27 826 12 817 825 28 13 816 29 14 818 30 15 821 3

16 818 INSTRUCTIONS On this format, list the average daily unit power level in MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.

(9/77)

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8503220456 850228 PDR ADOCK 05000346 R

PDR

OPERATING DATA REPORT DOCKET NO.

50-346

-DATE March 8, 1985 COMPLETED BY Bilal Sarsour TELEPHONE (419) 259-5000, Ext. 384 OPERATING STATUS Davis-Besse Unit 1 Notes

1. Unit Name:

2. Reporting Permd.

February 1985

3. Licensed Thermal Power (MWt):

2772

4. Namepiste Rating (Gross MWe):

915

5. Design Electrical Rating (Net MWe):

9M 918

6. Maximum C.,

'X Capacity (Gross MWe):

7. Maximum C.,

' ": Capacity (Net MWe):

874

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

j.

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

10. Reasons For Restrictions.lf Any:

j This Month Yr.-to-Date Cumulative 672 1,416.0 57,721.0 I1. Hours la Reporting Period

12. Number Of Hours Reactor Was Critical 672 1.066.1 34,097.6

13. Reactor Reserve Shutdown Hours 0.0 0.0 4,014.1 I4. Hours Generator On-Line 672 976.0 32,617.3

15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5

16. Gross Thermal Energy Generated (MWH) 1,712,431 2,196,755 77,182,177 l

17. Gross Elect.-ical Energy Generated (MWH) 573,184_,

726,155 25,572,499

18. Net Electrical Energy Generated (MWH) 543,374 672,269 23,962,525

19. Unit Semce Factor 100 68.9 56.5

20. Unit Availability Factor 100 68.9 59.5

21. Unit Capacity Factor (Using MDC Net) 92.5 54.1 47.5

22. Unit Capacity Factor (Using DER Net) 89.2 52.4 45.8

23. Unit Forced Outage Rate 0.0 0.0 16.9 L

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

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

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

Forecast Achieved l

i INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION (9/77) o

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t UNIT SHUTDOWNS AND POb:Gt REDUCTIONS DOCKET NO. 50-346 UNIT NAME Davis-Besse Unit 1 DATE March 8.

1985 i:

REPORT MONTil February'1985 COMPLETED BY Bilal Sarsour

, TELEPIIONE (419) 259-5000, Ext. 384 a'

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Cause & Corrective Licensee

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!gl Action to

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Date I-5g

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Event idu?

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j if, g Report

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L No unit shutdowns or significant f

power reductions

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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

{1 B-Maintenance of Test 2-Manual Scrani.

Entry Sheets for Licensee C Refueling 3-Automatic Scram.

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

E-Operator Training & License Examination l 5-Load Reduction F-Administ rative 9-Other (Explain) 5 1

G-Operatiimal Eirut (Explain)

Exfiibit I - Same Source 19/77) 11-01her (Explain) i 4

OPERATIONAL

SUMMARY

February, 1985 2/1/85'- 2/6/85 Reactor power was maintained at approximately 93% (reactor-power was limited to 93%'due to an inoperable main steam safety. valve) until 0300

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hours on February 6, 1985, when power was reduced to approximately 90% to stop Condensate' Pump'#1 for maintenance inspection and to repair its upper bearing oil cooler lines. After the completion of Condensate Pump #1

-inspection, reactor power was slowly increased to approximately 93% which was reached at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br /> on February 6, 1985.

2/7/85 - 2/20/85 Reactor power was maintained at approximately 93% until 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br /> on 3

February 19, 1985, when a manual power reduction to approximately 88% was l

initiated to perform control rod drive exercising testing.

At approximately.1506 hours0.0174 days <br />0.418 hours <br />0.00249 weeks <br />5.73033e-4 months <br /> on February 19, 1985,.while the unit was at g

g approximately.88%, Control Rod 2-3 dropped due to an improper connection-of.a wire cable connector to the transfer switch. The unit ran back to 52%. The rod was pulled, and reactor power was slowly increased at a steady rate.

f At'2057 hours0.0238 days <br />0.571 hours <br />0.0034 weeks <br />7.826885e-4 months <br /> on February 19,.1985, while the unit _was at approximately 62%, Control Rod 7-12 dropped.' The cause is suspected to be a bad transfer switch. Control Rod 7-12 was realigned with Control Rod Group 7.

i Power escalation continued until 1200 hour0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br />s-on February 20, 1985, when 93%

of full power was achieved. Reactor power level was maintained at 93%

until 1316 hours0.0152 days <br />0.366 hours <br />0.00218 weeks <br />5.00738e-4 months <br /> on February 20, 1985, when the unit ran back to approximately 53% power due to a dropped control Rod 1-3,'which was caused by an improper connection of-a wire cable connector to the transfer switch. Control Rod 1-3 was pulled and power escalation continued until 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br /> on February 20, 1985, when 93% of full power was achieved. These rod drops will be further investigated during the next outage of sufficient. length. The cause for the loose connectors was attributed to the extensive cleanup conducted in the cabinets during the recent refueling outage.

Efforts were made to tighten connections but accessibility is a problem with the cabinets energized.

I-t 2/21/85 - 2/28/85

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Reactor power was maintained at approximately 93% until 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br /> on February 23. 1985, when power was reduced to approximately 72% due to an L

unplanned runback of Group 7 caused by a defective command module in the

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' auxiliary power supply which was controlling Group 7.

Reactor power was slowly increased to approximately 94% which was reached at.2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> on February 24, 1985, and maintained at this power level for l'

the remainder of the month.

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. REFUELING INFORMATION-DATE: February, 1985

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EName of facility:- Davis-Besse Unit 1 22.-

Scheduled date for next refueling shutdown: Spring, 1986 3.

Scheduled _date for restart.following refueling: Summer, 1986

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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: Winter, 1985

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) 204 - 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)

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9.

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

Date:

1992 - assuming ability to unload the entire core into the spent fuel pool is maintained.

BMS/005 L._

COMPLETED FACILITY CHANGE REQUEST FCR NO:

79-430

SYSTEM: Auxiliary Feedwater (AFW)

COMPONENT: - Auxiliary Feedwater Flow Detector CHANCE. TEST OR EXPERIMENT: This FCR installed a nuclear safety related qualified auxiliary feedwater flow detector which met the requirements set forth in Regulatory Guide 1.89 and which is powered from an emergency power source. Work was completed August 4, 1982.

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' REASON FOR CHANGE: The original auxiliary feedwater flow detector was not qualified nuclear grada. The addition of the nuclear safety qualified detector was required to comply with NRC Requirements Task #18 of Lessons Learned dated October'1, 1979.

SAFETY EVALUATION: This change provided additional information to the reactor operator during abnormal reactor operation when auxiliary feedwater flow is required to remove decay heat.

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n-I COMPLETED FACILITY CHANGE REQUEST L

'FCR NO:.81-164

' SYSTEM: Clean Liquid Radioactive Waste COMPONENT: Clean Waste Monitor Tank Transfer Pump CHANCE. TEST OR EXPERIMENT: FCR 81-164 was implemented to replace the Clean Waste Monitor Tank Transfer Pump,.which was a Crane Chem-pump, with two Gould type pumps. This FCR includes the piping, electrical, and other modifications required to install the pumps to the system. Work involved with this FCR was completed April 4, 1984.

REASON FOR CHANGE: This change was needed in that the original Clean Waste Monitor Tank Transfer Pump was of the Crane Chem-pump type and unsuitable for the. intended function of the pump, which pumps water from the Clean Waste Monitor Tank to either the Primary Water Storage Tank, the waste system for further processing, or into Lake Erie.

If the water is sent on for further processing in the waste system, the water will have suspended solids present in it.

The Crane Chem-pumps are excellent for clear water service, but are susceptible to excessive wear and plugging l

when pumping suspended solid mixtures. Also, in the past, the Crane Chem-pumps have had to be returned _ to the manufacturer for maintenance which has shown to take at least one year to perform..This duration for-repair is unacceptable.

SAFETY EVALUATION: The purpose-of the Clean Waste Monitor Tank Pump is to pump water f rom the Clean Waste Monitor Tank to either the Primary Water Storage Tank, the waste system for further processing, or into Lake Erie.

Since this function is not decreased by the change made, an unreviewed~

safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST

- FCR-NO:'82-019 SYSTEM:. Fire Detection System COMPONENT: Fire Door 427A CHANGE, TEST OR EXPERIMENT: This FCR modified door 427A, a 3-hour fire rated door, by installing an electromagnetic hold / release device to hold the door in the normally open position. This-modification is temporary until a permanent battery heating system can be installed in Battery Room #2. Work was completed January 19, 1984.

REASON FOR CHANGE:

The recommended battery room temperature is approxi-mately 60*F at all times.'

In the past, this room has become very cold in the winter months.

The door hold / release device was installed to hold the door open during the winter months. This is to continue until a heating system can be installed in Battery Room #2.

SAFETY EVALUATION:

Because this door is a 3-hour fire rated door, it has been tied into the Fire Detection System and will automatically release in the event of an alarm. Because of this connection into the Fire Detection System, the intended safety function of door 427A is not decreased.'

Therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST

'FCR NO: 84-0046

-SYSTEM: Various COMPONENT: N/A CHANGE, TEST OR EXPERIMENT: This FCR revised the Safety Analysis Report to-incorporate pressure and temperature data for the following systems:

1) Auxiliary Feedwater System

2) Reactor Coolant Letdown System

3) Reactor Coolant Makeup System

4) Reactor Coolant Seal Water Supply System

5) Containment Spray System

6) Low Pressure Injection / Decay Heat Removal System

7) High Pressure Injection System

8) Spent Fuel Pool Cooling System

9) Radiological Monitoring System FCR 84-0046 was implemented January 22, 1984.

REASON FOR CHANGE: This FCR resolved inconsistencies present in drawing 12301-M-602 and the Updated Safety Analysis Report (USAR).

SAFETY EVALUATION:

It is concluded that these changes to drawing 12501-M-602 and the USAR will not result in an unreviewed safety question as defined by 10CFR50.59 or in undue risk to the health and safety of the public.

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i COMPLETED' FACILITY CHANGE REQUEST FCR NO: 84-047 SYSTEM: Various

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. COMPONENT: Drawing'12501-M-602 CHANGE.-TEST OR EXPERIMENT: FCR 84-047 revised drawing 12501-M-602, 1 piping class summary sheets, which corrected certain system pressures and temperatures..The affected-systems are:

1) Makeup and Purification System

2) Component Cooling Water System

3) Service Water System

4) Screen Wash System

5) Gaseous Radioactive Waste System

6) Miscellaneous Liquid Radioactive Waste System

7) Station Fire Protection System

8) Clean Liquid Radioactive Waste System This FCR was completed February 8,'1985.

REASON FOR CHANGE: This FCR was~ transmitted to provide resolution to inconsistencies between drawing 12501-M-602 and the Updated Safety Analysis Report.

4' SAFETY EVALUATION: It is concluded that these changes to drawing 12501-M-602 will not result in an unreviewed safety question as defined by 10CFR50.59 or' result in undue risk to the health and safety of the public.

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o COMPLETED FACILITY CHANGE REQUEST FCR No: 84-129 SYSTEM:

Fire Protection COMPONENT: Jockey Fire Pump, P-6 CHANCE, TEST OR EXPERIMENT: This FCR revised drawing 7749-C-395 to represent the correct "as-built" condition of the Jockey Fire Pump, P-6.

This revision involved updating the concrete pad and anchor bolts for the

. Jockey Fire Pump. FCR 84-129 was implemented December 19, 1984.-

REASON FOR CHANGE: This FCR was the result of work that was performed when the bolts installed to hold the Jockey Fire Pump were found broken off per Non-Conformance Report 84-0015. Therefore, changes to drawing 7749-C-395 were required to represent the correct "as-built" condition of the plant.

SAFETY EVALUATION:

Since this change does not decrease the safety function of the Fire Protection System, an unreviewed safety question does not

. exist.

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,. y TOLEDO

%m EDISON March 8, 1985 Log No. K85-488 File: RR 2 (P-6-85-02)

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, February 1985 Davis-Besse Nuclear Power Station Unit 1 Enclosed are ten copies of the Monthly Operating Report for Davis-Besse Nuclear Power Station Unit I for the month of February, 1985.

'If you have any questions, please feel free to contact Bilal Sarsour at (419) 259-5000, Extension 384.

Yours truly, f%w

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l Stephen M. Quenn Plant Manager Davis-Besse Nuclear Power Station SMQ/BMS/ljk Enclosures cc:

Mr. James G. Keppler, w/1 Regional Administrator, Region III Mr. Richard DeYoung, Director, w/2 Office of Inspection and Enforcement Mr. Walt Rogers, w/l NRC Resident Inspector LJK/002 d

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

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