ML20106J429

Monthly Operating Rept for Dec 1984

ML20106J429
Person / Time
Site:	Davis Besse
Issue date:	12/31/1984
From:	Quennoz S, Sarsour B TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
K85-166, NUDOCS 8502150628
Download: ML20106J429 (11)

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 UNIT Davis-Besse Unit 1 DATE January 10, 1985 COMPLETED BY Bilal Sarsour TELEPHONE (419) 259-5000, Ext. 384 MONTH December. 1984 DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 1 0

g7 0

2 0

gg -

0 3

0 0

g9 4

0 20 0

5 0

21 0

6 0

0 22 7

0 0

3 8

0 24 0

9 0

25 0

10 0

26 0

1I O

0 27 I2 0

0 28 13 0

0 29 14 0

0 30 IS 0

33 0

16 0

INSTRUCTIONS 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.

(9/77)

G502150628 841231 gDRADOCK 05000346 76'2,d/

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OPERATING DATA REPORT DOCKET NO. 50-346 DATE January 10, 1985 COMPLETED BY Bilal Sarsour TELEPHONE (419) 259-5000 OPERATING STATUS

-Ext. 384

1. Unit Name:

Davis-Besse Unit #1 Notes

2. Reporting Period:

December, 1984

3. Licensed Thermal Power (MWt):

2772

. 4. Nameplate Rating (Gross MWe):

915

15. 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 f.' anber 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 i1. Hours In Reporting Period 744 8,784.0 56,305.0

12. Number Of Hours Reactor Was Critical 0.0 5,529.0 33,031.5

13. Reactor Reserve Shutdown Hours 0.0 134.8 4,014.1

14. Hours Generator On.Line 0.0 5,489.5 31,641.3 i

15. Unit Reserve Shutdown Hours 0.0 0.0 1.732.5

16. Gross Thermal Energy Generated (MWH) 0.0 13,941,608 74,985,422

17. Gross Electrical Energy Generated (MWH) 0.0 4,554,151 24,846,344

18. Net Electrical Energy Generated (MWH) 0.0 4,291,557 23,290,256

19. Unit Service Factor 0.0 62.5 56.2

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20. Unit Availability Factor 0.0 62.5 59.3

21. Unit Capacity Factor (Using MDC Net) 0.0 55.9 47.3

22. Unit Capacity Factor (Using DER Net) 0.0 53.9 45.7 I

23. Unit Forced Outage Rate 0.0 11.0 17.3

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:

January 11, 1985

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

Forecast Achieved INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPER ATION (9/77)

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50-346 DOCKET NO.

UNIT St!UTDOWNS AND POM.:t REDUCTIONS UNIT NAME Davis-Besse Unit 1.*

DATE January 10. 1985 COMPLETED BY ' Bilal Rarannr REPORT MONTil December. 1984 TELEPilONE (419) 259-5000, Ext. 384 i

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Cause & Corrective k

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Prevent Recurrence a

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84 09 14 S

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4 NA NA NA The unit outage, which began on

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September 14, 1984, was still in

'l progress through the end of Decem-ber 1984.

See Operational Summary for further i

details.

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l F: Forced Reason:

Method:

Exhibit G Instructions 1

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-Rcrueling 3-Automatic Scram.

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

E-Operator Training & License Examination

5-Load Reduction F Administrative 9-Other (Explain)

S G-Operational Eiror (Explain)

Extiibit I Same Source 19/77)

Il-Oiher (Explain)

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o OPERATIONAL

SUMMARY

December, 1984 The unit outage, which began on September 14, 1984, was still in progress through the end of December, 1984.

The follow'ng are the more significant outage activities performed during i

December 1984:

1.

Completed all work on main turbine.

2.

The Integrated Leak Rate Test (ILRT) was successfully completed.

3.

The main condenser hydro ~and leak testing was completed.-

4.

Reactor vessel head was installed, and the Reactor Coolant System was filled.

5.

Completed repair of leaks and retesting of main generator for gas leakage.

6.

A plant heatup was delayed due to damaged connections at the output of the BD transformer.

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1 REFUELING INFORMATION

.DATE: December, 1984 1.

Name of ' facility: Davis-Besse Unit l' 2.

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

Scheduled _date forLrestart'following refueling: Summer, 1986'

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~4.

Will refueling or resumption of operation thereafter require a ftechnical 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, 1986 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) 140 - 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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-COMPLETED FACILITY CHANGE REQUEST ~

FCR NO: 78-036 99Kf 2

SYSTEM:

COMPONENT: Containment Cranes CHANGE, TEST OR EXPERIMENT: FCR 78-036 permitted the purchasing and installation of four hoists in Containment.- Work was completed August 29, 1980.

REASON FOR CHANGE: This alteration will alleviate the workload of the-

. polar crane during refueling outages.

SAFETY EVALUATION: Portions of the work for this FCR are routed near safety related equipment. Modifications under FCR 78-036 did not adversely affect the safety related equipment. Therefore, no unreviewed safety question exists.

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COMPLETED FACILITY CHANGE' REQUEST FCR NO: 79-373

-SYSTEM: Containment: Gas Analyzer _ System COMPONENT: AIT-5027'and AIT-5028 CHANGE, TEST 0R EXPERIMENT: This FCR was implemented to bring about

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several modifications to the Containment Gas Analyzer System.. Work for-this FCR was completed November 23, 1983.

The' containment hydrogen analyzers, AIT-5027 and AIT-5028 were reading values of 0.6% to 0.8% hydrogen concentration when the actual hydrogen concentration was 0.2%.

This offset was due.to the effects-of moisture in the containment atmosphere. Because of this, two modifications were performed. First, the desiccant dryer was removed because it was redundant to the sample cooler and heat tracing.

Second, an instrument zeroing system was added to allow the analyzer to be zeroed by containmeat gas.

This change required the continuous high range indication of containment hydrogen concentration in the Control Room per NUREG 0737. This indicator indicates a range of 0 to 10% hydrogen concentration under both.a negative and a positive pressure.

Finally, this FCR allowed for the incorporation of two electrical analog input signals to the plant computer from source transmitters AT-5027 and AT-5028 with a variable range of 0 - 10%. The 0 to 10% range was added to a selector switch range position by a logic network change and new meter scales.

REASON FOR CHANGE: This FCR was enacted for three major reasons. First, the desiccant dryer was removed, and the instrument zeroing system was i

added to eliminate the erroneous high indication of hydrogen in Containment due to the effects of moisture in the Containment atmosphere on the sensing instrument.

Secondly, the Control Room indicator with a range of O to 10% is the direct result of the NRC's review regarding the Three Mile Island 2 incident. Finally, the computer points were added to the plant computer so an indication of hydrogen concentration would be available 4

here to the Safety Parameter Display System and Station Process Computers.

SAFETY EVALUATION: Since the safety function of the hydrogen analyzers is-not altered but enhanced, an unreviewed safety question is not involved.

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COMPLETED FACILITY CHANGE REQUEST 1FCR NO: 83-068 SYSTEM: Radiation' Monitoring

-COMPONENT: 'RE-8442

'-CHANGE, TEST OR EXPERIMENT:.This FCR performed a 10CFR50.59' evaluation of the-secondary plant; drainage system discharge radiological monitoring configuration..This evaluation originated from the inoperability of a-radiation detector =that was to'be used as a final check for radioactive

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-releases through the storm sewer system from the plant to the Toussaint River. FCR 83-068 was approved for final closeout' October _ 25, 1984.

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REASON FOR CHANGE: This FCR-was requested to justify continued plant -

operation until the' installation of the new storm sewer radiation monitoring-equipment is complete.

SAFETY EVALUATION: The storm water system outlet did not have any radiation-detector device to monitor the activity level in the discharge. At Davis-Besse Unit 1,.the turbine building sump, condensate storage tank room drain, acid neutralizing tank room drain, turbine building office building, and' auxiliary building. roof drains, etc., are discharged to the storm water system and routed _to the Davis-Besse Training Center pond.

Out of these, there are only two sources of which water could be contaminated.

These are:

1)

A primary to secondary side leak (i.e., steam generator tube failure) 2)

Backwash system for the condensate polishers Steam generator tube leaks can be detected by the main steam line radiation monitoring detectors (RE-600 and RE-609) or the steam jet air ejector monitors (RE-1003A and RE-1003B).

There was a remote possibility that component leakages exceeding MPC from the condensate backwash system may flow into the floor drain and discharge offsite not monitored. This is because sampling is done only at the' inlet and outlet of the Davis-Besse Training Center pond on a weekly basis.-

Because of the remote possibility oflany activity exceeding the MPC values as listed in 10CFR20.103 and 20.106 and Appendix B. Tables I!and II, discharging offsite not monitored, it was concluded that the Station should continue to run until the installation of the new radiation monitor.

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Therefore, an unreviewed safety question is not involved.

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. COMPLETED' FACILITY CHANGE REQUEST

'FCR NO: 83-107 SYSTEM: Service Water (SW)-

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COMPONENT: SW-43 CHAhGE, TEST OR EXPERIMENT: ~FCR 83-107 eliminated the disk in check ~ valve.

SW-43.- With the internals removed from SW-43,' operator actions are required under certain conditions. -These actions and conditions are described in-Special Order #24. The.following is-from Special Order f24:

"If initially in a lineupfwith SW #1 side supplying secondary loads and a seismic event were to occur with a: loss of power to the No. 1

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side loads, the following operator actions-would be required:

Close SW-1424 (or~SW-1429 if 1-3~1ined up as 1-1) Component Cooling Water (CCW) heat exchanger outlet isolation valve locally in CCW Pump Room and close SW-1399 locally in the Irtake.

Structure Valve Room to provide-isolation. This will prevent backflow into the #1 SW loop.

If #2 SW side is initially supplying secondary loads SW-43 serves no safety function with respect to the events mentioned, and the above actions are not applicable.

Work was completed October 27, 1983.

REASON FOR CHANGE: The internals of check valve SW-43 had failed and were removed.=

SAFETY EVALUATION: The safety function of check valve SW-43 is to protect against backflow in the SW System from SW Loop 2. to Loop i to prevent an unanalyzed loss of SW from the intake forebay. This assumes a seismic event has cut off the intake forebay from Lake Erie and caused a rupture in the non-seismic portion of the SW System in the Turbine Building.

After the SW System was designed, an analysis was done to investigate a break in the non-seismic 30" line from the SW System to the cooling tower makeup line. This analysis showed that the operator would have three hours to isolate the break with an intake forebay level drop of only two feet. As a result of this analysis, the original safety function of SW-43 is no longer needed. Another analysis shows a far longer break would' result in an acceptable loss of water from the int.ake forebay.

As a result of the above analyses, the elimination of the. disk from SW-43 will not result in an'unreviewed safety question. This change will not create any new adverse environment.

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2 COMPLETED FACILITY CHANGE REQUEST.

FCR NO: 84-048' SYSTEM: Control' Room' Heating, Ventilation, and Air Condition'ing

. COMPONENT: FD-1159 and FD-1160 CHANGE, TEST OR EXPERIMENT:.This FCR allowed for drawing M-027A to be

' revised to saow actual plant conditions. Work was completed May 7, 1984.

REASON FOR CHANGE:.This change was made to show the installed fire l'

dampers FD-1159 and FD-1160, and the direction of air flow on drawing

.M-027A.

FD-1159 is located between Rooms 502 (Control Room Cabinet Room)

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and 510-(Computer Room).

FD-1160 is located between Rooms 511-(Shift Supervisor's Office) and 512 (General Office). Both. dampers allow the transfer of ductless air. Drawing M-027A has all the ductless air flow exhausting from these rooms via "to ceiling space".

This drawing does not i

show the supply side of ductless air into the rooms.

SAFETY EVALUATION: Because this change does not. decrease the margin of.

safety set forth in the Davis-Besse Unit 1 Technical Specifications, an unreviewed safety question does not exist.

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TOLEDO

%ms E eISON January 10, 1985 Log No. K85-166 File:

RR 2 (P-6-84-12)

Docket No. 50-346 Licer.se 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, December 1984 Davis-Besse 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 December 1984.

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

Yours truly, ppm Stephen M. Quennoz 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/1 NRC Resident Inspector LJK/002 7

THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MAO! SON AVENUE TOLEDO. OHIO 43652 gg C