ML20203E867

Monthly Operating Rept for Jan 1986

ML20203E867
Person / Time
Site:	Davis Besse
Issue date:	01/31/1986
From:	Khazrai M, Storz L TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
KB86-0177, KB86-177, NUDOCS 8604240468
Download: ML20203E867 (10)

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 g7c4;7 Davis-Besse Unit 1 February 10. 1986 DATE COMPLETED BY Morteza Khazrai TELEPHONE (419) 249-5000 Ext.

MONTH

' January, 1986 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net!

1 0

17 0

2 n

Ig 0

3 0

39 0

4 0

20 0

5 0

21 0

0 6

22 0

7 0

D 0

8 0

0 24 9

0 25 0

10 0

26 0

11 0

27 0

12 0

0 28 13 0

29 0

14 0

30 0

15 0

33 0

16 0

INS'IRtJCTIONS 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.

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OPERATING DATA REPORT DOCKET NO.

50-346 DATE hhrmary 10, 1986 COMPLETED BY.Mo.r.t;;;uLl(hazrai TELEPHONE (419) 249-5000, Ext.

7290 OPERATING STATUS

1. Unit Name:

Davis-Besse Unit 1 Notes

. 2. % Paiod January 1986

3. Licensed Thermal Power (MWr):

2772

4. Nameplate Rating (Gross MWe):

QM

5. Design Electrical Rating (Net MWe):

onA

6. Maximum Dependable Capacity (Gross MWe):

904

~1. !4axnnum Dependable Capacity (Net MWe):

860

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

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

_10. Reasons For Restrictions,if Any:

This Month YrAo-Date Cumalative

(

11. Hours in Repon.g Period 744 744 65.809 l

!2. Number Of Hours Reactor Was Critical 0.0 0.0 35.877.I

13. Reactor Reserve Shutdown Hours 0.0 0.0 4.058.8 i

14. Hours Generator On-Line 0.0 0.0 34.371.8 l

15. Unit Reserve Shutdown Hours 0.0 0.0

1. 732. 5

16. Gmss Thermal Energy Generated (MWH) 0.0 0.0 81.297.600

- 17. Gross Electrical Energy Generated (MWH) 0.0 0.0 25.233.I77 0.,0,

0.0 26,933,622

18. Net Electrical Energy Generated (MWH)

' 19. Unit Service Factor 0.0 0.0 52.2

20. Unit Availability Factor 0.0 0.0 54.9

21. Unit Capacity Factor iUsing MDC Net) 0.0 0.0 44.6

22. Unit Capacity Factor (Us;ng DER Net) 0.0 0.0 42.3 I

23. Unit Forced Outage Rate 100.0 100.0 26.:

AL Eb W!=us Scheduled Over Next 6 Months (Type. Date.and Duration of Each1:

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

26. Units In Test Status (Prior o Commercial Operation):

Forecast Achiesed INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIA L OPERATION (9/77 )

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DOCKET Nob

~ 50-346' UNIT SMrfDOWNS AND POVER REDUCTIONS UNIT MAME.

D:vix-Berse Unit

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DATE February 10, 1985 COMPLETED BY-Morteza Khazrai

. REPORT MONTH January, ~ 1986 TELEPHONE.

.(419) 249-5000 Ext. 7290;

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Licensee a,e n

Cause & Corrective 1'

1 UE i UE Event 3I gj e

Action to.

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.No.

Date N

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Roport #

EO 5o l'revent Recurrence 85 fi[

a 7

85 06 09 F

744 A

4 LER 85-013 JK SC The unit remained shutdown follow-

on't ing the reactor trip on June 9, 4

1985 j

See Operational Summary for further l

details.

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j F: Forced

. Reason:

Method:

Exhibit G - Instructions 4

i S: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparation of Data B-Maintenance or Test 2-Manual Scram Entry Sheets for Licensee 3

C-Refueling 3-Automatic Scram Event Report (LER) File j

D-Regulatory Restriction 4-Continuation from (NUREG-0161)

E-Operator Training & License Examination Previous Month F-Administrative 5-Load Reduction G-Operational Error (Explain) 9-Other (Explain)

Exhibit I - Same Source (9/77)

H-Other (Explain)

4 OPERATIONAL

SUMMARY

JANUARY 1986 The unit-remained shutdown the entire month of December following the reactor trip on June 9, 1985.

Investigation of the causes of the event and corrective actions' continue. See NUREG 1154 for further details.

Below are some of the major activities performed during this month:

1) Completed reinstallation of Power Operated Relief Valve (PORV)

2) Completed condenser hydro leak test

3) Repair and replacement of all the hydro motor actuators

4) Completed work on Decay Heat Loop #2

5) Repair condenser internal damage (heater shroud metal)

6) Replacing gasket and studs on steam generator primary side manway

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R UELING INFORMATION

-DATE: J nuary, 1986 gn 1.

Name of facility: Davis-Besse Unit 1

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Scheduled date for next refueling shutdown: March 1, l'987

2..

3.

Scheduled date for restart following refueling: May 10, 1987

-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: 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) 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) 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

c-COMPLETED FACILITY CHANGE REOUEST FCR NO: 77-221 SYSTEM: ' Auxiliary Feedwater

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COMPONENT: Auxiliary Feed Pump Turbine (AFPT) Speed Controller CHANGE, TEST OR EXPERIMENT: This FCR replaced the CW/ AUX and CCW/ AUX telays with relays and contacts that have adequate current ratings.

This FCR also replaced the dynamic braking resistor with a resistor with

. adequate wattage and voltage-rating.

This FCR and work was completed June 1, 1979.

REASON FOR CHANGE: The change was instituted due to the failures of the Couch /Deutsch relays used in the CW/ AUX and CCW/ AUX application.

Supplement 3 to FCR 77-221 involved the addition of a 50 ohm, 90 watt resistor in the AFPT control circuit. This resistor had been previously removed by the EIT for adverse environment.

SAFETY EVALUATION

SUMMARY

The safety evaluation of the Couch relays for CW/ AUX and CCW/ AUX in disconnect switch cabinets CDE 12A-1 and CDF 12A-1 did not work satisfactorily for the following reasons:

1. DC interrupting rating of contacts were not capable to break the

_Bodine motor locked rotor current.

Contacts arced on opening, causing excessive wear.

2. The.Agastat relay is a more robust relay with larger contacts and much better DC interrupting rating.

3. The addition of a 50 ohm, 90 watt resistor in the AFPT control circuit. This additional resistor in JT2703 and JT2704 has been previously reviewed by the EIT for adverse environment.

These changes do not change the function of the AFPT speed control circuit.

Reliability of the circuit and speed control will be enhanced by this modification. This modification will not adversely affect the safety function of the Auxiliary Feedwater System.

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COMPLETED FACILITY CHANGE REQUEST FCR NO:

78-297 SYSTEM:. Reactor Coolant System COMPONENT: RCP Neutron Power Interlock CHANGE, TEST OR EXPERIMENT: This FCR is requesting approval of testing to obtain operational data on the restarting of the fourth Reactor Coolant Pump (RCP) above 22% power. This will confirm,the simulator results

' obtained by B&W. The basic test would be to establish reactor power at 50%

2% with three RCPs running and then start the fourth RCP.

This FCR requested relief from Section 15.2.6.1 of the FSAR (Page 15-47) which describes the 22% interlock.

This FCR test and request were completed February 28, 1979.

- REASON FOR CHANGE:

By' resetting the setpoint of RCP interlock to a higher setpoint, the fourth RCP could be started at a higher power level then the present power level.. B&W recommends this test be performed in SOM #378 dated June 7, 1978.

SAFETY EVALUATION SQKHARY: An analyst: of the data provided by the B&W simulator at 50% full power reveals that the highest and lowest pressure after the starting of the fourth RCP are 2192 psig and 2090 psig, respectively, which are well within the pressure trip setpoints.

If the trip setpoints are reached, the reactor will trip and any limits of the accident analysis in Chapter 15 of the FSAR will not be exceeded.

The 22% interlock is a non-nuclear safety related interlock and was introduced in the original Integrated Control System (ICS) to avoid the possibility of-frequent ' reactor coolant pressure trips.

Based on the analysis of the data, it is informed that starting a fourth RCP at 50% full power will not compromise nuclear safety and.does not involve an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST

' FCR'NO: 80-278 SYSTEM: Safety Features Actuation System (SFAS)

COMPONENT: 'N/A CHANGE, TEST OR EXPERIMENT: This FCR changed the automatic transfer of the suctions of the decay heat and containment spray pumps from the BWST

' to the containment emergency sump to a manual transfer. Work was completed March 18, 1982.

REASON FOR CHANGE: This modification protects the decay heat and containment spray pumps from the possibility of damage due to an inadvertent transfer of pump suction to the dry containment emergency sump.

SAFETY EVALUATION

SUMMARY

The safety function'of this transfer is to assure proper net positive suction to the decay heat and containment spray pumps during recirculation mode of operation.

Manual transfer from the Control Room will be performed when BWST level indication is between 49.5 inches and 96 inches.

Minimum level requirements for the BWST are met when transfer from BWST is performed within four minutes after BWST level has reached 96 inches.

This manual transfer provides the same safety function as the previous automatic transfer; hence, this does not involve an unreviewed safety question.

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• COMPLETED FACILITY CHANGE REQUEST FCR NO:'83-046 SYSTEM: 'N/A COMPONENT: Borated Water Storage Tank (BWST)

- CHANGE, TEST OR EXPERIMENT: FCR 83-046 provided the addition of spray on foam insulation-to the exterior sides of the Borated Water Storage Tank (BWST).

To. protect the BWST from any direct contact with the foam insulation

- material, the tank had a protective base coating applied to its surface.

The BWST was also sealed and top coated. Work was completed October 5 1985.

REASON FOR CHANGE: This change is necessary due co stratification of temperature levels caused by tank heat losses, the uneven mixing'of water, and suspected icing on the inside of the BWST walls. Because of this, inaccurate temperature readings were received during cold weather.

Insulation added to the tank's external walls limits the tank heat losses such that the stratification of temperature levels are minimized and icing conditions are eliminated.

SAFETY EVALUATION

SUMMARY

The safety function of the BWST is to provide a water source for the Emergency Core Cooling System and the Containment Spray System. The BWST temperature should be maintained between 50*F and 90*F.

The low temperature level listed in the Technical Specifications is-35'F.

However, a minimum of 50*F should be maintained based on a containment spray' initiation. Because this modification will not adversely affect'the integrity of the BWST or the operation of the.BWST from performing its design function, an unreviewed safety question does not exist.

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TOLEDO

%s EDISON February 10, 1986 Log No. KB86-0177 File: RR 2 (P-6-86-01)

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, January 1986 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 January 1986.

If you have any questions, please feel free to contact Morteza Khazrai at (419) 249-5000, Extension 7290.

Yours truly, M'N Louis F. Storz Plant Manager Davis-Besse Nuclear Power Station LFS/MK/lj k Enclosures cc:

Mr. James G. Keppler, w/1 Regional Administrator, Region III Mr. James M. Taylor, Director, w/2 Office of Inspection and Enforcement Mr. Walt Rogers, w/1 g

NRC Resident Inspector IJK/002 y

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THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 43652

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