ML20125D334

Monthly Operating Rept for Dec 1979

ML20125D334
Person / Time
Site:	Davis Besse
Issue date:	01/08/1980
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML20125D326	List: ML20125D322 ML20125D334
References
NUDOCS 8001140255
Download: ML20125D334 (10)

Text

AVERAGE DAILY UNIT POWER LEVEL v

DOCKET NO.

Davis-Besse Unit 1

.lT January 8, 1980 DATE Erdal Caba COMPLETED BY 419-259-5000, Ext.

TELEPfl0NE 236 December, 1979 MONTil i

DAY AVER AGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (MWe-Net )

(MWe Net) 0 1

0 37 0

2 0

gg 0

0 39

,3 0

0 4

20 0

0 21 5

0 0

22 6

0 0

7 23 0

0 24 8

0 0

9 25 0

10 0

26 O

0 II 27 0

0 12 28 0

0 29 13 0

0 30 14 0

0 15 31 0

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

90017282 em, N

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'4 OPERATING DATA REPORT DOCKET NO. 50-346 DATE January B, 1980 COMPLETED BY Erdal Caba TELEPilONE 419-259-5000, Ext.

236 OPERATING STATUS Davis-Besse Unit 1 Notes

1. Unit Name:

December, 1979

2. Reporting Period:

2772

3. Licensed Thermal Power (MWt):

925

4. Nameplate Rating (Gross MWe):

906

5. Design Electrical Rating (Net NWe):

p; s'

6. MEimum Dependable Capacity (Gross MWe): to be deterrained

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t be determined l

7. Masimum Dependable Capacity (Net MWe):

(/

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

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i None

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

10. Reasons For Restrictions,if Any:

This Month Yr. to Date Cumulative i

744 8,760 20,525

11. Ilours In Reporting Period 29.1 4,332.4 10,964.2

12. Number Of flours Reactor Was Critical 0

2,085.5 2,875.8

13. Reactor Reserve Shutdown Hours 0

4,141.6 9,874.8

14. Hours Generator On Line 0

1.728.2 1.728.2

15. Unit Resene Shutdown ilours 806 10,011,937 20,199,507

16. Gross Thermal Energy Generated (MWil) 0 3,339,756 6,723,511 j

17. Gross Electrical Energy Generated (MWil) 0 3,129,118 6,170,578

18. Net Electrical Energy Generated (MWil) 0 47.3 49.5

19. Unit Service Factor 0

67.0 58.8

20. Unit Availability Factor to be determined

21. Unit Capacity Factor iUsing MDC Net) 0

,_39.4 36.2 J

22; Unit Capacity Factor (U>ing DER Net) 100 28.5 27.6 l

23. Unit Forced Outage Rate

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

Refueling Outage, March 1980 12 weeks January 5, 1980

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

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

Forecast Achiesed

. INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCI A L OPERATION 90017283 m72)

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OPERATIONAL SUFD!ARY

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DECEMBER, 1979 The unit-shutdown which' was initiated on November 30, 1979 to investigate the motor lower bearing oil level alarm on Reactor Coolant Pump (RCP) 1-2; to fix.

Group 7 Rod 5 and Group 5 Rod 11 APIS was still in. progress throughout the month of December. The following is a list of-the major work items performed during the outage:

1.

The RCP 1-2 motor lower bearings oil leak was thoroughly investigated.

' j No. concrete reason for the low IcVel was found.- It was possibly due to filling the motor.when it was running which caused a syphon effect.

2.

Group 7 Rod 5 and Group 5 Rod'11 APIS were fixed.

3.

Shortly after the shutdown, RTD RC3A3 failed, and it was though~t that there was a primary leak through the RTD.

It was later determined that the leakage was through the furmanited gasket and into the thermocouple extension piece. The furmanite box had come into contact with a Reactor Coolant System (RCS) piping whip restraint bending the RTD boss and RTD, j

cracking the extension piece. This in turn allowed water to leak into

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the RTD head and short out_the signal.

Subsequently, it was found that

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the two furmanited RTDs on Loop 2 were both damaged by the furmanite boxes contacting whip restraints. On Loop 1 there was a slight dent in the bosses due to the whip restraints contacting the pipe. _ Three of the J

bosses were replaced and new well type RTDs were also installed. The 1

repairs to the bosses _ were completed on December 26, 1979.

4.

Due to the extended outage several eighteen month' surveillance tests were completed.

90017284 4

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50-346' DOCKET NO.

Davis-Fesse Unit l' UNIT SHUTDOWNS AND FOWER REL" '.TIONS UNIT NAME January 8. 1930 DATE COMPLETED BY Erdal Caba REPORT MONTil December, 1979 TELEPi!ONE 419 M9-5000 Ext. 236 Cause & Corrective Eg 3

Y Licensee Eg o

No.

Date g

5=

jis Event gg 9's Action to Prevent Recurrence yU mv H

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Report e 6

20 79 11 30 S

11.4 B

1 NA NA NA Maintenance outage due to a low (Continued) bearing oil level alarm on Reactor F

732.6 Coolant Pump 1-2.,

see Operational Summary for further details.

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CD C3 m

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- N I 2

Method:

Exhibit G-Instructions W

F: Forced Reason:

1-Manual for Preparation of Data W

5: Schedu!ed A-Equipment Failure (Explain) 2 Manual Scram.

Entry Sheets for Licensee B-Maintenance of Test Event Report (LER) File (NUREG-3-Automatic Scram.

C Refueling 4-Other (Explain) 0161)

D Regulatory Restrictiim E Operatur Trainin;& License Examination 5

F-Administ rative

' Exhibit I - Same Source G-Operation.il Enor (Explain)

(9/77)

I1.Other (INplain) p-

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DATE:

Decenher, 1979

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REFUELING INFORMATIO.1 Davis-Besse Nuclear Power Station Unit 1 1.

Name of f acility:

March, 1980 Scheduled date for next refueling shutdown:

2.

June. 1980 Scheduled date for restart following refueling:

3.

Will refueling or resumption of operation thereaf ter require a technical If answer is yes, whar, 4.

specification change or other license amendment?If answer is no, has the reload f in general, will those be?

r Plant Safety Review Committee and core configuration been reviewed by you' to determine whether.any unreviewed safety questions are associated with the core reload (Ref.10 CFR Section 50.59)?

Yes, see attached Scheduled date(s) for submitting proposed licensing action and supporting 5.

information.

December, 1979 Important licensing considerations associated with refueling, e.g., new or

~ ifferent fuel design or supplier, unreviewed design or performance analysis 6.

methods, significant changes in fuci design, new operating procedures.

d The spent fuci pool capacity expansion program was approved by the NRC 1, 1979.

in Amendment 19 to the operating license received August The number of fuel assemblics (a) in the core and (b) in the spent fuel 7.

storage pool.

0 (zero) 177' (b).

(a)

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, 8.

in number of fuel assemblies.

0 (zero)

Present 735 Increase size by The projected date of the last refueling that can be discharged to.the spent 9

fuel pool assuming the present licensed capacity.

j 1989 (assuming ability to unload the entire core into the spent fuel pool is maintained and the unit goes to an 16 monta retueling cycle)

Date-9.0017286

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i REFUELING INFORMATION Continued Page 2 of 2 December, 1979 4.

The following Technical Specifications (Part A) will require revision:

2.1.1 & 2.1.2 - Reactor Core Safety Limits (and Bases) 2.2.1 - Reactor Protection System Instrumentation Setpoints (and Bases) 3.1.3.6 - Regulating Rod Insertion Limits 3.1.3.7 - Rod Program 3.2.1 - Axial Power Imbalance (and Bases)

The following Technical Specifications (Part A) may also require revision:

3.1.2.8 & 3.1.2.9 - Borated Water Sources (and Bases) 3.2.4 - Quadrant Power Tilt (and Bases) 3.2.5 - DNB Parameters (and Bases) 90017287 f

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g-COMPLETED FACILITY CHANGE REQUESTS i

FCR NO:

79-378 (including Supplements 1 through 4)

SYSTEM: Various Class 1E Systems COMPONENT:

125 VDC Control Relays CHANGE, TEST, OR EXPERIMENT:

On November 19, 1979, the installation and testing associated with the implementation of FCR 79-378 and its four supplements was com-

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plcted. As requested by the FCR, arc suppression diodes were installed in various saf ety related and non-saf ety related circuits where the contacts of Couch relays are controlling highly inductive loads (generally the coils of other relays). A total of 26 safety related circuits was involved including 4 control relays in the auxiliary feedwater pump controls, 18 in switchgear controls, and 4 in emergency diesel generator controls.

l REASON FOR THE FCR:

It was found that failures of Couch relays which occurred in the reactor coolant pump interlock circuitry (see Licensee Event Report NP-33-7 9-126) had been caused by the high transient voltage which occurs when the Couch relay contacts interrupt the current flow through a highly inductive load.

The high voltage arcs across the contacts of the Couch relays resulting in a short circuit. The diodes added under this FCR preclude this from occurring by shunting out the transient voltages through a path separate of the relay contacts.

SAFETY EVALUATION: This change will not adversely affect the function of the relays.

It will improve the reliability of the circuits by providing a path for the discharge of the energy in the coil through the diode, rather than through the contacts of the interrupting relay. This is not an unreviewed safety gyestion.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-405 SYSTEM: Reactor Protection System (RPS)

COMPONENT: Power Range Nuclear Instrument (NI) 5 CHANGE, TEST, OR EXPERIMENT: On November 15, 1979, the physical work and testing associated with the implementation of FCR 79-405 was completed.

This FCR moved cables 2LRPSA01C and 2LRPSA01X, which are the leads for the bottom ion chamber signals of NI-5, from penetration cable GW3 to penetration cable DW3 both located in penetration P2L4EX/P2L4GI.

REASON FOR THE FCR: During the performance of IC 2002.03, NI Detector Post Installa-tion Test, it was found that the inner and outer shields on penetration cable GW3 were shorted together.

SAFETY EXPLANATION: The change outlined was reviewed on Bechtel Drawing E-530 and determined to have no adverse impact on safety. The change of modules within This penetration P2L4G will utilize a spare which was provided for this purpose.

is not an unreviewed safety question.

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OPERATIONAL

SUMMARY

NOVEMBER, 1979

.The unit shutdown which was initiated on October 25, 1979, when Reactor Coolant Pump (RCP) 2-2 tripped from a blown fuse, continued until 1811 hours0.021 days <br />0.503 hours <br />0.00299 weeks <br />6.890855e-4 months <br /> on November 20, 1979, when the turbine-generator was synchronized.

Below is a list of the major work items performed during the outage:

1.

All four RCP seals were replaced.

2.

A new expansion joint for 1-4-2 heater extraction was rewelded.

3.

A design change to the Couch relays in the RCP starting interlock circuits was made. This initiated a similar change in sixteen safety related cir-cuits.

4.

NI-3 was pulled and a new detector was installed. A new section of cable between the containment penetration and the NI-l pre-amp was installed.

1 5.

Several pir restraints were modified as a result of IE Bulletin 79-14 findings.

11/20/79 - 11/21/79 The unit returned on line at 1811 hours0.021 days <br />0.503 hours <br />0.00299 weeks <br />6.890855e-4 months <br /> on November 20, 1979, and reactor power was increased to 81% of full power with generator gross load at 760 MWe on November 21, 1979.

11/22/79 Reactor Coolant System (RCS) flow inidcation was determined to be less than required by Technical Specification 3.2.5 at 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br />.

Reactor power was decreased to 73% at which point RCS flow indication was back within allowable limits. At 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> generator gross load was 684 MWe.

11/23/79 - 11/25/79 Reactor power was maintained at approximately 73% until 0620 hours0.00718 days <br />0.172 hours <br />0.00103 weeks <br />2.3591e-4 months <br /> on November 25, 1979, when RCS flow indication was again determined to be less than that required by Technical Specifi-cations.

Reactor power was decreased to 67% at which point the RCS flow indication was again back within tolerance. The cause of the low flow indication is still being investigated.

11/26/79 Reactor power was maintained at approximately 69% of full power.

At 2145 hours0.0248 days <br />0.596 hours <br />0.00355 weeks <br />8.161725e-4 months <br /> on November 26, 1979, RCP 1-2 was manu-ally tripped due to a lower motor bearing low oil level alarm.

11/27/79 - 11/28/79 The unit remained at approximately 68% of full power with generator gross load of 635 MWe until 0645 hours0.00747 days <br />0.179 hours <br />0.00107 weeks <br />2.454225e-4 months <br /> on November 28, 1979, when it was determined that the power range nuclear instrumentation (NIs) were reading slightly greater than 2%

below the calculated heat balance.

Reactor power was reduced to 507-to bring the NIs in tolerance and to run another heat balance to get the NIs in tolerance.

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90017290

OPERATIONAL

SUMMARY

,. NOVEMBER, 1979 PAGE 2 0F'2 11/29/79 Group 7, Rod 5 was declared inoperahle at 0640 hours0.00741 days <br />0.178 hours <br />0.00106 weeks <br />2.4352e-4 months <br />.initiat-1 ing.a reduction in reactor power to approximately 42% and reset-ting the high flux trip-setpoint 'to 54%.

11/30/79

- Reactor power was reduced and.the turbine-generator taken off line at.1910 hours0.0221 days <br />0.531 hours <br />0.00316 weeks <br />7.26755e-4 months <br /> to investigate the lower motor bearing oil level alarm on RCP 1-2; to fix Group 7,- Rod 5 and Group 5, Rod 11 API.

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