ML19242D070

Monthly Operating Rept for Jul 1979

ML19242D070
Person / Time
Site:	Davis Besse
Issue date:	08/07/1979
From:	Caba E TOLEDO EDISON CO.
To:
Shared Package
ML19242D067	List: ML19242D066 ML19242D070
References
NUDOCS 7908140465
Download: ML19242D070 (11)

Text

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___, ,g AVERAGE DAILY UNIT POWER LEVEL 50-346 DOCKET NO.

UNIT Davis-Besse August 7, 1979 DATE _

COMPLETED BY Erdal Caba TELEPl{ONE 259-5000 Ext. 236 Julv, 1979 MONTil DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-NetI 0 882 g

g7 0 18 878 2

0 881 3 39 879 4 . 0 :o 0

887 5

21 869 6

0 22 0 873 7 23 0 24 871 8

0 2.' 872 9

0 866 10 26 0 27 874 11 259 873 12 28 440 077 13 29 739 876 14 30 798 3, g77 15 847 16 INSIRL*CTIONS On this format. fist the average da !y unit power level in MWe Net for each day in the reporting nonth. Coinpure to the nearest whole megawart.

(9/771 790814ve -

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. .a OPERATING DATA REPORT

- DOCKET NO.

50-346 DATE August /, 1979 CO3!PLETED BY E. Caca TELEPHONE 239-5000 Ext. 236 OPERATING STATUS Notes

1. Unit Name:

Davis-Besse Unit 1

2. Reporting Period; Julv. 1474

3. Licensed Thermal Power t.1Wtp:3 2772

4. Nameplate Rating (Gross alwel: 425 906

5. Design Electrical Rating (Net 31We):

6. 512ximum Dependable Capacity (Gross 31We): To be det.

7. Afaximum Dependable Capacity iNet 31We): ro be det.

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

Zero (until Julv 6, 1979)

9. Power Level To which Restricted. !f Anv (Net 3!We):

10. Reasons For Restrictions. lf Any:

'iRC OIE Bulletins and Shutdown Orcers

' Ibis 51onth Yr.-to-Date Cumulative 744 5,087 16,852

11. Ilours In Reporting Period 8,877.2 498 2,245.4

12. Number Of Hours Reactor was Critical 264.2 1,876.4 2,666.7

13. Reactor Reserve Shutdown Hours 7,833.1 479.8 2,154.9

14. liours Generator On-Line 1,723.2 264.2 1,728.2

15. Unit Resene Shutdown Hours 5,109,543 13.297,118 1,230,451

16. Gross Thermal Energy Generated ISIWH)

17. Gross Electrical Energy Generated (SIWH) 409,950 _ _ ,

1.703.279 5.0Q6.973 381,814 1,394.372 a.635,332

18. Net Electrical Ener;y Generated (5th H) 64 5 42.4 48.3

19. Unit Senice Factor 100 76.3 59.9

20. Unit Availability Factor To be det.

21. Unit Capacity Factor (Using SIDC Net) 56.6 34.6 33.7

22. Unit Capacity Factor (Uung DER Net) 0 3.8 21.3

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Oser Next 6 31onths (Type. Date.and Duration of Each):

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

26. Units In Test Status tPrior to Commercial Operation): Forecast Achiesed INITIA L CRITICALITY INITIAL ELECTRICITY CO.\ t3tERCIAL OPEx u iu.N _

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t DOCKET NO.

50-345 UNIT SIIUTDOWNS ANDIOWER REDUC 130NS- UNIT NAME _ _ Dwi s-Paano Mq

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REPORT MONTil July. 1979 COMPLETED llY y gggg,IIONE l'.rit a 1 caba 259-$000 Ext. 236 "L -

l Licensec Eg Cause & Corrective ,

.  ; .$ 2 "h .h* -

98 Action to No. Date g 3g s jg& Event 37 8O Prevent Recurrence H

f5 $ 3 ;p,

<. =g Reporia in L o .

O i The unit remained in an outage l 9 79 03 30 S 264.2 D 1 N/A N/A N/A until July 12, 1979. Refer to  !

the outage summary of July, 1979, for further detail's.

3 4 I 2 Method: Exhibit C o lnstructions F: Forced Reason:

T- S: Schedu!ed A Equipment Failure (Explain) l-Manual for Preparation of Data Entry Sheels for Licensee B. Maintenance ci Test 2 M.inual Scram.

3- Autonutie Scrain. Event Iteport (t.Elt) File (NUREG-C-Itefueling Olb!)

I D lterniatory Itesteiction 4-Ottier (Explain) 1 ], .. i Operator Tr.iining & License Examination

'- F- Ao onmtiative Exhibit I - Same Source G Opei nional I nor (1:xplain)

(9/77) 11 Oihei il splain) ,

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OPERATIONAL

SUMMARY

FOR JULY, 1979 The unit outage, which began at 2142 hours0.0248 days <br />0.595 hours <br />0.00354 weeks <br />8.15031e-4 months <br /> on March 30, 1979, was in progress the first nine days of the nonth. The outage was extended longer than antici-pated due to additional NRC startup restraints which were imposed as a result of an ongoing analysis of the Three Mile Island incident. NRC released the unit to go to Mode 3 on July 2, 1979, and the NRC shutdown order was lifted on July 6,1979.

7/10/79 Reactor criticality was established at 2338 hours0.0271 days <br />0.649 hours <br />0.00387 weeks <br />8.89609e-4 months <br />.

At 2339 hours0.0271 days <br />0.65 hours <br />0.00387 weeks <br />8.899895e-4 months <br />, the startup range nuclear instrument NT-1 failed and a reactor shutdown was initiated.

7/11/79 Reactor criticality was re-established at 0558 hours0.00646 days <br />0.155 hours <br />9.22619e-4 weeks <br />2.12319e-4 months <br />.

7/12/79 The turbine-generator was synchronized on line at 0011 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />. Reactor power was increased, but was administra-tively limited to 60%. This limitation was initiated to investigate an asymmetric rod fault for Rod 2 of Group 2.

7/13/79-7/16/79 Reactor power was increased to 90% on July 14, 1979, and maintained until July 16, 1979. Reactor power was increased to 100% full power at 2223 hours0.0257 days <br />0.618 hours <br />0.00368 weeks <br />8.458515e-4 months <br /> on July 16, 1979.

7/17/79 - 7/31/79 The unit was maintained between 99% and 100% reactor power the remainder of the month.

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DATE: July, 1979 REFUELING INFORMATION

1. Name of f acility: Davis-Besse Nuclear Power Station Unit 1 March, 1980

2. Scheduled date f or next refueling shutdown:

Scheduled date for restart following refueling: May, 1980 3.

4. Will refueling or resumption of operation thereaf ter 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 Flant Safety Review Committee to determine whether any unreviewed safety questions are associated with the core reload (Ref.10 CFR Section 50.59)?

Yes, see attached

5. Scheduled date(s) for submitting proposed licensing action and supporting information. December, 1979

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.

The spent f .1 pool capacity expansion program was approved by the NRC in Amendment 19 to the operating license received August 1,1979.

7. The number of fuel asse=blies (a) in the core and (b) in the spent fuel storage pool.

(a) 177 (b) 0 (zero)

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 260 Increase size by 475 (75. :otal)

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

Date March, 1980 - May, 1930 (assuming ability to unload the entire core into the spent fuel pool is maintained. )

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REFUELING INFORMATION (Continued)

Ju2y, 1979 Page 2 of 2

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

2.1.1 & 2.1.2 - Reactor Core Safety Limits (and Eases) 2.2.1 - Reactor Protection System Instrumentation Setpoints (and Bases) 3.1.3.6 - Regulating Rod In* .a 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 Para =eters (and Bases) e q

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FACILITY CHANGE REOUESTS COMPLETED DURING JULY, 1979 FCR NO: 79-151 SYSTEM: Service Water (SW) and Component Cooling Water (CCW)

COMPONENT Component Cooling Water Heat Exchanger Service Water Outlet Valves (SW 1424, SW 1429, SW 1434)

CHANGE, TEST, OR EXPERDIENT: On July 3, 1979, implementation of Facility Change Request 79-151 was completed. This change modified the actuator linkages on valves SW 1424, SW 1429, and SW 1434 to provide a more secure and positive attachnent of the actuator linkage arms to the disk arms of the valves.

REASON FOR THE FCR: The retaining nut for the valve linkage was being loosened by vibt ' tion which caused slippage and misalignment of the valve operator linkage. See License Event Reports NP-33-79-74, NP-33-78-147, NP-33-78-120 for further details.

SAFETY EVALUATION: This change will prevent the head capscrew from backing out and the valve linkage from vibrating loose. The safety function of the service water system will not be adversely affected. This change is expected to increase the reliability of these service water valves. This is not an unreviewed safety question.

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FACILITY CHA'!GE REOUESTS COMPLETED DURING JULY,1979 FCR NO: 79-170 f.5 TEM: Reactor Protection System (RPS) f1MPONENT: High Pressure Trip BL' table Setpoints CHANCE, TEST, OR EXPERIME'IT: On May 26, 1979, work was completed which readjusted the trip setpoints of the Reactor Protection System (RPS) from 2351.4 PSIG to 2296.4 PSIG. The system operating procedure, as well as the applicable surveillance test procedures, have been revised to reflect this change. A request for an amend =ent to Table 2.2-1 of the Davis-Besse Unit 1 Technical Specifications has been submitted in order to reflect the above change.

REASON FOR THE FCR: The reduction in the RPS high pressure trip setpoint was made to preclude actuation of the pressurizer Power Operated Relief Valve (PORV) .

This is in resgonse to Nuclear Regulatory Commission Bulletin 79-05B.

SAFETY EVA1.UATION: The proposed reduction in the RPS high pressure trip set-point does not degrade the safety of the plant and does not invalidate any of the safety analyses presented in the Davis-Besse Unit 1 FSAR or in the safety evaluation submitted to the NRC on December 22, 1978 (Serial No. 475). The possibility of an accident or a malfunction of a different type than any eval-uated in the FSAR is not created. Also, the margin of safety as defined in the bases for technical specification is not reduced. Pursuant to the above, the propc sed change does not involve an unreviewed safety question. For further details, see Toledo Edison response to NRC Bulletin 79-05B.

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FACILITY CHANGE REOUESTS COMPLETED DURING JULY, 1979 FCR No: 79-244 SYSTEM: Main Steam COMPONENT: Snubber mounts on both SR-7 seismic restraint locations CHANGE, TEST, OR EXPERI'fENT: On June 20, 1979, work was completed which extended the length of the snubber mountings on both SR-7 seismic restraint locations lh inches. All applicable drawings were revised by the unit Architecht-Engineer Bechtel Company.

REASON FOR THE FCR: The former mounting arrangement caused these two snubbers to be completely extended when the piping was cold. This change placed these snubbers in the middle of their stroke when cold, as designed. These snubbers were operable and s_ill are operable when ar normal operable temperature.

SAFETY EVALUATION: This change moves the mounting of the snubber such that it will be in the designed position with the piston properly centered. There-fore, operation of the snubber in the unlikely event of a seismic event will be as designed. This is not an unreviewed safety question.

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FACILITY CHANGE REOUESTS COMPLETED DURING JULY, 1979 FCR NO: 79-257 SYSTEM: Main Steam COMPONENT: Seismic restraints SR-2, SR-4. SR-6, and SR-7 CHANGE, TEST, OR EXPERIMENT: On June 27, 1979, work was completed which added webb reinforcing plates to sei.smic restraints SR-2, SR-4, SR-6, and SR-7. Since seismic restraints to SR-2, SR-4, SR-6, and SR-7 are located on both main steam lines, a total of eight seismic restraint locations were affected. The necessity for addition of the web reinforcing plates to these eight seismic restraint loca-tions was determined as a result of an investigation conducted by the unit archi-tect/ engineer, Bechtel Company. All affected drawings were revised to rc#'.ect the change.

REASON FOR FCR: A review of design calculations by Bechtel discovered a design deficiency in that- the impact points of these eight seismic restraint locations

- could have deforme.1 if a design basis earthquake had occured. The addition of thz web reinforcing plates corrects this deficiency. (For further details, see License Event Report NP-32-79-08).

SAFETY EVALUATION: Bechtal Engineering evaluation of as-built seismic restraints on the main steam lines showed that eight seismic restraint locations (four on each main steam line) may be subject to I-beam web deformation. To prevent this web deformation, web stiffeners were welded to both sides of the I-beam web. This will stiffen the web and distribute the load, thus preventing web deformation. This modification will not degrade the safety function of the main steam system. This is not an unreviewed cafety question.

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FACILITY CHANGE REOUESTS COMPLETED DURING JULY,1979 FCR NO: 79-261 SYSTEM: [uxiliaryFeedWater(AFW) System COMPONENT: Auxiliary Feed Pumps and Turbines 1-1 and 1-2 CHANGE, TEST, OR EXPERIMEN'T: On June 28, 1979, a 72-hour endurance run of both auxiliary feedwater pumps was successfully completed. After a subsequent cool-down period, both pumps were restarted and run for one hour. This test was conducted with the unit in Mode 5. No water was pumped into the steam generators; pump discharge was to the station drainage system or back to the condensate storage tank.

REASON FOR THE FCR: The above mentioned test runs were made to fulfill an NRC commitment to verify the auxiliary feedwater pumps would operate properly for an extended period of time.

SAFETY EVALUATION: This test did not degrade the safety of th,e unit since it was conducted in Mode 5, and the AFW system is only required in Modes 1, 2, and 3.

Also, a surveillance test has been conducted upon the system before the unit is started up in order to ensure its operability. This is not an unreviewed safety question.

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