ML19339C860

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Monthly Operating Rept for Jan 1981
ML19339C860
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 02/06/1981
From: Sarsour B
TOLEDO EDISON CO.
To:
Shared Package
ML19339C859 List:
References
NUDOCS 8102120135
Download: ML19339C860 (11)
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AVERAGE DAILY UNIT POWER LEVEL

$ DOCKET NO. 50-346 Davis-Besse Unit 1 UNIT DATE February 6, 1981 COMPLETED BY Bilal Sarsour TELEPHONE (419) 259-5000, Extension 251 MONTH January, 1981 ,

I DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAILY POh ER LEVEL i (MWe-Net) (MWe. Net) 464 0 g g 2 463 gg 0 3 470 g9 0 4 . 472 20 0 3 444 21 0

121 0 6 22 7 0 23 0 0 0 8 24 9 0 25 0

10 0 - 26 0 II O 27 0 i

12 0 2g 0 13 0 29 0

14 0 30 0 IS 0 33 0 16 0 -

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

  • hole megawatt.

(9/771 O

O.102120\% .

1 - + -w- -m y w -w p.-----m =, e4..,.ee< win - w m -- =waa e #,,e up + 9 ye m?

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, OPERATING DATA REPORT i

DOCKET NO. 50-346 DATE February 6, 1981 ,

COMPLETED BY Bilal Sarscur TELEPliONE ( '.1 0 251-5000,  ;

Extension 251 OPERATING STATUS j

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Davis-Besse Unit . Notes

2. Juporting Period: January, 1981 I 3. Licrnsed Thermal Power t\lWt): 2772 ,
4. Nameplate Rating IGross MWe): 925
5. Design Electrical Rating (Net MWe): 906
6. Maximum Dependable Capacity (Gross MWe): 934
7. Masimum Dependable Capacity (Net alwe):
8. If Changes Occur in Capacity Ratings (Irems Number 3 Through 7) Since Last Report.Give Reasons:
9. Power Lesel To which Restricted. If Any (Net 31We):
10. Reasons For Restrictions,if Any:

i , This Month Yr.-to Date Cumularise

11. flours in Reporting Pen.od 744 744 30,053 i
12. Number Of Ilours Reactor Was Critical 134.0 134.0 14,518.2
13. Reactor Resene Shutdown flours 0 0 2,882.1
14. Ifours Gen:rator On Line 127.9 127.9 13.175.7
15. Unit Resene Shutdown flours 0 0 1,731.42
16. Gross Thermal Energy Generated IMWII) 191,873 191,873 27,096,679 .
17. Gross Electrical Energy Generated (MWil) 62,434 _, 62,434 9,037,768
18. Net Electrical Energy Generated (31Wil) 51,729 51,729 8,316,230
19. Unit Senice Factor 17.2 17.2 44.4 '
20. Unit Asailability factor 17.2 17.2 50.6
21. Unit Capacity Factor ICsing SIDC Net) 7.8 7.8 _

32.9

22. Unit Capacity Factor IUsing DER Net) 7.7 7.7 32.3
23. Unit Forced Outage Rate 82.8 82.8 27.6,
24. Shutdowns Scheduled Oser Nest 6 Months (T)pe. Date,and Duration of Facht:
25. If Shut Down At End Of Report Period. Estimated Date of Startup:

February 3, 1981

26. Units in Test Status IPrior o Commercial Operation): Forecast Achiesed INITitL CRITICA LITY INITIAL ELECTRICITY CO\tMERCI AL OPER ATION (9/77 )

t,

. 50-346 DOCKET NO.

UNIT SilUIDOWNS AND POb:.;t REDUC 11DNS*

UNIT NAME _Davia-Bcasc_L' Februasy o, 1981 nit 1 *g DATE i COMPLE1ED BY Bilai Sirsour REi' ORT MONTil January, 1981 TELEPl!ONE (419) 2S9-5000 Ext. 251{

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x a e_ et Y 186 Licensee E* E% Cause & Corrective ,

1p 53 i 'E $ U Event 55 N ^I""'"

No. Date g3 g gg Report * @G gO Psevent Recurrence O- u my u 6

11 81 1 6 F 616.1 A 1 NP-33-81-01 CB Pla!PXX The unit was shutdown to repair tiie seals in the Reactor Coolant Pumps _

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  • - 1-2 and 2-1. See Operational Summary for further details'.

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- Method: Exhibit G-Instructions F: Fmced Reason:

for Preparation of Data S: Schedu!cd A Equipment Failure (Emplain) 1 Manual 2 Manual Scram. Entry Sheets for Licensee B. Maintenance of Test 3.Autonutic Scram. Esent Repost (LERI File (NUREG-C Refueling D Regulatory Restriction XMLL334G351)Gl0 0161)

E Operator Training & Ucense Examination 4-Continuation 5-Reduction 5 F Adnunntr.itive '

G-Operational l'uor (Expfain) Eshibit i . Same Source 6-Other (9/77) ll Oiher (E xplai..

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'6 OPERATIONAL SLDCL\RY January, 1981 Reactor power was maintained at approximately 55 percent power until 0734 hours0.0085 days <br />0.204 hours <br />0.00121 weeks <br />2.79287e-4 months <br /> on January 6, 1981 when a shutdown was initiated because of problems with the seals in the Reactor Coolant Pumps 1-2 and 2-1. The unit was taken off line at 0751 hours0.00869 days <br />0.209 hours <br />0.00124 weeks <br />2.857555e-4 months <br /> on Janaury 6,1981 and remained 2.here for the rest of the month.

The reactor coolant pump seals were repaired. One new seal cartridge was in-stalled, the other three were rebuilt and installed.

While maintenance was being performed on, an electrical penetration, water was discovered in the containment electrical penetration. An ensuing investigation resulted in finding moisture in all cast electrical penetrations. A comprehen-sive plan to deal with the moisture problem was developed and is being imple-mented.

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DATE: January, 1981 REFUELING INFORMATION Davis-Besse Nuclear Power Station Unit 1

1. Name of facility:

February, 1982

2. Scheduled date for next refueling shutdown:

May, 1982

3. Scheduled date for restart following refueling:
4. Will refueling or resumption of operation thereafter require a technical If answer is yes, what, specification change or other license amendment?

in general, will these be? If answer is no, has the reload fuci 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)?

Reload analysis is scheduled for completion as of December, 1981. No technical specification changes or other license amendments identified to date.

5. Scheduled date(s) for submitting proposed licensing action and supporting information. January, 1982 j 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.

None identified to date.

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7. The number of fuel assc=blics (a) in the core and (b) in the spent fuel storage pool. '44 - Spent Fuel Assemblies 177 (b) 8 - New Fuel Assemblies (a)
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, l in number of fuel assemblies.

i 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.

fuel Date 1988 (assuming ability to unload the entire core into the spent

. pool is maintained)

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\

) COMPLETED FACILITY CilANGE REQUESTS i

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CHANGE, TEST OR EXPERIMENT: On May 25, 1979, modifications to the Pressurizer Pilot Operated Relief Valve (PORV) for FCR 79-169 were completed. The modifications included changing the relief setpoints for the PORV from 2255 psig i to 2400 psig, and changing the low setpoint to 2350 psig.

REASON FOR CHANCE: The above change in the setpoint for the PORV in conjunction with the change in the Reactor Protection System (RPS) high pressure trip setpoint (FCR 79-170) should prevent the actuation of the PORV during anti-

cipated transients, r

SAFETY EVALUAT. ION:

All safety analysis for B & W plants assume that the vent capacity of the ,

i PORV will not be available. Thus, these analysis are unchanged by an increase in its setpoint. -

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! The change proposed in FCR 79-169 does not constitue and unreviewed safety 4 questian for the following reasons.

(1) The probability of occurence or the consequences of an accident or malfunction of equipment important to safety, previously i evaluated int the Final Safety Analysis Report (FSAR) has not been increased.

(2) The possibility of an accident or malfunction of a different type

other than any evaluated previously in the FSAR has not been created.

! (3) The margin of safety a's defined in the basis for any technical specification has not been reduced.

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COMPLETED FACILITY CIIANCE REQUESTS FCR: 80-065 SYSTEM: Lake Water ;fstem COMPONENT: Motor Operated Valves MV2927 and MV2928 CHANGE TEST, OR EXPERIMENT: FCR 80-065 has been written to revise Electrical Drawing No. E-48B Sheet 27, Revision 6 to reflect the as built conditions.

REASON FOR CIIANCE: While investigating NCR 210-78, it was found that more interlocks were connected in the MV2927 and MV2928 control schemes in addition to the TIS contacts shown in the Elementary wiring diagram E-48B, Sheet 27, Revision 6. MV2927 and MV2928 are located in the control scheme inside control panels C6708 and C6709.

SAFETY ANALYSIS: This FCR involves only the revision of drawings to as-built condi-tions and therefore an unreviewed safety question does not exist.

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j COMPLETED FACILITY CHANGE REQUESTS FCR NO: 80-067 SYSTEM: Containment Ventilation System l COMPONdNT: Containment Cooler Fan #3 a

CHANGE, TEST, OR EXPERIMENT: FCR 80-067 has been written to revise Bechtel Drawings E-58B, Sheets 2A and 2C to reflect the as built conditions.

i REASON FOR CHANCE: To correct a possible draf ting error on Drawings E-58B, Sheet 2A and 2C, so as to reflect the as built conditions of the internal wiring of-breakers BE1501 and BF1501.

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SAFETY EVALUATION: This FCR involves only a drawing revision to reflect as built

! conditions, and therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO:80-104 SYSTEM: Reactor Coolant System COMPONENT: RC Flow Transmitters CHANGE TEST EXPERIMENT: FCR 80-104 has been written to revise Bechtel' Drawing M-567 to reflect the actual plan view location of FTRCIA1, A2, A3, A4 and 1A, and FTRClB1, B2, B3, B4 and 1B as shown on Johnson Service Co. Drawings YF-FP-RCOLA Sheet 1, and 1/F-FP-RC01B Sheet 1.

REASON FOR CHANGE: The Bechtel instrument location drawings should accurately reflect the actual instrument location for the purposes of training, trouble shosting and maintenance. Furthermore, this information is extremely important in times of limited access to instruments.

SAFETY EVALUATION: This FCR involves only a drawing revision to reflect as built conditions and therefore, an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REOUESTS FCR NO: 80-122 SYSTEM: Reactor Coolant System COM20NENT: RC42, RC43, RC44, RC45 CHANGE, TEST OR EXPERIMENT: On May 29, 1980, modifications on Reactor Coolant Valves RC42 RC43 RC44 and RC45 for FCR 80-122 were completed. The modifications made to the valves included grinding of the canopy and seal welding the bonnet to the valve using a fillet veld.

REASON FOR CHANGE: In order to remove the bonnet for valve repair, the canopy' weld must be ground out. Tolerances for rewelding the canopy can not be met to allow reassembly via a canopy weld.

SAFETY EVALUATION: Sealing the bonnet to the body with a fillet veld will not affect the ability of valve to perform its safety function. This valve modification has the concurrence of the valve manufacturer, Rockwell International. An un-reviewed safety question does not exist.

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COMPLETED FACILITY CHMGE REQUESTS FCR NO: 80-167 SYSTEM: Seismic Supports COMPONE!Tr: Hanger PSV-1-H1 CHANGE, TEST, OR EXPERIMENT: On June 19, 1980, the rear snubber in hanger PSV-1-H1 was changed from a remote reservoir to a local reservoir.

REASON FOR CHANGE: To enhance snubber maintenance.

SAFETY EVALUATION: Changing the hydraulic fluid reservoir for PSV-1-H1 from remote to local will not affect the functioning of the snubber. Furthermore, proper func-tioning will probably be enhanced, since there will be less opportunity to trap air in the snubber tubing, and hence, adversely affect snubber operations. An unreviewed safety question does not exist.

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