ML19323H608

Monthly Operating Rept for May 1980

ML19323H608
Person / Time
Site:	Davis Besse
Issue date:	06/09/1980
From:	Sarsour B TOLEDO EDISON CO.
To:
Shared Package
ML19323H607	List: ML19323H606 ML19323H608
References
NUDOCS 8006130275
Download: ML19323H608 (8)

Text

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AVERAGE DAILY UNIT POWER LEVEL O -

DOCKET NO. 50-346 UNIT Davis-Besse Unit 1 DATE COMPLETED BY Bilal Sarsour TELEPHONE (419) 259-5000, Ext. 251 MONTH May, 1980

. DAY AVERAGE DAILY POWER LEVEL '

DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net i 0 g7 0

1 0 i 0

2 18 0 g9 0 3  !

M 0 )

4 .

0 21 O l 5

0 22 0

6 0

, 0 v 0 4 o 8

0 25 0 9

0 26 0 10 0 27 0-l1 0 0 12 28 0

0 13 29 0 0 14 30 0 33 0

15 16 0

4 INSTRUCTIONS On this format. list the aserage daily unit power level in MWe-Net for each day in the ieporting month. Compute to the nearest whole megawatt. ,

(9/77)

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OPERATING DATA REPORT DOCKET NO. 50-346 DATE June 9, 1980 COMPLETED BY Bilal Sarsour TELEPHONE (419) 259-5000

' Ext. 251 OPERATING STATUS Davis-Besse Unit 1 Notes

1. Unit Name:

2. Reporting Period: May, 1980

3. Licensed Thermal Pow er (MWr): 2772

4. Nameplate Rating (Gross MWe): 925

5. Design Electrical Rating (Net MWe): 906

6. Maximum Dependable Capacity (Gross MWe): 934

7. Maximum Dependable Capacity (Net MWe): 890

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

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

10. Reasons For Restrictions. If Any.

1his Month Yr.-to-Date * ~ Cumulative

11. Hours in Reporting Period 744 3,647 24,172

12. Number Of Hours Reactor Was Critical 0 2,078 13,042

13. Reactor Reserve Shutdown Hours 0 0 -28,758

14. Hours Generator On Line 0 2,008.7 11,gg3,g

15. Unit Reserve Shutdown Hours 0 0 1.728

16. Gross Thermal Energy Generated (MWH) 0 -4,687,305 24.886.812 -

17. Gross Electrical Energy Generated (MWH) 0 . _ . 1.583.559 8.307.070

18. Net Electrical Energy Generated (MWH) 0 1,483,787 7,654,365

19. Unit Service Factor 0 55,1 50.4

20. Unit Availability Factor 0 55.1 58.2

21. Unit Capacity Factor (Using MDC Net) .

O _ 45.7 38.3 44.9

22. Unit Capacity Factor (Using DER Net) 0 37.6

23. Unit Forced Outage Rate 0 14.3 25.6

24. Shutdowns Scheduled Over Next 6 Months (Type. Date,and Duration of Each): ,

25. If Shut Down At End Of Report Period. Estimated Date of Startup: August 1, 1980

26. Units In Test Status (Prior to Commercial Operation): Forecast , Achieved INITIAL CRITICALITY INITIA L ELECTRICITY COMMERCIAL OPERATION .

(9/77)

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50-346' .

UNITSHUTDOWNS ANDIOWER REDUCTIONS ~ '

UNIT NAME Davis-Besse 17 nit 1 DATE June 9. 1980 COMPLETED BY Bilal Sarsour REPORT MONTil

• May, 1980 TELEPil0NE (419) 259-5000 Ext. 251 l

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$.E5 Licensee Event Ev, gg h

91 Cause & Corrective Action to No. Date .s s s H

fE $ j g, g Rep.>rt a mV yb Prevent Recurrence 5

4 80 04 7 S 744 C 4 . NA NA NA Unit outage which beg:In on April 7, 1980 was still.in progress through the end o'f May, 1980.

(See operational summary for further details). .

i 2 3* 4 -

1 F: Forced Reason: Method: Exhibit G-Instructions S: Schedu!ed A Equipment Failure (Explain) I Manual for Preparation of Data B-Maintenance of Test 2 Manual Scram. Entry Sheets for Licensee C Refueling 1-Automatic Scram. Event Repor (LER) Fife (NUREG-D Regulatory Restriction 4 Continuation 0161)

E-Operator Training & License Examination 5-Reduction 5

F Administrative '9-Other Exhibit I Same Source G-Oper itional Enror (Explain)

(9/77) Il Other (Explain)

_O_________-_________________ _ _ _ - _ _ _ _ _ _ __ _ _ _ _ _ __ __

OPERATIONAL

SUMMARY

MAY, 1980 5/1/80 - 5/31/80 The unit outage which began on April 7,1980, was still in pro-gress through the end of May, 1980.

The following are the more significant outage activities per-formed during the month of May:

1. Replacement of reactor coolant pump seals for Reactor Coolant Pumps 1-1 and 2-2.

2. Overhaul of many station pumps and valves.

3. Performance of electrical preventive maintenance.

4. The turbine work is approximately 75% completed. The high pressure turbine and main feed pump turbine 1-1 are re-assembled, and the low pressure turbine is being reassembled.

5. The modificacions to the moisture separator reheater are 70%

completed.

6. The majority of the circulating water canal work is completed.

7. The modifications to the fuel transfer system were completed.

Fuel shuffle was commenced and was completed'95 hours0.0011 days <br />0.0264 hours <br />1.570767e-4 weeks <br />3.61475e-5 months <br /> later.

8. A holddown spring on assembly C-3 was discovered to be broken. The subsequent video investigation of fuel assemblies in the core and spent fuel pool revealed definite spring prob-lems.

l Irradiated springs have been sent to Babcock and Wilcox Lynch-burg Research Center for analyses.

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DATE: May, 1980 REFUELINGINFOP5ATION

1. Name of facility: Davis-3 esse nuclear Power Station Unit 1

2. Scheduled date for next refueling shutdown: April, 1980

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

4. Will refueling or resumption of operation thereaf ter require a technical l specification change or other license amendment? If answer is yes, what, l in general, will these be? If answer is no, has the reload fuel design i 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)?

Yes, see attached

5. Scheduled date(s) for submitting proposed licensing action and supporting  !

information. , February, 1980 (revision submittal expected) l I

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.

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

44 - Spent Fuel Assemblies (a) 177 (b) 8 - New 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) 1

9. The projected date of the last refueling that can be discharged to the spent l fuel pool assuming the present licensed capacity. . l Date 1989 (assuming ability to unload the entire core into the spent fuel pool is maintained and the unit goes to an is month reIueling cycle)

REFUELING INFORMATION Continued 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 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) 4 0

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COMPLETED FACILITY CHANGE REQUESTS FCR NO.78-315 3' STEM: Safety Features Actuation System (SFAS)

COMPONENT: SFAS wiring for valve CC1407B CHANGE, TEST, OR EXPERIMENT: On April 26, 1980, FCR 78-315 was closed out. FCR 78-315 was written on June 26, 1978 to document work performed on SFAS Channels 2 and 4 on that date with the verbal approval of Toledo Edison Power Engineering.

The modification involved making wiring changes in two SFAS cabinets such that the white wire on pin "K" of connectors J208 and J408 was moved to pin "C" of each connector which was a spare.

REASON FOR THE CHANCE: The wire running between the two cabinets via pin "K" in connectors J208 and J408 was broken which prevented actuation of valve CC1407B (component cooling water outlet isolation valve from containment). Substituting the spare conductor restored the ability of the SFAS to operate this valve. See Licensee Event Report NP-33-78-88 for further details.

SAFETY EVALUATION: This FCR provides for changing internal cabinet connection wires in channels 2 and 4 of the SFAS in connectors J208 and J408. Presently, these wires show a discontinuity which prevents actuation of valve CC1407B (compon-ent cooling outlet isolation valve from containment) on SFAS logic L422B/L424B.

The wires would be replaced by good wires which would' enable actuation of the above valve as required by the SFAS.

Replacing of defective wires by good wires does not constitute an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO.78-511 SYSTEM: Solid Raduaste '

COMPONENT: Radwaste Solidication System 3 CHANGE, TEST, OR EXPERIMENT: On April 10,1980, the 10CFR50.59 review requested j by FCR 78-511 was completed. The purpose of this review was to comply with the j NRC requirement that such a review be undertaken when a utility utilizes a Chem- {

Nuclear Systems, INC. , mobile radwaste solidification system. System Procedure SP 1104.28, " Solid Radioactive Waste Disposal", was modified to reflect the use of the Chem-Nuclear system.

REASON FOR THE CHANGE: The Chem-Nuclear system is being utilized because the plant's ,

installed solidification system is inoperable. '

SAFETY EVALUATION: Liquid radwaste at Davis-Besse is being solidified using a Chem-Nuclear mobile solidification system. The system is located in the fuel handling area of the auxiliary building on elevation 585', near the existing radwaste drumming station area. This evaluation is based on the skid and liners for the mobile solidi-fication system remaining in their present location in the fuel handling area. The solidified product meets all current NRC and DOT requirements for shipping and burial.

The procedures used to transfer radwaste to the liners and the procedures used for the solidication system (SP 1104.28) have been approved b the Station Review Board.

Liquid radwaste (2003ft / liner is transferred to a 300 ft liner in the fuel handling crea where it will be solidified by a Chem-Nuclear operator utilizing their mobile colidification system. Once liquid has been placed in a liner, the liner is not moved until it has been solidified, thus eliminating the hazard of spilling due to mishandl-ing. Should a liquid filled liner leak, or the waste supply line to the liner develop a leak, the liquid would flow into the drains in the fuel handling area or the radwaste drumming area drains which are all routed to a sump in the auxiliary building and then into the miscellaneous liquid radwaste system. The drains and the sumps are adequate to handle this material and contain it within the auxiliary building. Normal fuel handling ventilation filtration consists of prefilters and HPEA filters. On high radiation in the fuel handling area, the normal ventilation cystem is actuated for the area to ensure radiation limits at the site boundaries are well within the 10CFR Part 100 guidelines. Therefore, the ventilation systems are adequate to handle the solidification activities in the area. An unreviewed safety question is NOT involved with this activity for the following reasons:

1. The probability of occurrence or the consequences of an accident or malfunc-tion of equipment important to safety previously evaluated in the final safety analysis report is NOT increased.

2. A possibility for an accident or malfunction of a different type than any evaluated previously in the final safety analysis report is NOT created.

3. The margin of safety as defined in the bases for any technical specification is P27 reduced.

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