ML19296D897

Monthly Operating Rept for Feb 1980

ML19296D897
Person / Time
Site:	Davis Besse
Issue date:	03/07/1980
From:	Caba E, Sarsour B TOLEDO EDISON CO.
To:
Shared Package
ML19296D894	List: ML19296D893 ML19296D897
References
NUDOCS 8003130467
Download: ML19296D897 (8)

Text

a

• AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 Davis-Besse Unit 1 UNIT F! arch 7, 1980 DATE COMPLETED BY Erdal Caba/Bilal 419-259-5000yrsour TELEPJf0NE Ext. 236/251 February, 1980 MONTil

- DAY AVERAGE DAILY POWER LEVEL ' DAY AVER AGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 886 37 879 g

884 gg 816 2

884 39 822 3 _

881 20 815 4 .

559 21 821 5

0 818 6 22 0 23 820 7

0 24 819 8

0 25 817 9

145 816 10 26 687 27 819 33 844 28 821 12 876 29 805 13 14 879 30 878 15 16 879 INSTRUCTIONS On this format.hst the average daily unit power leselin MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.

(9[77 )

8002;.30V6f

OPERATING DATA REPORT 50-346 DOCKET NO.

March 7, 1980 DATE CONIPLETED BY Erdal Caba/Bilal TELEPliONE 419-259- Sarsou 5000, Ext. 236/251 OPERATING STATUS Notes

1. Unit Name:

Davis-Besse Unit 1 February, 1980

2. Reporting Feriod:

2772

3. Licensed Thermal Power (1 twt p:

925

4. Nameplate Rating (Gross 51We):

906

5. Design Electrical Ratine(Net stWe):

934

6. h!aximum Dependable Capacity (Gross 51We):

7. Sf aximum Dependable Capacity (Net $1We):

890

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

Modified based on capacity test

9. Power Lese! To Which Restricted.lf Any (Net 31We):

10. Reasons For Restrictions,if Any:

This 5fonth Yr.-to-Dat e Cumulatise 696 1,440 21,965

11. Ilours in Reporting Period 12,202.9 614.6 1,238.7

12. Number Of flours Reactor Was Critical 0 0 2,875.8

13. Reactor Reserve Shutdown flours 1,180.5 11,055.3 587.4

14. Ilours Generator On-Line 0 0 1,728.2

15. Unit Resene Shutdown flours 3,065,778 21,265.285 1,495,351

16. Gross Thermal Energy Generated (51WH) 7,755,960 505,400_ _ _, 1,032,449

17. Gross Electrical Energy Generated (MWH) 7,143,695 476,480 973,117

18. Net Electrical Energy Generated (Mbfl) 84.4 32.0 _51.8

19. Unit Service Factor 82.0 60.5 84.4

20. Unit Asailability Factor 39.7 76.9 75.9

21. Unit Capacity Factor tUsing 31DC Net) 39.0 75.6 74.6

22. Unit Capacity Factor (Using DER Net) 26.7

23. Unit Forced Outage Rate 15.6 18.0

24. Shutdowns Scheduled Over Next 6 Months (Type.Date,and Duration of Each1:

Refueling Outage April, 1980 12 weeks

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

26. Units in Test Status tPrior to Commercial Operationi: Forecast Achiesed INITIAL CRITICALITY .

INITIAL ELECTRICITY COMMERCIA L OPER ATION (9/77)

t i

I 0-346 __

DOCKETNO. WVis-uesse Unit 1 [

UNIT SlIUTDOWNS ANDIOWE!t REDUCTIONS ' UNIT NAME March /, 1980 l DATE COMPLETED !!Y Erdal Caba/Bilal Sarsour lt TELEPI:0NE 419-259-5000 E_x t . 236/

REPORT MONT!! ' February 1980 251  !

1 C'

w $ Cause & Corrective ge,

,$ g 3 $Yl Licensee Ee, As. tion to Date o- M2 4 jE5 Event u9 93 Prevent Recurrence No.

S $ 5dc2 Report a <# 0 8O

$8 <.

v ~

d .

NA NA NA Reactor trip on high pressure caused 1 80 02 05 F 108.6 A 3

- apparently by a false indication of turbine overspeed. See Operational -

Summary for further details. ,,

i D

4 3

I 2 Method: Exhibit G Instructions F: Forced Reason: for Preparation of Data l-Man ual S: Schedu!cd A Equipment Failure (Explain) 2 M.tnual Scram.

Entry Sheets for Licensee

~

B Maintenance of Test Event Report (LER) File (NUREG-3 Autonutic Scram.

C Refucting 4-Orher (Exp! in) 01611 D Regulatory Restriction

!? Operator Training & License Examination 5

• F Adininistrative Exhibit I - Same Source G-Operational Einor (Explain) 11 Other (laptain)

(9/77)

OPERATIONAL SUMMAKi FEBRUARY, 1980 2/1/80 - 2/4/80 Reactor power was maintained at 100% of full power with the genera-tor gross load at 920 1 10 MWe. On February 2, 1980, condensate chemistry indicated that a tube leak had occurred in the high pres-sure condenser.

2/5/80 The Reactor Coolant Pump (RCP) 1-1 first stage seal experienced a par-tial destaging increasing second seal cavity pressure to 1900 psi.

At 1539:57 hours on February 5, 1980, the electrohydraulic control (EHC) system closed the high pressure turbine control valves and the combined intermediate valves apparently in response to a false indication of turbine overspeed. At 1540:01 hours, the Reactor Protection System (FPS) tripped the reactor on high Reactor Coolant System (RCS) pressure. Station personnel, with concurrence from General Electric field engineers replaced all the circuit boards which could have been the source of the faulty curbine speed signal.

During the shutcown, the condenser tube leaks were located and the tubes plugged.

2/9/80 The reactor was critical at 0110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br />.

2/10/80 - 2/12/80 The turbine generator was synchronized on line at 0145 hours0.00168 days <br />0.0403 hours <br />2.397487e-4 weeks <br />5.51725e-5 months <br />.

Reactor power was slowly increased and attained 100 percent full power on February 12, 1980 with generator gross load at 920 i 10 MWe.

2/13/80 - 2/18/80 Reactor power was maintained at 100 percent until the morning of February 18, 1980 when the #4 turbine control valve partially closed. This caused main steam header pressure to increase which in turn opened #3 control valve fully in an attempt to lower the header pressure. Because header pressure remained excessive, reac-tor power was reduced until #3 control valve gained control of header pressure (approximately 88 percent) and the #4 control valve was fully closed. Control valve #4 was then electrically disabled.

Complete evaluation of this problem will take place during the refueling outage. Reactor power was increased to approximately 92% which corresponds to the maximum power at which #3 control valve will reliably control header pressure.

2/19/80 - 2/29/80 Reactor power was maintained at 92% until the evening of February 29, 1980. Reactor power was then decreased to approximately 72%

to stretch allowable core burnup to the refueling date and also to enable a safe shutdown of RCP i-1 should the third cavity seal pressure increase.

February, 1980

. DATE:

JtEFUELING IMF0biA".ICT .

Name of facility: Davis-Besse Nuclear Power Station Unit 1 1.

April, 1980

2. Scheduled date for next refueling shutdown:

following refueling: _ July, 1980

3. Scheduled date for restart

4. Will refueling or resumption of operation thereafIf ter require answcr a technical is yes, what, specification change or other license amendment?If answer is no, has the reload in general, will these be? Safety Review Ccanittee and core configuratica been reviewed by your Plant ih to determine whether any unreviewed safety questions are associated w t the core reload (Ref.10 CFR Section 50.59)?

Yes, see attached

5. Scheduled date(s) for submitting proposed licensing action and supporting February, 1980 information.

6. Important licensing considerations associated with refueling, e.g. , new or different fuel design or supplier, unreviewed design or perf ormance analysis methods, significant changes in fuel design, new operating procedures. .

(a) in the core and (b) in the spent fuel

7. The nenber of fuel assemblics storage pool. ,

0 (zero) 177 _ (b) -

(a)

8. The present licensed spent fuel pool storage capacity in number of fuel assemblies.

Increase size by 0 (zero)

Present 735 9.

The projected date of the last licensed refuelingcapacity.

that can be discharged to the spent .

fuel pool assuming the present 1989 (assuming ability to unload the entire core into.the spent fuel Date pool is maintained anu tiie unit goes to an lo monta retueling cycle;

REFUELING I!; FORMATIO : Centinued Page 2 of 2

4. The following icchnical 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 Uater Sources (and Basts) 3.2.4 - Quadrant Power Tilt (and Bases) 3.2.5 - D:1B Parameters (and Bases)

COMPLETED FACILITY CllANCE REQUESTS FCR NO: 78-009 SYSTEM: Station Service Building No. 2 COMPONENT: N/A

_ CHANCE, TEST, OR EXPERIMENT: On February 6,1980 work was completed on Supplement 4 of FCR 78-009 completing the physical work for this FCR. This FCR was written to provide heat, ventilation and sufficient electrical service to support the use of Service Building No. 2 as a welding fabrication and training area.

REASON FOR FCR: The Davis-Besse Maintenance Department requires additional space to support plant maintenance programs. Since there is no space availabic for expan-sion within the Maintenance Shop, Station Service Building No. 2 was adapted via this FCR to provide the required space for maintenance activities. Supplement 4 of this FCR, the only one of four supplements which is nuclear safety related, is nuclear safety related by virtue of the fact that Post-Inspection Construction Authorization (PICA) review was required to ensure that the installation of the required electric.nl service would not result in the creation of any environment adverse to nuclear safety related equipment.

SAFETY EVALUATION: Portions of the work required to impicment this FCR supplement are nuclear safety related due to PICA requirements. Installation in accordance with PICA requ,.rements will ensure that no adverse environment is created fo-nuclear saf ety related equipment due to the addition of this equipment. ,

il

, COMPLETED FACILITY CHANGE REQUESTS FCR NO: 78-175 SYSTEM: Process and Area Radiation Monitoring COMPONENT: Instrument voltage to current (E/I) converters, RIS 105, 106, 110, 201, 202, 203, 205, 207, 305, 306, 310, 401, 601, 602, 603, 605, 607, 608, 701, 702, 703, .

705, 707, 708, 801, and 802  !

I f

l CHANCE, TEST, OR EXPERIMENT: On May 14, 1979, tmplementation of FCR 78-175 was com-  !

pleted. As requested by this FCR, all wires connected to the above converters were f disconnected and marked as spares. This change was made with the guidance of the unit architect / engineer, Bechtel Company.

REASON FUR THE FCR: The outputs of the above liste. ' converters were not connected and so there was no need for the inputs to be connected. The FCR properly tags the unused wires and allows the above listed converters to be used as spares.

SAFETY EVALUATION: As stated in the FCR, a number of E/I converters have no outputs and therefore do not need to be hooked up. Some of these have been used for spares, j as described in the FCR. Removing the E/I converters, lifting the associated wires, taping the wire ends, and 1 caving the wires in place (for future use) will not adversely affect the safety of the plant.

.