ML20207D591

Monthly Operating Rept for Aug 1986

ML20207D591
Person / Time
Site:	Davis Besse
Issue date:	08/31/1986
From:	Khazrai M, Storz L TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
KB86-0697, KB86-697, NUDOCS 8612310098
Download: ML20207D591 (13)

Text

h AVERAGE DAILY UNIT POWER LEVEL 50-346 DOCKET NO.

UNIT Davis-Besse Unit 1 DATE September 10, 1986 COMPLETED BY Morteza Khazrai TELEPHONE (419) 249-5000, Ext. 7290 MONTH August 1986 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 1 0

0 g7 2

0 gg 0

3 0

0 g,

4 0

0 20 5

0 21 0

0 0

6 22 7

0 n

0 8

0 24 0

9 0

25 o

10 0

0 26 1I O

27 0

12 0

0 28 13 0

0 29 14 0

0 30 IS 0

33 0

16 0

INSTRUCTIONS On this format, list the average daily unit power levelin MWe-Net for each day in the reporting month. Compute to the nearest *

• ole megawatt.

(9/77 )

_TEB4 8612310098 860831 ij PDR ADOCK 05000346 i

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d OPERATING DATA REPORT '

DOCKET NO. ' 50-346 DATE September' 10, 1986 COMPLETED BY Morteza Khazrai TELEPHONE (419)249-5000, Ext.

7290 OPERATING STATUS

1. Unit Name:

Davis-Besse Unit 1 Notes

. 2. Reporting Period:

Aunust 1986

3. Licensed Thermal Power (MWs):

2772

4. Nameplate Ratir.g (Gross MWe):

925

5. Design Electrical Rating (Net MWe):

906

6. Maximum Dependable Capacity (Gross MWe):

904

7. Maximum Dependable Capacity (Net MWe):

860

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

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

10. Reasons For Restrictions,If Any:

This Month Yr.-to.Date Cumulative

11. Hours In Reporting Period 744 5,831 70,896

12. Number Of Hours Reactor Was Critical 0.0 0.0 35,877.I

13. Reactor Reserve Shutdown Hours 0.0 0.0 4,058.8 l

14. Hours Generator On.Line 0.0 0.0 34,371.8-

15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5

16. Gross Thermal Energy Generated (MWH)

O.0 0.0 81.297.600

17. Gross Electrical EnerEy Generated (MWH) 0.,0,,

0.0 26,933,622

18. Net Electrical Energy Generated (MWH) 0.0 0.0 25.233,177

19. Unit Service Factor 0.0 0.0 48.5

20. Unit Availability Factor 0.0 0.0 50.9

21. Unit Capacity Factor (Using MDC Net) 0.0 0.0 41.4

22. Unit Capacity Factor (Using DER Net) 0.0 0.0 39.3

23. Unit Forced Outage Rate 1on n 100.0 33.7

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:

November 10, 1986

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

Forecast Achiesed INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCI AL OPER ATION i

i (4/77 )

4

~

DOCKET NO.

50-346-UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Davis-Besse Unit 1 DATE September 10.-1986 COMPLETED BY Morteza Khazrai REPORT MONTH August 1986 TELEPHONE (419) 249-5000. Ext. 7290 1

4 g

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Licensee s -e g

Cause & Corrective Event g

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y3y g7 gg Action to No.

Date e

uo e

u :s Report i

>, o eo Prevent Recurrence 55 fis e

e a

7 85 06 09 F

744 A

4 LER 85-013 JK SC The unit remained shutdown follow-Contd ing the reactor trip on June 9, 1985.

See Operational Summary for-further details.

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l I F: Forced Reason:

Method:

Exhibit G - Instructions j

S: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparation of Data B-Maintenance or Test 2-Manual Scram Entry Sheets for Licensee C-Refueling 3-Automatic Scram Event Report (LER) File D-Regulatory Restriction 4-Continuation from (NUREG-0161)

E-Operator Training & License Examination Previous Month F-Administrative 5-Load Reduction 5

G-Operational Error (Explain) 9-Other (Explain)

Exhibit I - Same Source (9/77)

H-Other (Explain) l

OPERATIONAL

SUMMARY

AUGUST 1986 The unit remained shutdown the entire month of August following the reactor trip on June 9, 1985. Corrective actions and system upgrades continue.

Below are some of the major activities performed during this month:

1)

Continued testing as part of the System Review and Test Program.

2)

Continued Motor Operated Valves Analysis Test (MOVATS) activities.

3)

Continued Raychem investigation and followup corrective actions.

4)

Completed integrated Safety Features Actuation System (SFAS) testing.

5)

Work on Reactor Coolant Pumps (RCPs) 1-1, 1-2, and 1-3 is in progress.

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4 REFUELING INFORMATION DATE: August 1986 1.

Name of facility: Davis-Besse Unit 1 2.

Scheduled date for next refueling shutdown: October, 1987 3.

Scheduled date for restart following refueling: December, 1987

4. -

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

Ans: Expect the Reload Report to require standard reload fuel-design Technical Specification changes (3/4.1 Reactivity Control Systems and 3/4.2 Power Distribution Limits).

5.

Scheduled date(s) for submitting proposed licensing action and supporting information: Summer, 1987 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.

Ans: None identified to date.

7.

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

(a) 177 (b) 204 - Spent 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) 9.

The projected date of the last refueling that can be discharged to

~

the spent fuel pool assuming the present licensed capacity.

Date:

1995 - assuming ability to unload the entire core into the 4

spent fuel pool is maintained.

9 BMS/005

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COMPLETED FACILITY CHANGE REQUEST FCR NO.80-058 SYSTEM:

Reactor Coolant System COMPONENTS:

Various CHANGE, TEST OR EXPERIMENT This FCR performed the following modifications:

1.

Revised the pressurizer' heaters controls (drawing E52B, Sht. 42E, Rev.'6 and B-H-D 8034805C) so that a loss of NNI power would deenergize the heaters.

2.

Revised the pressurizer spray line valve drawing E52B, Sht. 59, Rev. 1.

3.

Revised the power operated relief valve RC2A drawing 52B Sht. 13, Rev. 6.

4.

Changed the pressurizer power operated relief valve control switch so what when turned and pulled the circuit energizes and locks in until turned off.

5.

Changed the Integrated Control System (ICS) and NNI computer points and anuunciator windows, per the following:

a.

Q533 ICS fuse blown )

I window, 2 comp. pts.

Q718 NNI fuse blown )

b.

Q526 ICS or NNI 24 VDC )

power supply TROUBLE

)

No window, I comp. pt.

alarm

)

c.

Q525 ICS (X) 24 VDC bus trip )

I window, 2 comp. pts, Q527 ICS (Y) 24 VDC bus trip )

d.

Q715 NNI (x) 24VDC bus trip )

I window, 2 comp. pts.

Q716 NNI (y) 24VDC bus trip )

6.

Added an additional power supply monitor to the NNI and ICS systems.

This FCR was closed June 23, 1986.

LREASON FOR CHANGE As a result of the Crystal River loss of NNI incident, Toledo Edison had been committed to modify the control circuits of selected critical equipment. The review

1..

0 FCR-80-058' Page 2.of 2 of loss of power to the non-nuclear instrumentation indicates that unde-sirable fail positions result with certain critical equipment. Of major concerns were the following:

1.

The power operated relief valve would open and remain open.

2.

-The pressurizer spray valve would open and remain open.

3.

The pressurizer heaters would energize and remain energized.

This change ensures a loss of power will result in both the power operated relief valve and the pressurizer spray valve to close and the pressurizer heaters to deenergize. Manual control is then available.

Another improvement was the addition of Station computer alarms to indicate when loss of power occurred to the Non-Nuclear Instrumentation or the Integrated Control System.-

SAFETY EVALUATION

SUMMARY

This FCR involved modification on the following nuclear safety related items:

1.

Pressurizer spray line valve 2.

Pressurizer heaters This FCR involved modification on the following non-nuclear safety items:

1.

Power operated relief valve 2.

NNI cabinet 3.

ICS cabinet This FCR installed a power supply monitor and an auxiliary relay module in the NNI (X) panel. The power monitor was installed and modified so as to have a tripping time approximately 22 VDC decreasing of approximately 3M sec. Also an existing switch at panel C5705 was replaced. New cable was also pulled in existing raceways. The~ separation of non-nuclear and nuclear circuits is provided through the 27 X relay.

Based on the above information, no unreviewed safety question exists, mj b/70

. w COMPLETED FACILITY CHANGE REQUEST.

FCR N0;81-189 SYSTEM:

Non-Nuclear Instrumenation (NNI)

COMPONENTS:

-NNI. Power Supplies-CHANGE. TEST OR EXPERIMENT This FCR modified. Control Room panel C5722 by adding four-(4) indicating lights.to monitor NNI power.

Lens colors are as follows:

Y-AC~and Y-DC - Amber X-AC and X-DC - Blue This FCR was closed June 11, 1986.

REASON FOR CHANGE This FCR modification will help the Control Room operators determine a loss of NNI power has occurred.

SAFETY EVALUATION

SUMMARY

The addition of these indicating lights does not adversely. affect the function of the panel.

It does enhance operation by providing indication to the operators of a loss of NNI power.

No adverse environment was created by the drilling or minor griding required to install the indicating lights.

Installation was in accordance with "Q" DCNs and proper control of the core drill cutouts report preclude those portions of the work from creating any new adverse environment.

This change does not constitute an unreviewed safety question.

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.y-COMPLETED FACILITY CHANGE REQUEST FCR NO.

'84-115 SYSTEM:

Service' Water Syst'em COMPONENTS:

Check and Gate Valves CHANGE, TEST OR EXPERIMENT

.This FCR repositioned the following valves:

1.

Repositioned valves SW-86,.SW-94 and SW-102 on Service Water Train #1.

2.

' Repositioned valves SW-108 and SW-116 on Service Water Train #2 associated with its Service Water Pump 1-2 from locked open to locked closed.

This:FCR was closed February _ 21, 1986.

REASON FOR CHANGE To eliminate' leak test requirements for check valves in the supply line to Emergency Core Ccoling System (ECCS) pump room coolers,_ decay heat room cooler and Hydrogen Dilution System blowers.

In addition, this change is required to maintain the redundancy of service water system for single failure criteria.

SAFETY EVALUATION

SUMMARY

This FCR ensures that if Service Water Train #1 which provides flow to ECCS Pump Room Coolers 1-4 and 1-5 in Room 105 fails, the redundant Train #2 will provide flow to the ECCS Pump Room Coolers 1-1 and 1-2 in

-Room 115 and the Decay Heat Room Cooler 1-3 in Room 113.

If Service Water Train #2, which provides flow to the ECCS Pump Room Coolers 1-1 and 1-2 in Room 115 and Decay Heat Coolers 1-3 in Room 113

. fails, the redundant Train #1 will provide flow to ECCS Room Coolers 1-4 and.1-5 in Room 105 and will have adequate capacity to remove the heat loads in Rooms 105 and 113. As a result of the above, there is no unre-viewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO.84-192 SYSTEM:

Davis-Besse License-NPF-3 COMPONENTS:

N/A

-CHANGE, TEST OR EXPERIMENT This FCR added license condition 2.C(3)(t) for the Startup Feedwater Pump (SUFP) by incorporating the following:

1.

Dedicated operator during SUFP operation.

2.

Isolate water sources when not in operation a)

Suction (DST) b)

Discharge (FW) c)

Pump cooling (TPCW) 3.

Install new SUFP before commencing Cycle 6.

This FCR was closed March 4, 1986.

REASON FOR CHANGE This license' condition was requested by NRC to allow startup and operation utilizing the current SUFP configuration.

SAFETY EVALUATION

SUMMARY

The SUFP System itself performs no safety function.

It is, however, used as a backup to the Main and Auxiliary Feedwater Systems for supplying water to the steam generators in case of the total loss of these two systems.

The main concern with the location of the SUFP is the potential for pipe whip and jet impingement in Room 238 and flooding and high temperature in Room 237 or'238. These concerns are only realized during startup and shutdown of the' reactor. An operator shall be positioned at the Auxiliary Feedwater room area when the SUFP is operating in Modes 1 through 3.

Upon indication of a leak, the operator will trip the SUFP or contact the Control Room to trip the SUFP. The operator would then close the appropriate valves.

Based on the above actions, it has been deemed an unreviewed safety question does not exist.

s

.4 COMPLETED FACILITY CHANGE REQUEST FCR NO.

85-0028 SYSTEM:

Demineralized' Backwash Recirculation Tank (BWRT)

-COMPONENTS:

T-110 CHANGE, TEST OR EXPERIMENT This FCR replaced a 90 elbow with a tee and installed a lateral cleanout

-in 3" HBD-115.

3" HBD-115 is the discharge line from the Backwash Recircu-lation Tank to the Settling Basin.

This FCR was closed July 29, 1986.

REASON FOR CHANGE Discharge line 3" HBD-115 periodically became clogged. The modification facilitated cleaning of the clogged line.

SAFETY EVALUATION

SUMMARY

FCR 85-0028 Rev. A replaced a 90* elbow with a tee which has a capped end.

It also replaced a straight.section of pipe with a lateral which has a flange. Both changes were on 3" HBD-115 drain line from the Backwash Recirculation Tank to the Settling Basin. The reason for the changes was because this line periodically clogged, and these changes allowed access of cleaning equipment without cutting and rewelding the line.

This does not create a possibility for an accident different than any evaluated previously in the USAR, or reduce the margin of safety as defined in the bases for any Technical Specifications. Therefore, an unreviewed safety question'does not exist.

COMPLETED FACILITY CHANGE REQUEST FCR NO.85-078 SYSTEM:

D. C. Distribution COMPONENTS:

Station Batteries CHANGE, TEST OR EXPERIMEhT This FCR adjusted the gap between the end stringer and battery cells so that it measured between 0" and 1/4".

The gap is maintained by placing a tempered masonite spacer between the end stringer and battery cells.

The FCR also adjusted the gap between the side stringer and battery cells so that it is measured 0" to 3/8".

The gap is achieved by adding spacers or flat washers between the battery rack rails and the vertical support.

This FCR was closed February 12, 1986.

REASON FOR CHANGE These adjustments were made in response to a letter dated March 27, 1985 to Toledo Edison from GNB Batteries Inc. The letter was in response to Toledo Edison letter dated March 19, 1985 which had a safety concern on seismic adequacy of batteries.

GNB Batteries' recommendation is a result of their seismic testing conducted at Wyle Laboratories.

SAFETY EVALUATION

SUMMARY

The safety function of the battery racks is to support Station batteries before and after a seismic event. Our Engineering evaluation indicated the original condition would not create an adverse effect. Due to seismic testing conducted at the Wylie Laboratories, Davis-Besse Engineering recommended the above modifications. These changes do not create any adverse environment and do not constitute an unreviewed safety question.

TOLEDO

%mm EDISON September 10, 1986 Log No. KB86-0697 File: RR 2 (P-6-86-08)

Docket No. 50-346 License No. NPF-3 Mr. Norman Haller, Director Office of Management and Program Analysis U. S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Haller:

Monthly Operating Report, August 1986 Davis-Besse Nuclear Power Station Unit 1 Enclosed are ten copies of the Monthly Operating Report for Davis-Besse Nuclear Power Station Unit 1 for the month of August 1986.

If you have any questions, please feel free to contact Morteza Khazrai at (419) 249-5000, Extension 7290.

Yours truly, Louis F. Storz Plant Manager Davis-Besse Nuclear Power Station LFS/MK/ljk Enclosures cc:

Mr. James G. Keppler, w/1 Regional Administrator, Region III Mr. James M. Taylor, Director, w/2 Office of Inspection and Enforcement

,][6f d' Mr. Paul Byron, w/1

l NRC Resident Inspector LJK/002 THE TOLEDO EOISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO. OHIO 43652

,-