ML20210U886

Monthly Operating Rept for Jan 1987

ML20210U886
Person / Time
Site:	Davis Besse
Issue date:	01/31/1987
From:	Khazrai M, Storz L TOLEDO EDISON CO.
To:	Harold Denton, Haller NRC, Office of Nuclear Reactor Regulation
References
KB87-00019, KB87-19, NUDOCS 8702180625
Download: ML20210U886 (10)

Text

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-346 Davis-Besse Unit 1 WIT DATE February 13, 1987-COMPLETED BY.

M rteza Khazrai-(419) 249-5000, TELEPHONE Ext. 7290 MONTH-January 1987 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net) 50

-411

,7 2

0 18 413 3

186 410 g9 4

397 20 411 5

404 21 410 6

406 412 22 7

409 412 23 8

410 24 410 9

398 25 418 10 386 26 438 11 412 27 469 12 411 492 28 13 410 29 466 14 412 461 30 15 410 3

471

.16 414 INSTRUCTIONS On this format, list the average daily unit power level in MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.

(9/77) 8702180625 870131 DR ADOCK 0500 6

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OPERATING DATA REPORT s

DOCKET NO. '50-346 DATE February 13, 1987-COMPLETED BY Morteza Khazrai TELEPHONE 419-249-5000, Ext.

7290 OPERATING STATUS

1. Unit Name:

Davis-Besse Unit 1 Notes

. 2. Reporting Period:

January 1987

3. Licensed Thermal Power (MWt):

2772

4. Nameplate Rating (Gross MWe):

925-906

5. Design Electrical Rating (Net MWe):

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 Last Report. Give Reasons:

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

1

10. Reasons For Restrictions,if Any:

This Month Yr.-to-Date Cumulative

11. Hours In Reporting Period 744 744 74.'68

12. Number Of Hours Reactor Was Critica]

708.5 708.5 36,763.6

13. Reactor Reserve Shutdown Hours 35.5 35.5 4,660.3 j

14. Hours Generator On.Line 697.6 697.6 35,186.2.

15. Unit Reserve Shutdown Hours

-0.0 0.0 1.732.5

16. Gross Thermal Energy Generated (MWH) 1.009.895 1.009.895-82.436.559

17. Gross Electrical Energy Generated (MWH) 319.592,_.

319.592

_-27.281.979

18. Net Electrical Energy Generated (MWH) 287,599 287,599 25,524,262

19. Unit Service Factor 93.8 93.8 47.2

20. Unit Agailability Factor 93.8 93.8 49.5

21. Unit Capacity Factor (Using MDC Net) 44.9 44.9 39.8

22. Unit Capacity Factor (Using DER Net) 42.7 42.7 37.8

23. Unit Forced Outage Rate 6.2 6.2 36.7

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

i

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

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

Forecast Achiesed 4

l INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIA L OPERATION (4/77 )

_,_,_,_,,__1.

DOCKET No.

50-346 UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Davis-Besse Unit 1 DATE February 13, 1987 COMPLETED BY REPORT MONTH January 1987 TELEPHONE - Morteza Khazrai 419-249-5000, Ext. 7290 o

U E,s "e

EE U Licensee a.

En Cause & Corrective 8,

!! 0 0

l$ U $

Event 3 4l

@ 4l Action to No.

Date

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Report #

$0 0' O Prevent Recurrence 85 M59 o

8 3

86 12 30 F

2.6 A

1 N/A SB PSP Turbine generator was taken off line Con't due to two steam leaks on the steam lines downstream of control valves.

1 87 01 01 F

43.9 A

3 87-001 SJ TRB Main.Feedwater Pump Turbine (MFPT)

1. 1 tripped as a result of high vibra-tion due to bearing failure. The loss of feedwater initiated an Anti-cipatory Reactor Trip System (ARTS)/

Reactor Protection System (RPS) trip which shutdown the unit.

I F: Forced Reason:

Method:

Exhibit G - Instruccions 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) 4

7 OPERATIONAL

SUMMARY

JANUARY 1987

.The turbine generator was synchronized' on line at 0235 hours0.00272 days <br />0.0653 hours <br />3.885582e-4 weeks <br />8.94175e-5 months <br /> on January 1,

- 1987.-.

L The ' reactor -power was maintained at' approximately-40% power.until l'039 hours4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> on January.1, 1987, when a reactor trip occurred.. The trip was the.

result of the loss _'of Main Feedwater Pump Turbine (MFPT) #1. The MFPT #1-

. tripped as a result of high vibration.due to bearing; failure.- MFPT-f2 was-out of. service at the. time. The loss of feedwater initiated an Anticipatory-Reactor. Trip System (ARTS)/ Reactor Protection System (RPS) trip which shutdown the~ unit.

The reactor; criticality was established at 2212 hours0.0256 days <br />0.614 hours <br />0.00366 weeks <br />8.41666e-4 months <br /> on January 2, 1987.

The turbine generator was synchronized on line at 0630 hours0.00729 days <br />0.175 hours <br />0.00104 weeks <br />2.39715e-4 months <br /> on January 3, 1987.

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Reactor power was slowly increased to 50% which'was attained at approximately.

1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> on February 3,1987. - The power increase was limited between 50% and 70% power the rest of the month due to the testing and troubleshooting of MFPT #1.

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s REFUELING INFORMATION

.DATE:, January'1987

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1.

Name of facility: Davis-Besse Unit 1 2.

Scheduled-date for.next' refueling shutdown: February'1988 3.

Scheduled date for' restart f6110 wing refueling: April 1988

- 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~asrociated with the core reload:(Ref. 10-CFR Section 50.59) ? --

Ans: Expect the Reload Report to require standard re' load 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,ae.g.,

n2w or different. fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures.

L Ans: None identified to date.

7.'

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

j (a) 177 (b) 204 Spent Fuel Assemblies 8.

The present licensed spent fuel pool storage capacity an'd. the size of

. any increase in licensed _ storage capacity that has been -requested or is planned, in number 'of fuel assemblies.

I.

Present: 735 Increase size by: 0 (zero)

R 9.

The projected date of the last refueling that can be discharged to the spent 1 fuel pool assuming the present licensed capacity.

.Date:

1996 - assuming ability to unload the entire core into the spent fuel pool is maintained.

BMS/005

w COMPLETED FACILITY CHANGE REQUEST FCR NO.85-027 SYSTEM Auxiliary Steam COMPONENT-Supports No. HBD 87-H9 and HBD 87-H10 CHANGE,' TEST OR EXPIREMENT L

' FCR 84-027 performed the following changes and. modifications; 1.

Replaced the bottom channel ~on s'upport HBD 87-H9 2.

Replaced the top and bottom channels on support HBD 87-H10 3.

Modified the weld design.which had welds around the entire perimeter of the channels to welding only three (3) sides of the channels.

This FCR 84-027 was closed November 4, 1986.

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REASON FOR CHANGE i

FCR 84-027 was initiated due to NCR 83-123 and 83-0068 which identified nonconforming welds on structural members of subject supports.

SAFETY EVALUATION

SUMMARY

l-This FCR 84-027 replaced horizontal structural members on supports HBD 87-H9 i

and HBD 87-H10 and modified the weld design from all around perimeter to three sides only.

The safety. functions of the supports.is.to restrain the piping under a postulated seismic loading. The supports-have been analyzed and found to be acceptable for short term operation (BT-14402). Modification performed by FCR 84-027 will allow long term operations. Based on-the above issue an unresolved safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST-FCR No.85-131 SYSTEM Reactor Coolant System COMPONENT PSV-RCO2 4

-CHANGE, TEST OR EXPERIMENT

^

.FCR 85-131 modified the PORV discharge line by adding a pair of flanges in the PORV discharge line.

This FCR 85-131 was closed October 23, 1986.

REASON FOR CHANGE This modification was incorporated to provide easier' access to the PORV for removal and reinstallation.

SAFETY EVALUATION

SUMMARY

The purpose of this modification improved conditions related to PORV i

maintenance. The installation of these flanges permitted the removal of:

the valve.and spool piece as a unit, eliminating lengthy exposure and maneuvering of the valve around existing piping and equipment. The installation of these flanges does not alter the intended design function of the piping system.

Therefore, the occurrence, consequences, malfunction, probability and possibility of an accident previously evaluated in the Updated Safety Analysis Report (USAR) is not increased as a result of this modification.

The margin of safety as defined in Tech. Spec. has'not been reduced. This change does not involve an unreviewed safety question.

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-COMPLETED' FACILITY CHANGE REQUEST FCR No.85-071

' SYSTEM Main Steam' System COMPONENT.

Various CHANGE, TEST OR EXPERIMENT' FCR 85-071 modifie'd the various pipe supports on the main steam. inlet' lines to the Auxiliary Feedwater Steam Turbines 1-1 and:1-2.

This FCR 85-071 was closed on November 11, 1986.

REASON FOR CHANGE s

This modification was implemented due to Nonconformance Report 85-0019, which identified several supports needing repairs, ranging from replace-ment of expansion anchors to replacement of pipe saddles.

SAFETY EVALUATION

SUMMARY

The function'of a pipe support is to provide support for'all design loads.

The pipe support from the main steam inlet lines to the auxiliary feedwater-pumps was analyzed by Bechtel with one bolt inactive. This was found to.

meet all short and long term conditions. However, the base plate for this support was found to be damaged and required rework per NCR 85-0071. -The damaged anchor bolt was replaced and this restored the support to the original design, i

This modification does not create a reduction in the margin of safety as defined in the basis for any Technical Specification. Therefore, an-unreviewed safety question does not exist.

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' COMPLETED FACILITY CHANGE REQUEST 1

1

- FCR No.84-193 SYSTEM Moisture Separator and Reheater System

- COMPONENT HSR Safety Valve Vents CHANGE, TEST OR EXPERIMENT FCR 84-0193 allowed the installation of a threaded reheat safety valve

. vent line to the condenser.

-This FCR'84-193 was closed February 21, 1986.

REASON FOR CHANGE The safety valve vent lines were hard piped to the condenser during valve maintenance these lines needed to be cut.

Installing unions facilitates maintenance work on the safety valves.

SAFETY EVALUATION

SUMMARY

The subject vent lines are used to vent the area above the bellows in the case of a bellows leak in these valves. These lines serve no nuclear safety related function. Addition of unions does not effect any nuclear safety related items. This change does not constitute an unreviewed safety question.

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TOLEDO h EDISON EDISON PLAZA 300 MADISON AVENUE TOLEDO OHIO 43652 February 13, 1987 KB87-00019 File: RR 2 (P-6-87-01)

Docket No. 50-346 License.No. NPF-3 Mr. Harold Denton, Director Office of Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Haller:

Monthly Operating Report, January 1987 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 January 1987.

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

Yours truly, F

orz T

P ant Manager Davis-Besse Nuclear Power Station LFS/MK/ljk Enclosures cc:

Mr. A. Bert Davis, w/1 Regional Administrator, Region III Mr. James M. Taylor, Director, w/2 Office of Inspection and Enforcement Mr. Paul Byron, w/1 NRC Resident Inspector Nuclear Records Management, Stop 3220 LJK/002 x'N'k

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