ML20206R987

Monthly Operating Rept for Mar 1987

ML20206R987
Person / Time
Site:	Davis Besse
Issue date:	03/31/1987
From:	Khazrai M, Storz L TOLEDO EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
KB87-00050, KB87-50, NUDOCS 8704220309
Download: ML20206R987 (11)

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-346 UNIT Davis-Besse Unit 1 April 14, 1987 DATE COMPLETED BY Morteza Khazrai TELEPHONE (419) 249-5000, Ext. 7290 March 1987 DAY AVERAGE DAILY POWER LEVEL- DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 482 0 37 2 482 83 Ig 3 486 0 19 20 654 5 482 878 21

' 9 22 882 7 484 880 23 8 495 880 24

' 9' 25 875 10 655 880' 26 il 02 892.

27 12 726 872 28 13 254 880 29 14 0 874 30 15 0 887 3

16 0 INSTRUCTIONS On this format, list the average daily unit power level in MWe-Net for each day in the reporting month. Comput the nearest whole megawatt.

19!77) 8704220309 870331 PDR ADOCK 05000346

e OPERATING DATA REPORT DOCKET NO. 50-346 DATE APr11 14. 1987 COMPLETED BY c1 rteza anazrai TELEPHONE 419-249-3000, Ext. 7290 OPERATING STATUS Davis-Ben.se Notes s

1. Unit Name:

2. Reporting Penod March 1987

3. Licensed Thermal Power (MWt):

2772

4. Nameplate Rating (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 Last 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 2,160 76,056

12. Number Of Hours Reactor Was Critical 635.6 2,016.1 38,071.2

13. Reactor Reserve Shutdown Hours 108.4 143.9 4,768.7

14. Hours Generator On-Line 599.6 1,969.2- 36,457.8

15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5

16. Gross Thermal Energy Generated (MWH) 1,347,857 3,441,493 84,868,157 .

17. Gross Electrical Energy Generated (MWH) 441.338 ,_, 1.113,521 28.075.908  ;

18. Net Electrical Energy Generated (MWH) 410.057 1.082.240 28.044.627

19. Unit Service Factor 80.6 91.2 47.9

20. Unit Availability Factor 80.6 91.2 50.2

21. Unit Capacity Factor (Using MDC Net) 64.1 58.3 42.9

22. Unit Capacity Factor (Using DER Net) 60.8 55.3 40.7

23. Unit Forced Outage Rate 19.4 8.8 36

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:

26. Units In Test Status (Prior to Commercial Operation): Forecast Achieved INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION (9/77)

_ _ - - -. _ - - - _ _ _ ~ _ . . - . - -. - - . _ _ _ . .

- _ = - - . - . . -- - . . -

DOCKET NO. 50-346 .

UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Davia-Besne Unit 1 ,

DATE April 14, 1957 COMPLETED BY Morteza Khazrai .

REPORT MONTH March 1987 TELEPHONE 4190249-5000, Ext. 7290 i

w O M Em "e EU Licensee s .e $ in Cause & Corrective E. U0 @ lUE Event 3$ $$ Action to No. Date N O5

$ 6$" Report #

$0 @*O Prevent Recurrence 4

^ a5 259 O e

2 87 03 13 F 118.7 H 3 006 SJ SHV The reactor was tripped by the Reactor Protection System (RPS) as a result of high Reactor Coolant System (RCS). pressure caused by the isolation of feedwater flow to Steam Generator (SG) #2. The turbine generator automatically tripped as a result of the reactor trip. See Operational Summary for further details.

3 87 03.18 F 25.7 A 1 N/A TA TRB The turbine was manually tripped due to the turbine steam leak. See

. Operational Summary for further details.

i I 4 F: Forced Reason: Method: Exhibit G - Instructions 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 G-Operational Error (Explain) 5 9-Other (Explain) Exhibit I - Same Source (9/77) H-Other (Explain) 4 1

REFUELING INFORMATION DATE: March 1987-i.

1. Name of facility: Davis-Besse' Unit 1

2. Scheduled date for next refueling shutdown: February - 1988

3. Scheduled date for restart following 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 associated with the core reload (Ref. 10 CFR Section i 50.59)?.

1 Ans: -Expect the Reload Report to require standard reload fuel design Technical Specification changes (3/4.1 Reactivity Control-Systems and 3/4.2 4

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.

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

fuel storage pool, fa) 177 (b) 204 - Spent Fuel Assemblies

8. The present licensed spent fuel pool storage capacity and the size of y any increase in licensed storage capacity that has been requested or j is planned, in number of fuel assemblies.

t Present: 735 Increase size by: 0 (zero) i

9. The projected date of the last refueling that can be discharged to the spent 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 2

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OPERATIONAL

SUMMARY

MARCH 1987 Reactor power was maintained at approximately 58 percent power until 1730 hours0.02 days <br />0.481 hours <br />0.00286 weeks <br />6.58265e-4 months <br /> on March 9, 1987, when reactor power was slowly increased to approxi-mately 92 percent power at 2230 hours0.0258 days <br />0.619 hours <br />0.00369 weeks <br />8.48515e-4 months <br /> on March 10,-1987.

Reactor power was maintained at approximately 92 percent of full power until Control Rod Group 7 Rod 10 dropped into the core due to a blown fuse. A plant runback was initiated which reduced plant power to approxi-mately 51 percent. The fuse was replaced and Rod 7-10 repulled. Reactor power was restored to 92 percent power.

Reactor power was maintained at approximately 92 percent power until 0822 hours0.00951 days <br />0.228 hours <br />0.00136 weeks <br />3.12771e-4 months <br /> on March 13, 1987, when the reactor tripped as a result of high Reactor Coolant System (RCS) pressure following a loss of feedwater flow to Steam Generator (SG) #2. This transient was initiated when Feedwater Valve 601 was inadvertently closed. The closure resulted from a construc-tion worker bumping the local control switch which sent a signal to the motor operator to close the valve.

The reactor was critical at 2047 hours0.0237 days <br />0.569 hours <br />0.00338 weeks <br />7.788835e-4 months <br /> on March 17, 1987. The turbine generator was synchronized on line at 0659 hours0.00763 days <br />0.183 hours <br />0.00109 weeks <br />2.507495e-4 months <br /> on March 18, 1987.

The turbine was manually tripped at 1923 hours0.0223 days <br />0.534 hours <br />0.00318 weeks <br />7.317015e-4 months <br /> on March 18, 1987 due to the turbine steam leak, caused by the failure of a capped pressure tap.

The turbine-generator was synchronized on line at 2107 hours0.0244 days <br />0.585 hours <br />0.00348 weeks <br />8.017135e-4 months <br /> on March 19, 1987, following repair of a steam leak.

The reactor power was increased steadily and attained 100 percent power at 2204 hours0.0255 days <br />0.612 hours <br />0.00364 weeks <br />8.38622e-4 months <br /> on March 20, 1987, and maintained at this level for the rest of the month.

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COMPLETED FACILITY CHANGE REQUEST FCR NO. 84-0202 SYSTEM RCS High Point Vents COMPONENT High Point Vent Switches HIS-4608A, HIS-4608B, HIS-4610A-and HIS-4610B CHANGE, TEST OR EXPERIMENT FCR 84-0202 provides covers for the high point vent switches HIS-4608A, HIS-4608B, HIS-4610A and HIS-4610B.

I This FCR was closed September 30, 1985 REASON FOR CHANGE l FCR 84-0202 was incorporated to add a guard to the high point vent _ switch-l es located on the post accident indicating panel to prevent their acciden-tal operation.

SAFETY EVALUATION

SUMMARY

This FCR 84-0202 will minimize accidental operation of the high point vent switches without affecting their safety function. FCR 84-0202 will not increase the probability of occurrence or the consequences or an accident

, or talfunction of safety related equipment as evaluated in the USAR. Nor will implementation of FCR 84-0202 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 specification.

Therefore an unreviewed safety question does not exist.

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, COMPLETED FACILITY CHANGE REQUEST ,

FCR No.79-410 SYSTEM Reactor Coolant System COMPONENT Pilot Operated and Safety Relief Valves RC2A. RC13A and RC13B CHANCE, TEST OR-EXPERIMENT This FCR 79-410 provided a positive,' redundant and safety grade indication and alarm of flow through valves RC2A, RC13A and RC13B.

This FCR 79-410 was closed December 12, 1986.

1 REASON FOR CHANGE This FCR 79-410 was implemented in response to NRC requirement per NUREG 0578.

SAFETY EVALUATION

SUMMARY

I This FCR 79-410 provided for a positive and redundant indication of flow through the pilot operated relief valve (RC2A) and the safety relief valves (RC13A and RC13B). This change satisfies the NRC NUREG 0578 requirements of unambiguous valve position indications. These valves are part of the lessons learned TMI #2. This change enhanced system reliabil-4 ity and operability. No new adverse conditions have been created. This change does not constitute an unresolved safety question.

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COMPLETED FACILITY CHANGE' REQUEST FCR NO.79-446 SYSTEM Control Cabinets C5798 and C5799 in Control Room COMPONENT New Equipment CHANGE, TEST OR EXPERIMENT FCR 79-446 installed new cabinets in the control room and cabinet room to facilitate the installation of new indication and control equipment.

This FCR 79-446 was closed December 1, 1986.

REASON FOR CHANGE l This modification for new equipment was installed per TM1-2 lessons learned task numbers 007, 009,.010, 022, 024, 037 and 039.

SAFETY EVALUATION

SUMMARY

Installation of the new control room cabinets in the control. room area, and the cabinet room does not create any new adverse environment because they were seismically mounted and they are seismically qualified for the area of installation, therefore an unreviewed safety quastion does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO 81-049 SYSTEM 480 Volt Power Supply COMPONENT PORV Block Valve RC-Il CHANGE, TEST OR EXPERIMENT This FCR 81-049 changed the power feed to PORV valve RC-ll from Channel 2 to Channel 1 this FCR 81-049 upgraded the installation inside the contain-ment to Class IE.

This FCR 81-049 was closed June 17, 1986.

REASON FOR CHAEGE This change was required per TM1-2 lessons learned task force report.

The change was required in order to obtain redundancy in the pressurizar vent paths which in turn requires that PORV block valve (RC-II) be powered from Channel 1.

SAFETY EVALUATION

SUMMARY

The PORV block valve itself is not a safety grade component. In the event of PORV malfunction such that the PORV stuck open, isolation would be performed by the PORV block valve. In the case of an electrical failure associated with the non-IE PORV block valve the Class IE power feed would be isolated from the rest of the Class IE system by means of the Class IE system circuit breaker in the Motor Control Center (MCC).

Upgrading the power supply of PORV block valve to Class IE is a require-ment of TM1-2 lessons learned task force report.

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Based on the above an unreviewed safety question does not exist.

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COMPLETED FACILITY CHANGE REQUEST FCR NO.79-113 SYSTEM Station Computer COMPONENT Bailey 855/50 CHANGE, TEST OR EXPERIMENT FCR 79-113 modified the unit process computer (Bailey 855/50 by' hardware replacement).

REASON FOR CHANGE.

Design lifetimes of principal components have been or will shortly be exceeded due to prior use by Bailey. Obsolete design of electronics prohibits use of current technology in troubleshooting problems, and spare parts availability is very poor. Current computer problems indicate CPU and memory degradation which could impact unit operation.

This FCR 79-113 was closed November'6, 1986.

SAFETY EVALUATION

SUMMARY

This FCR 79-113, Supp. 2 replaced the operator's two 13" (inch) console with two (2) similar 13" (inch) color units. The operator's keyboard was replaced with a larger unit providing expanded function keys. Present cutouts were enlarged and new mounting panels replaced existing panels.

No supporting cabinet structure was modified.

The main control board display was replaced with a 25" color display.

This cabinet did not require cutting for the installation. Some shielding was added to reduce CRT-interaction but no supporting structure was modified. -Based on the above this modification did not constitute an unresolved safety question.

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TOLEDO

%ss EDISON April 14, 1987 KB87-00050 File: RR 2 (P-6-87-03)

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, March 1987 Davis-Besse Nuclear Power Station Unit 1 Enclosed are ten copies of the Monthly Operating Report for Davis-Besse Nuclear Power Station Unit I for the month of March 1987.

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

Yours truly, ce u a d Louis F. Storz Plant Manager Davis-Besse Nuclear Power Station LFS/MK/lj k Enclosures cc: Mr. A. Bert Davis, w/l Regional Administrator, Region III Mr. James M. Taylor, Director, w/2 Office of Inspection and Enforcement Mr. Paul Byron, w/l 7 NRC Resident Inspector Nuclear Records Management, Stop 3220 / \

LJK/002 THE TOLEDO EOISON COMPANY EOISON PLAZA 300 MAOISON AVENUE TOLEOO, OHIO 43652 1