ML20214H175
| ML20214H175 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/30/1987 |
| From: | Khazrai M, Storz L TOLEDO EDISON CO. |
| To: | Harold Denton, Haller Office of Nuclear Reactor Regulation |
| References | |
| KB87-00179, KB87-179, NUDOCS 8705270301 | |
| Download: ML20214H175 (11) | |
Text
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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.
Davis-Besse Unit 1 UNIT May 13, 1987 DATE Morteza Khazrai COMPLETED BY TELEPHONE
~
Ext. 7290 April 1987 MONnt DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 877 887 3
g7 883 875 2
gg 886 876 3
39 877 872 4
20 879 866 5
21 881 885 6
22 7
879 880 23 8
881 875 24 9
881 431 25 876 663 10 26 879 871 II 7
878 874 12 28 879 871 13 29 14 874 873 39 879 15 31 16 867 INSTRUCTIONS On this format list the average daily unit power leselin MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.
(9/77) k(
8705270301 870430 p
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OPERATING DATA REPORT 50-346 DOCKET NO.
DATE May 13, 1987 COMPLETED BY M rteza shazrai TELEPHONE 419-z4v-3000, Ext. 7290 OPERATING STATUS a
s-esse Unit 1 Nota
- 1. Unit Name:
- 2. Reporting Period:
April 1987 2772
- 3. Licensed Thermal Power (MWt):
925
- 4. Nameplate Rating (Gross MWe):
906
- 5. Design Electrical Rating (Net MWe):
904
- 6. Maximum Dependable Capacity (Gross MWe):
- 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 719 2,879 76,775
- 11. Hours in Reporting Period
- 12. Number Of Hours Reactor Was Critical 719 2,735.1 38,790.2 0
143.9 4,768.7
- 13. Reactor Reserve Shutdown Hours
+
- 14. Hours Generator On-Line 719 2,688.2 37,176.8
- 15. Unit Reserve Shutdown Hours 0.0 0.0 1,732.5
- 16. Gross Thermal Energy Generated (MWH) 1,947,275 5,388,768 86,815,432
- 17. Gross Electrical Energy Generated (MWH)
.646,789 _,
1,760,310 28,722,697
- 18. Net Electrical Energy Generated (MWH) 614,902 1,635,988 26,872,651 100 0.4 e.4
- 19. Unit Service Factor
- 20. Unit Availability Factor 100 93.4 50.7
- 21. Unit Capacity Factor (Using MDC Net).
99.4 66.1 40.7
- 22. Unit Capacity Factor (Using DER Net) 94.4 62.7 36.6
- 23. Unit Forced Outage Rate 0.0 6.6 M.5
- 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 9
INITIAL CRITICALITY INITIAL ELECTRICITY COMMERCIAL OPERATION (9/77)
50-346 DOCKET.NO.
UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME-Divis-Breca Unit 1 DATE May 13, 1987 COMPLETED BY Morteza Khazrai REPORT MONTH April 1987 TELEPHONE 419-249-5000, Ext. 7290 8m "c
Licensee s.e En Cause & Corrective E,
U0 E
3UE Event 0$
Action to No.
Date N
OS E
f$*
Report #
EO 0' O Prevent Recurrence 85 M69
'o 8
4 87 04 25 S
33.7 B
M/A N/A U/A U/A Reactor power was reduced to 50% for planned Containment entry and high pressure condenser water box diff-erential pressure investigation (see Operational Summary for further details).
5 I 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 5
C-Operational Error (Explain) 9-Other (Explain)
Exhibit I - Same Source (9/77)
H-Other (Explain)
OPERATIONAL
SUMMARY
APRIL 1987 The reactor power was maintained at approximately 100% power until 0415
' hours on Ap'ril 25, 1987, when reactor power was reduced to 50% power for planned Containment entry and high condenser water box differential pressure investigation.
The Low Pressure Condenser Loop 2 water box was cleaned and differential pressure dropped from 30 ft. H O (before cleaning) to 20.5 ft. H O (after 2
2 cleaning).
Containment entries were made for repair of Hot Leg Level Monitoring System Channel 2 transmitter and for inspection of the PORV and code safety valves for leakage and for leakage in the Decay Heat Valve Pit.
The reactor power was increased to approximately 100% power at 1310 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.98455e-4 months <br /> on April 26, 1987, and maintained at this power level for the rest of the
- month, i
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REFUELING INFORMATION.
DATE: -April 1987 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 50.59)?
Ans: Expect the Reload Report to require standard reload fuel design Technical Specifications changes (2. Safety Limits and Limiting Safety System Settings, 3/4.1 Reactivity Control Systems, 3/4.2 Power Dis-tribution Limits and 3/4.4 Reactor Coolant System.)
I 5.
Scheduled date(s) for submitting proposed licensing action and supporting information: December, 1987 6.
.Important licensing considerations associated with. refueling, e.g.,
-i -
new or different' fuel design or supplier, unreviewed design or performance analysin 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.
1
'(a) 177 (b) 204 - Spent Fuel Assemblies 8.
-The present licensed spent fuel pool storage capacity and the size of l
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)
{
9.
The projected date of the last refueling that can be discharged to j
the spent fuel pool assuming the present licensed capacity.
s-g' Date:
1995 - assuming ability to unload the entire core into the 1
spent fuel pool is maintained.
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COMPLETED FACILITY CHANGE REQUEST-FCR NO:- 79-0293 SYSTEM: Auxiliary Feedwater System COMPONENT: Pressure Switches.
CHANGE, TEST OR EXPERIMENT: FCR 79-0293 replaced the Static-0-Ring pressure switches with Static-0-Ring stainless steel diaphrages and stainless steel pressure port type switches.
This FCR was closed March 20, 1987.
REASON FOR CHANGE: This modification was incorporated due to corrosion of the piston due to permeation of condensate through the Buna-N diaphragm onto the piston.
SAFETY EVALUATION
SUMMARY
- The new Static-0-Ring pressure switches with stainless steel diaphragm and ports have the same pressure ratings and setpoints as the original switches. The modification of these Static-0-Ring pressure switches does not create an unreviewed safety question.
l COMPLETED FACILITY CHANGE REQUEST FCR NO: 85-0016 SYSTEM: Control Room Emergency Ventilation System COMPONENT: Volume Control Dampers 4
CHANGE, TEST OR EXPERIMENT: FCR 85-0016 modified the damper actuator support to each Control Room Emergency Ventilation System cooling coil volume control damper.
This FCR was closed April 23, 1987.
REASON FOR CHANGE: This FCR is corrective action for Surveillance Report 84-47.
SAFETY EVALUATION
SUMMARY
- To conform to standard design practices to ensure long term integrity, the damper actuator supports were modified under this FCR. This modification does not adversely affect the safety function of the damper or the Emergency Ventilation System.
This modification does not create an unreviewed safety question.
l
-COMPLETED FACILITY CHANGE REQUEST FCR NO: 85-0048
-SYSTEM: Emergancy Diesel Generators COMPONENT:
PCV2987. PCV2988, PCV2989, and PCV2994 CHANCE,-TEST OR EXPERIMENT:
FCR 85-0048 changed the setpoints on PCV2987, PCV2988, PCV2989, and PCV2994 from 200 psig to a new setpoint of 180 psig.
This FCR was closed on February 17, 1987.
REASON FOR CHANCE: The air start solenoids in use are rated at 200 psig maximum. Lowering line pressure to 180 psig prevents the air start solenoids from being over-ranged.
SAFETY EVALUATION
SUMMARY
- This change loweted the operating setpoint of the Emergency Diesel Generator air start system from 200 psig to 180 psig. A pressure of approximately 180 psig is adequate to at
& art the Emergency Diesel Generators. This change prevents overpressurization of the air start solenoid valves and helps avoid leakage through the valves. The lower pressure setpoints for the air pressure regulators will also result in less wear and maintenance since run time of the air compressors to maintain pressure will be reduced.
.o COMPLETED FACILITY CHANGE REQUEST FCR NO: 85-0238 SYSTEM: Auxiliary Building Non-Radioactive COMPONENT: MV-5314 CHANGE, TEST OR EXPERIMENT: FCR 85-0238 modified the control circuit to ensure the ventilation path is open for Room 428 in the event offsite power is lost.
This FCR was closed March 20, 1987.
REASON FOR CHANGE: This modification ensures that damper MV-5314 is open to provide cooling / ventilation for Room 428 (Low Voltage Switchgear Room) during loss of offsite power.
SAFETY EVALUATION
SUMMARY
- The safety function of the Low Voltage Switch-gear Room Exhaust Fan is to maintain the temperature below a level which would adversely affect the reliability of the switchgear. No changes to the normal operation of the fan and damper are being made. This FCR changed the damper failure position from closed to open on a loss of electrical power.
This modification does not increase the probability of occurrence of a malfunction of equipment in accordance with the safety evaluation in the SAR. This FCR does not involve an unreviewed safety question.
COMPLETED FACILITY CHANGE REQUEST FCR NO: 85-0259 SYSTEM: Station and Instrument Air COMPONENT: N/A CHANCE, TEST OR EXPERIMENT:
FCR 85-0259 installed an air filter in the station air header bypass line.
. This FCR was closed April 24, 1987.
REASON FOR CHANGE: SA-2010 exceeded the permissible leak rate test limit and was found to have a dirty seat and internals.
SAFETY EVALUATION
SUMMARY
FCR 85-0259 installed an in-line filter (F86) upstream of Containment Isolation Valve SA2010. This improves the reliabiity of the Station Air System by reducing particle contamination of Containment breathing air, which will minimize the likelihood of local leak rate test failure.
The changes to the Station Air System in no way degraded the system. The modification does not increase the possible consequences of an accident previously evaluated in the SAR.
Based on the foregoing consideration, it is concluded the addition-of an air filter in the station air supply does not involve an unreviewed safety question.
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MXEDO
%mm EDISON May 13, 1987 KB87-00179 File: RR 2 (P-6-87-04)
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, April 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 April 1987.
If you have any questions, please feel free to contact Morteza Khazrai at (419) 249-5000 Extension 7290.
Yours truly, ved
' ry Louis F. Storz Plant 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 L
[
t Nuclear Records Management, Stop 3220 t
'q (k
LJK/002 THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 43652
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