ML20129H156
| ML20129H156 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/30/1985 |
| From: | Andrews R, Matthews T OMAHA PUBLIC POWER DISTRICT |
| To: | Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| LIC-85-318, NUDOCS 8507180538 | |
| Download: ML20129H156 (10) | |
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t AVERAGE DAILY UNIT POWER LEVEL 50-285 DOCKET NO.
Fort Calhoun Station ugi7 July 8, 1985 DATE T. P. Matthews COMPLETED BY TELEPIIONE (402) 536-4733 June, 1985 MONm DAY AVER AGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (MWe-Net)
(MWe-Net) 476.6 474.1 i7 479.4 is 475.8 2
479.4 477.7 3
i, 479.5 477.4 4
29 480.0 475.0 5
3, 479.2 474.2 6
22 4
7 23 451.0 473.4 g
24 450.1 470.2 9
25 454.6 469.0 39 3
475.0 11 27 470.9 478.8 474.8 12 28 479 7 475.2 13 29 479.3 472.5 i4 39 477.4 is 3,
474.8 16 INSTRUCTIONS On this format. fist the average daily unit power levelin MWe Net for each day in the reporting month. Compute to the nearest whole megawatt.
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8507180538 850630 PDR ADOCK 05000285 R
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OPERATING DATA REPORT DOCKET NO.
50-285 DATE July 8, 1985 COMPLETED BY T. P. Matthews TELEPHONE (402) 536-4733 OPERATING STATUS
- 1. Unit Name:
Fort Calhoun Station Notes
- 2. Reporting Period:
June, 1985 1500
- 3. Licensed Thermal Power (MWt):
502
- 4. Nameplate Rating (Gross MWe):
478
- 5. Design Electrical Rating (Net MWe):
502
- 6. Maximum Dependable Capacity (Gross MWe):
470
- 7. Maximum Dependable Capacity (Net MWe):
- 8. If Changes Occur in Capacity Ratings (Items Number 3 Through 7) Since Last Report. Give Reasons:
N/A
- 9. Power Lesel To which Restricted. If An{o(Net MWe):
n ne
- 10. Reasons For Restrictions.If Any:
This Month Yr.-to-Date Cumulative II. Hours In Reporting Period 720.0 4.343.0 103.120.0
- 12. Number Of Hours Reactor Was Critical 720.0 4,316.6 79.596.8
- 13. Reactor Reserve Shutdown flours 0.0 0.0 1.309.5
- 14. Hours Generator On Line 720.0 4,308.7 78,976.l'
- 15. Unit Reserve Shutdown Hours 0.0 0.0 0' 0
- 16. Gross Thermal Energy Generated (MWH) 1,069,152.0 6,368,248.5 100,555.015.5
- 17. Gross Electrical Energy Generated (MWil) 357,360.0 2,158,384.0 32,928,009.0
- 18. Net Electrical Energy Generated (MWH) 340.692.1 2,059,786.4 31.471.423.7 100.0 99.2 76.6
- 19. Unit Service Factor
- 20. Unit Availability Factor 100.0 99.2 76.6
- 21. Unit Capacity Factor (Using MDC Net) 99.0 99.2 66.3
'22. Unit Capacity Factor (Using DER Net) 99.0 99.2 64.1 0.0 0.0 3.6
- 23. Unit Forced Outage Rate
- 24. Shutdowns Scheduled Over Nest 6 Months (Type. Date.and Duration of Each):
1985 Refuelino Shutdown is tentatively scheduled for October.1985 with start up in December. 1985.
- 25. If Shut Down At End Of Report Period. Estimated Date of Startup:
N/A
- 26. Units in Test Status (Prior io Commercial Operation): N/A Forecast Achiesed INITIAL CRITICA LITY INITIAL ELECTRICITY COMMERCIAL OPER ATION (9/77) i
UNIT SHUTDOWNS AND POWER REDUCTIONS DOCKET NO.
50-285 UNIT NAME Fort Calhoun Station DATE July 8. :.985 June, 1985 COMPLETED BY T. P. Mauthews REPORT MONTH TEl.EPHONE (402) 536-4733 t
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3 Y
Lisensee E r, f%
Cause & Currective No.
Date g
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Event gg Q
Action to 3;jj=g Repor =
mV hL Prevent Recurrence
$5 5
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There were no unit shutdowns during the month of June.
I 2
3 4
F: Forced Reason:
Method:
Exhibit G -Instructions S: Scheduled A. Equipment Failure (Explain) 1 Manual for Preparation of Data B-Maintenance of Test 2 Manual Scram.
Entry Sheets for Licensee C-Refueling 3 Automatic Scram.
Event Report tLER) File (NUREG-D Regularory Restriction 4-Other (Explain 1 01611 E-Operator Training & Ucense Examination F-Administratise 5
G-Operational Error (Explain)
Eshibit I-Same Source (9/77)
Il-Ot her (Explain)
Refueling Information Fort Calhoun - Unit No. 1 Report for the month ending June, 1985 1.
Scheduled date for next refueling shutdown.
October, 1985 2.
Scheduled date for restart following refueling.
December, 1985 3.
Will refueling or resumption of operation thereafter require a technical specification change or other license amendment?
Yes a.
If answer is yes, what, in general, will these be?
Technical Specifications change to accommodate increased radial peaks due to further reduction in radial leakage.
b.
If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Comittee to deter-mine whether any unreviewed safety questions are associated with the core reload.
c.
If no such review has taken place, when is it scheduled?
4.
Scheduled date(s) for submitting proposed licerc.ng action and support information.
September, 1985 5.
Inportant 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.
Methodology Changes June, 1985 6.
The number of fuel assenblies: a) in the core 133 assemblies b) in the spent fuel pool 305 c) spent fuel pool storage capacity 729 d) planned spent fuel pool May be increased storage capacity via fuel pin consolidation 7.
The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity.
1996 Prepared by ed/
Date July 1.1985
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OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No. 1 June, 1985 Monthly Operations Report I.
OPERATIONS
SUMMARY
Fort Calhoun Station operated at 100% power throughout June,1985.
Spent fuel assembly D005 was shipped off-site as part of a joint study by the Department of Energy and the Combustion Engineering Owners Group on high burnup fuel.
The first shipment of new fuel for Cycle 10 arrived on-site June 30.
The efforts to reduce the amount of waste stored on-site continued.
Shipments of waste requiring high integrity containers have reduced the inventory of long-term waste in the Auxiliary Building.
The first of the replacement low pressure feedwater heaters arrived on-site in preparation for the Fall refueling outage.
Two engineers sat for the NRC Senior Reactor Operator exam and four operators sat for NRC Reactor Operator exams during June.
A team of 13 evaluators from the Institute of Nuclear Power Operations (INP0) conducted the annual review of Fort Calhoun Station. The team interfaced with all areas of plant activities and provided a good exchange of information.
No safety valve or PORV challenges or failures occurred.
A.
PERFORMANCE CHARACTERISTICS
'None B.
CHANGES IN OPERATING METHODS None C.
RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS 1
None
Monthly Operations Report June, 1985 Page Two D.
CHANGES, TESTS AND EXPERIMENTS CARRIE0 OUT WITHOUT COMMISSION APPROVAL Procedure Description SP-FAUD-1 Fuel Assembly Uplift Condition Detection.
This procedure.did not constitute an unreviewed safety question as defined by 10CFR50.59 since it only involved the evaluation of data from a surveillance test to verify that a fuel assembly uplift condition did not exist.
SP-VA-80 Hydrogen Purge System Test.
This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because this procedure only checks operability of fans and cleanliness of the filters.
SP-SFS-1 Shipment of Spent Fuel in NLI 1/2 Cask.
This procedure, which provided for the shipment of a spent fuel bundle, did not constitute an unreviewed safety question because technical specifiutions prohibiting using the crane from carrying a load over irradiated fuel were observed, an approved cask was used for the shipment and appropriate radiological require-ments were addressed.
System Acceptance Committee Packages for June,1985:
Package Description / Analysis l
EEAR FC-83-182 Dry Pipe Fire Protection in Diesel Generator Rooms.
This modification converted the existing wet pipe sprinkler system to a dry pipe system.
This conversion will decrease the probability of frozen piping during cold weather, thus increasing system availability.
This modifi-cation has no adverse effect on the safety analysis.
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Monthly Operations Report June, 1985 Page Three D.
CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued)
System Acceptance Committee Packages for June,1985:
(Continued)
Package Description / Analysis EEAR FC-85-53 Potable Water System Isolation Valves.
This modification provided for the installation of additional valves in the potable water header.
These valves will provide the capability to isolate sections of the header for maintenance.
This modification has no adverse effect on the safety analysis.
EEAR FC-82-47 New Dumping of Concentrates.
This modification provided for the repiping of the concentrates dump line with a more direct path to the concentrate tanks.
This modifica-tion provides better dumping ability for the evaporator and original piping code standards were maintained.
This modification has no adverse effect on the safety analysis.
EEAR FC-85-74 Isolation of Demineralized Water from Gas Stripper.
This modification permanently isolates demin-eralized water feed to the gas stripper which is not used.
This modification has no adverse effect on the safety analysis.
EEAR FC-84-139 TMLP Delta Noise Elimination.
This modification provided for the installation of filter capacitors across the inputs to the reactor protective system thermal margin-low pressure (TMLP) calculator. These capacitors will filter out most of the disruptive noise in the signals which could lead to a spurious trip.
No change to the existing operation of these circuits was made.
This modification has no adverse effect on the safety analysis.
w e
Monthly Operations Report June,-1985 Page Four D.
CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued)
System Acceptance Comnittee Packages for June,1985:
(Continued)
Package Description / Analysis EEAR FC-85-35 VHPT Dual Pot Replacement.
This modification provided for the one-for-one replacement of the original RPS equipment with' upgraded modules to eliminate experienced failures.
The function and operation of the new modules are identical to the original equipment.
This modification has no adverse effect on the safety analysis, i
EEAR FC-82-108, ATCOR Startup.
Part 1 This modification made only minor equipment changes that will make the waste solidifica-tion system more reliable and simpler to use.
This modification has no adverse effect on the safety analysis.
EEAR FC-84-29 RPS Dual Pot Upgrade.
This modification provided for the one-for-one replacement of existing system hardware with a higher reliability, less failure-prone piece of equipment.
The RPS function was not altered in any way.
This modification has no adverse effect on the safety analysis.
EEAR FC-84-205A Barrier Installation.
This modification did not involve safety related equipment; therefore, has no adverse effect on the safety analysis.
EEAR FC-85-51 Replace Portal Radiation Monitors.
This modification replaced the existing portal radiation monitors with monitors that increased detection time, are more sensitive and easier to calibrate. This modification has no adverse effect on the safety analysis.
Monthly Operations Report June, 1985 Page Five E.
RESULTS OF LEAK RATE TESTS None F.
CHANGE IN PLANT OPERATING STAFF None G.
TRAINING Six candidates sat for an NRC operator's license exam on June 18 (four for reactor operator and two for senior reactor operator).
All six candidates passed the oral walkthrough portion of the exam.
Training for Equipment Operator-Nuclear (Turbine Building) commenced. General employee, C/RP Technician, emergency plan and
-maintenance training continued.
H.
CHANGES, TESTS AND EXPERIMENTS REQUIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT T0 10CFR50.59 None II. MAINTENANCE (Significant Safety Related)
None W. Gary Gates Manager Fort Calhoun Station
6 i
Omaha Public Power District 1623 Harney Omaha Nebraska 68102 402/536 4000 July 12, 1985 LIC-85-318 Mr. James M. Taylor, Director Office of Inspection and Enforcement U. S. Nuclear Regulatory Comission Washington, DC 20555
Reference:
Docket No. 50-285
Dear Mr. Taylor:
June Monthly Operating Report Please find enclosed ten (10) copies of the June,1985 Monthly Operating Report for the Fort Calhoun Station Unit No.1.
Sinc r ly, M/
R. L. Andrews Division Manager Nuclear Production RLA/TPM/dao Enclosures
-cc: NRC Regional Office Office of Management & Program Analysis (2) f1r. R. R. Mills - Combustion Engineering Mr. T. F. Polk - Westinghouse Nuclear Safety Analysis Center INP0 Records Center American Nuclear Insurers NRC File
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