ML20073H298

Monthly Operating Rept for Mar 1983

ML20073H298
Person / Time
Site:	Fort Calhoun
Issue date:	04/12/1983
From:	Matthews T OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20073H284	List: ML20073H279 ML20073H298
References
NUDOCS 8304180543
Download: ML20073H298 (16)

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• AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 UNIT Fort Calhoun Station DATE April 12, 1983 COMPLETED BY T. P. Matthews TELEPIIONE (402)S36-4733 MONTli fia rch,1983 DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAILY POhER LEVEL (MWe-Net) (Mwe. Net) 1 0.0 37 0.0 2 0.0 18 0.0 3 0.0 39 0.0 4 0.0 20 0.0 5 0.0 21 0.0 6 0.0 22 0.0 7 0.0 23 0.0 8 0.0 24 0.0 9 0.0 25 0.0 to 0.0 26 0.0 11 0.0 27 0.0 12 0.0 28 0.0 13 0.0 29 0.0 14 0.0 39 0,0 15 0.0 3, 0,0 16 0.0 INSTRUCTIONS On this format. list the average daily unit power levelin MWe Net for each day in the reporting ini ,th. Compute to the nearest whole megawatt.

(9/771 0304180543 830412 PDR ADOCK 03000285 R PDR

OPERATING DATA REPORT DOCKET NO. 50-285 DATE April 12, 1983 CO.1PLETED

\ BY T. P. Matthews TELEP110NE (402)S36-4733 OPERATING STATUS Fort Calhoun Station Notes

1. Unit Name:

2. Reporting Period: March, 1983

3. Licensed Thermal Power (51Wt):

1500

4. Nameplate Rating (Gross A!We): 501

5. Design Electrical Rating (Net 31We): 478

6. Sta[imum Dependable Capacity (Gross SIWE): 501

7. Staximum Dependable Capacity (Net SIWE): 478

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

ft/A

9. Power Level To which Restricted.If Any(Net 31We): N/A

10. Reasons For Restrictions. If Any: None This Alonth Yr..to.Date Cumulative I1. Ilours In Reporting Period 744.0 2,160.0 83,401.0

12. Number Of flours Reactor Was Critical 0.0 0.0 64,110.5

13. Reactor Reserve Shutdown flours 0.0 0.0 1,309.5

14. Ilours Generator On.Line 0.0 0.0 62,947.5

15. Unit Resene Shutdown liours 0.0 0.0 0.0

16. Gross Thermal Energy Generated (31Wil) 0.0 0.0 77,616,548.4 .

17 Gross Electrical Energy Generated 151WH) 0.0 0.0 25.735.333.5 l 18. Net Electrical Energy Generated (51Wil) 0.0 0.0 24,330,034.4

19. Unit Senice Factor 0.0 0.0 75.5

20. Unit Availability Factor 0.0 0.0 75.5

21. Unit Capacity Factor (Using SIDC Net) 0.0 0.0 63.3

22. Unit Capacity Factor (Using DER Net) 0.0 0.0 63.0 1 23. Unit Forced Outage Rate 0.0 0.0 3.9 l 24. Shutdowns Scheduled Oser Next 6 31onths(Type. Date,and Duration of Each):

l 25. If Shut Down At End Of Report Period. Estimated Date of Startup: April 7, 1983

26. Units in Test Status (Prior to Commercial Operation): N/A Forecast Achiesed INITI A L CRITICA LITY INITi \ L ELECTRICITY CO\1%!ERCIAL OPER ATION (9/77)

1 s .

DOCKET NO. 50-285 UNITSilUTDOWNS AND POWER REDUCTIONS UNIT NAME Fort Calhoun Station DATE Aoril 1?. 14R3 REPORT MONTil flarch 1983 COMPLETED BY T. P. Matthews TELEPilONE (4021853fi-4731 n.

- E c

, jE 3 j .Y 3 Licens.:e E r, je, Cause & Corrective No. Date g 3g ,3 g 5 Event gy g3 Action to H 5,s mV o' 3: 3 dj =!

Report a u Prevent Recurrence 6

82-06 821206 S 2,766 C 4 N/A XX XXXXXX 1982/1983 refueling outage com-menced December 6, 1982.

I 2 3 4 F: Forced Reason: Method: Exhibie 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 (LER) File (NUREG-D-Regulatory Restriction 4-Other (Explain) 0161)

E-Operator Training & License Examination F Administrative 5 G-Operational Error (Explain) Exhibit i . Same Source (9/77) II-Other (Explain t

Refueling Information Ebrt Calhoun - Unit No.1 Report for the month ending March 1983 .

1. Scheduled date for next refueling shutdown. March 1984

2. Scheduled date for restart following refueling. May 1984

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?

A Technical Specification Change

b. If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Comnittee 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 licensing Methodology - Dec. 1983 action and support information. Tech. Specs.- Feb. 1984

5. Important licensing considerations associated with refueling, e.g. , new or different fuel design or i supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures.

6. The number of fuel assenblies: a) in the core 133 assemblies b) in the spent fuel pool 265 c) spent fuel pool storage capacity 483 d) planned spent fuel pool r

storage capacity 728

7. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity. 1985 Prepared by / (W,{/) Date April 1, 1983 o o

OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No.1 March, 1983 Monthly Operations Report I. OPERATIONS

SUMMARY

Fort Calhoun Station continued refueling outage operations during the month of March.

Reactor core loading was performed March 4 through 7 using the new loading pattern to reduce neutron fluence on the vessel. The reactor head was set March 11. Reactor coolant system heatup began March 28 reaching hot shutdown March 30.

Many system hydrostatic tests were perfomed including the RCS hydro at 2275 psig on March 29 and the "B" steam generator hydro on March 17.

The high pressure turbine rotor has been reassembled.

Two new personnel were added to the Operations staff March 14.

No safety valve or PORV challenges occurred.

A. PERFORMANCE CHARACTERISTICS LER Number Deficiency 83-001 During the refueling outage the pressurizer code safety relief valves were removed and sent to an outside laboratory for a check of the setpoint pressure with steam. The setpoint for one of the two safety valves was found to be i

beyond the Technical Specification 2.1.6(1) limit l

of 2500 psia to 2545 psia i 1%. RC-141 actually lifted at 2577 psia. The lift setpoint on the redundant valve RC-142 was within the specified tolerance.

I

a e e Monthly Operations Report March, 1983 Page Two B. CHANGES IN OPERATING METHODS None.

C. RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS Surveillance tests as required by the Technical Specifications Section 3.0 and Appendix B, were performed in accordance with the annual surveillance test schedule. The following is a summary of the surveillance tests which resulted in Operation Incidents and are not reported elsewhere in the report:

Operation Incidents Deficiency 01-1664 ST-RPS-8, During the performance of ST-RPS-8, F.2, D/765 F.2 pressure switch found out-of-calibration.

01-1612 ST-MSSV-1, During the performance of ST-MSSV-1, F.1, MS-275, F.1 MS-278, MS-280, MS-282 failed to actuate.

01-1658 ST-PORV-1, During the perfonnance of ST-PORV-1, F.2, l

F.2 pressure transmitter PT-115 and anplifier R4-115 found out-of-calibration.

O. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT C0iMISSION APPROVAL PROCEDURE DESCRIPTION SP-IC-11 Disposal of Irradiated Incore Detectors. Completed per procedure.

Not an unreviewed safety question because all operations were performed in accordanc< 4! 'n the Facility Operating License; approved writter. p ;edures; and a valid Radia-tion Work Permit.

SP-CSF-1 Carbon Steel Fastenert Inser ice Testing. Completed per procedure.

Not an unreviewed safety question because it is only an inspection.

SP-ECT-1 Eddy Current Testing of Heat Exchanger Tubes.

Not an unreviewed safety question because each one of these SP's only provided for testing on non-safety related equipment.

Monthly Operations Report March, 1983 Page Three D. CHANGES, TESTS AND EXPERIMENfS CARRIED OUT WITHOUT COMMISSION APPROVAL System Acceptance Committee Packages 1982:

Package Description / Analysis EEAR/FC-81-100 Main Feedwater Isolation This modification will enhance safety by providing a new safeguard signal to isolate main feedwater on certain events. Primary events which may be affected by this modification are MSLB and SB-LOCA. Main feedwater isolation reduces the severity of RCS cooldown for the MSLB, and reduces containment pressure for both events. This modification will increase plant safety and consequently, has no adverse effect on the safety analysis.

EEAR/FC-79-162 Station Lighting Transformers - 4160 V Supply This modification provides a safer and quicker means to transfer station lighting from one 4160 V bus to the other during surveillance testing and/or maintenance. This modification has no adverse effect on the safety analysis.

EEAR/FC-81-21B Component Cooling Water Isolation to Reactor Coolant Pumps This modification ensures that conponent cooling water will not be isolated to the reactor coolant puaps following either a CIA signal or loss of DC power. It is acceptable to allow valves HCV-438B and HCV-438D to remain open in all cases except where a CIA signal and component cooling water pressure low signal are simultaneously present. This modification will ensure that the reactor coolant punps will operate in a post-DBA condition while a CIA signal is present.

This modification will increase plant safety and consequently, has no adverse effect on the safety analysis.

EEAR/FC-81-139 Beta Radiation Shielding (Penetration Lead Wires)

This modification increases the reliability of certain safety related cables under accident and post-accident conditions. Existing cables are being modified only to provide greater radiation protection. This modification has no adverse effect on the safety analysis.

Monthly Operations Report March, 1983 Page Four D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL Package Description / Analysis (continued)

EEAR/FC-81-22 Replace LCV-1173 This modification is only to replace the existing vavle with the new one. This modification has no adverse effect on the safety analysis.

EEAR/FC-83-22 Emergency Feedwater Tank Ventilation This modification does not compronise the functional ability of the tank. The integrity of the tank is being maintained by using Section VIII ASME Design Criteria and constructing the modification in accordance with that code. The probability of a failure or modification of FW-19 is not increased.

This modification has no adverse effect on the safety analysis.

EEAR/FC-79-182E Safety Injection Pump Room level Indicators This modification installs level indicators which will provide a means to monitor excessive leakage into the sumps. This modification has no adverse effect on the safety analysis.

EEAR/FC-81-99 Installation of Heated Junction Thermocouple Probe Part 9 Holder in the Upper Guide Structure This modification installs heated junction thermo-couple probe holders inside two enpty CEA shrouds which protect said probe holders fran normal operating cross-flow loads as well as flow-down loads. This modification has no adverse effect on the safety analysi s.

EEAR/FC-81-99 tiodification of Electrical Penetrations A-10 and D-11 Part 3 Modifications of these two electrical penetrations are necessary to allow for the installation of the reactor vessel level monitoring system. This modification has no adverse effect on the safety analysis.

EEAR/FC-81-99 Installation of Modified ICI Flanges Part 5 This modification was necessary in order to allow for installation of a reactor vessel level monitoring system. This modification has no adverse effect on the safety analysis.

Monthly Operations Report March, 1983 Page Five D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL Package Description / Analysis (continued)

EEAR/FC-82-56 TEC Charge Converter Replacment This modification is to replace the TEC charge converter. This modification has no adverse effect on the safety analysis.

EEAR/FC-82-38 Diesel Generator 1 and 2 Vent Line Replacement This modification is to replace the cooling water vent lines in the diesel generators with new flexible hoses. These hoses will eliminate the transmission of vibrations from the diesel generators to the accessory racks. This modification has no adverse effect on the safety analysis.

EEAR/FC-82-28 Steam Generator and Pressurizer Pressure Transmitter Replacement This modification provides for replacement of the

, existing pressure transmitters with new nuclear qualified pressure transmitters. This modification will increase plant safety and consequently, has no t

adverse effect on the safety analysis.

EEAR/FC-81-99 Machining of ICI Flanges in Place on Reactor Vessel Part 11 This modification will allow for machining of the ICI flanges to restore the gasket seating surfaces thereby improving the probability of a leak-tight connection.

This modification has no adverse effect on the safety analysis.

EEAR/FC-83-35 HCVS-2908 Seismic Restraint This modi fication modi fies the support on HCVS-2908 in order to ensure that the valve will cycle properly and j to ensure that it is adequately supported. This modification has no adverse effect on the safety analysis.

i Monthly Operations Report March, 1983 Page Six

0. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT C0'1 MISSION APPROVAL Packane Description / Analysis (continued)

EEAR/FC-82-143 Key-Lock Switches for Containment Purge Valve Control Ci rcui ts This modification installs new key-operated switches in panel AI-44. These switches (HC-742-1C, HC-742-10, HC-742-2C, and HC-742-20) are part of the containment purge valve control system. This modification will provide a better administrative control of the containment purge valves. This modification has no j adverse effect on the safety analysis. I EEAR/FC-82-11 Reactor Coolant Pump Motor Cover l

This modification installs handles on the reactor coolant pump motor covers. This modification will i have no effect on plant safety and consequently, has no effect on the safety analysis.

EEAR/FC-82-171 Bypass Valves for Check Valves SI-194,197, 200, 203 This modification is necessary to allow for bypass valves in order to perform the safety injection hydro test. This modification has no adverse effect on the safety analysis.

EEAR/FC-78-37 Personnel Airlock Doors This modification was done in order to allow for easier personnel airlock door operation and to ensure that there is a reliable interlocking system during outages. This modification has no adverse effect on the safety analysis.

EEAR/FC-79-15 Replacement of Reactor Pressure Vessel Head Insulation This modification allows for the removal of existing RPV head insulation and the installation of new reflective type insulation. This modification has no adverse effect on the safety analysis.

EEAR/FC-82-33 Cable Spread Roan Halon Dampers This modification installs halon system dampers in the cable spread room. This modi fication will increase safety by setting alarms and closing dampers upon manual initiation. This modification has no adverse effect on the safety analysis.

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Monthly Operations Report March, 1983 Page Seven D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL Packace Description / Analysis (continued)

EEAR/FC-82-32 Volume Control Tank Relief Repiping This modification is to replace check valve CH-284 and install a new block valve CH-534 to give added protection against radioactive gases releasing back from the volune control tank. The existing check valve CH-284 will be replaced with another better seating valve. The new block will be for isolation purposes between the volume control tank and the check valve only when the check valve needs servicing. This modification has no adverse effect on the safety analysis.

EEAR/FC-82-167 220 Volt Power for Temporary Trailers This modification was done in order to provide 220 volt power for tanporary office trailers to be used during the outage. This modification is an administrative type only and has no effect on the safety analysis.

EEAR/FC-82-40 SI-222 Relief Valve Discharge This modification was done in order to reroute the discharge of SI-222 from the waste gas header to the reactor coolant drain tank. This modification has no adverse effect on the safety analysis.

EEAR/FC-80-29 Turbine Room Floor bodification This modification is for additional structural support to the grating of the turbine room floor. This modification is for the secondary plant side and has no effect on safety analysis.

EEAR/FC-81-98 Component Cooling Water Drain Valves This modification installs 3/4-inch drain valves upstream of existing valves AC-283, 284, 285 and 286 on containnent coolers VA-1 A,1B, VA-8A, 88. This modification will enhance maintenance on the component cooling relief valves thereby increasing the margin of sa fety. This modification has no adverse effect on the safety analysis.

l Monthly Operations Report March, 1983 ,

Page Eight l D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL Package Description / Analysis (continued)

DCR-75B-28 AC/RW Interface Valves This modification was performed to provide air supply to the AC/RW interface valves. This modification has no adverse effect on the safety analysis.

EEAR/75B-20 Discharge Tunnel Sampling Apparatus This modification involved design of a structure to raise and lower sample nets from the discharge tunnel.

This modification has no impact on the safety 4 analysis.

EEAR/DCR-FC-76-54 Spent Fuel Pool Crosstie

. This modification provides for an installation of a system for a means of a crosstie between the spent fuel pool cooling system and the shutdown cooling system. This modification will have no adverse effect on the safety analysis.

EEAR/FC-79-39 Replacement of Raw Water Valves i

This modification provides for replacement of raw water isolation valves. This modification has no adverse effect on the safety analysis.

EEAR/FC-80-125 Control of Gas Leaks from WD-32 Part 1

( This modification removed valve HG-102 and capped the line that it was in, in an effort to remove an unused part of the hydrogen gas systems. This modification i has no adverse effect on the safety analysis.

EEAR/FC-82-136 Securing Containment Integrity This modification provides for installation of time delays and voltage dropping resistors on HCV-921 and l 922. This modification will provide for better i

isolation in case of an accident. This modi fication i

has no adverse effect on the safety analysis.

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Monthly Operations Report March, 1983 Page Nine D. CHANGES, TESTS AND CXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL Packace Description / Analysis (continued)

EEAR/FC-80-38/ Replacment and Relocation of Containment High and 82-23/81-123 Low Pressure Safety Injection Flow Transmitters This modification provides for replacement of FT-313 with a new nuclear qualified flow transmitter. This modification will increase plant safety and consequently, has no adverse effect on the safety analysis.

EEAR/FC-81-125 Reactor Coolant Pump Hatch Safety Items This modification installed hand rails and rail sockets in order to allow for easier accessibility during maintenance. This modification has no effect on plant safety or the safety analysis.

EEAR/FC-81-161 Safety Injection Tank Level Switches Parts 1 and 2 This modification provides for the installation of the safety injection tank narrow range level system. The modification of the safety injection tank instrumentation does not alter the performance of the system, but provides for a redundant system that will allow for more reliability. This modification will have no adverse effect on the safety analysis.

EEAR/FC-81-174 Access Platforms for RC-138 and RC-139 This modification installed access platforms to RC-138 and RC-139 to provide for better maintenance access.

This modification has no effect on plant safety, nor on the safety analysis.

EEAR/DCR-74A-95 Hotel Waste Tank Modification This modification was done in order to discharge hotel waste tanks to overboard discharge header immediately before WD-632 so RM0-55 would monitor the waste release, but monitor tanks would be bypassed. This modification has no adverse effect on the safety analysis.

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Monthly Operations Report March, 1983 Page Ten D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL Packaqe Description / Analysis (continued)

EEAR/FC-82-123 Preparation of Spare Electrical Penetration E-11 The purpose of this modification was to prepare the spare penetration E-11 so it could be used for the 1983 reactor vessel examination. This modification has no effect on plant safety, nor the safety analysis.

EEAR/FC-82-58 Battery Discharge Test Equipnent The purpose of this modification was to install equipment necessary for completing battery discharge test to be completed during the refueling outage.

This modification has no effect on plant safety and consequently, has no effect on the safety analysis.

EEAR/FC-82-66 Alternate Safe Shutdown / Pressurizer Heater Control This modification provides for local control of the back-up pressurizer heaters, isolation of control room circuits, and local start of back-up pressurizer heaters. This modification has no effect on plant safety and consequently, has no effect on the safety analysis.

EEAR/FC-82-57 Replacement of ASCO Solenoid Valves This modification provided for the replacement of

existing ASCO solenoid valves with qualified

( repl acements. This modification has no effect on plant safety and consequently, has no effect on the safety analysis.

EEAR/FC-83-43 Control of Heavy Loads This modification provides for redundant upper-travel l

limit switches on the polar crane auxiliary and main hoist. This modification has no effect on plant safety and consequently, has no effect on the safety a nalysi s.

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Monthly Operations Report March, 1983 Page Eleven D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL Package Description / Analysis (continued)

EEAR/FC-82-06 Load Shed of 480 Volt Raw Water Strainers This modification installed a contact start-stop switch on the raw water pump strainers. In this way, a manual restart of the raw water pump strainer will not be necessary after a load shed. This modification has no effect on plant safety and consequently, has no effect on the safety analysis.

E. RESULTS OF LEAK RATE TESTS All leak rate tests perfonned during January, February and March will be reported to the PRC per a special report (which concerns itself with all CONT-2, 3, 7 and associated tests). This special report will be submitted to the PRC prior to starto .

F. CHANGES IN PLANT OPERATING STAFF Two new operators: Harold Barnett and Tim Demanett.

G. TRAINING Training for operators included General Employee Training, Cycle 8 Modifications, and fire brigade. Maintenance training included crane operators and signalmen, hydrostatic testing and General Employee Training.

During the 1983 Refueling approximately 400 personnel received Initial General Employee Training and approximately 450 home personnel received Annual General Employee Refresher Training.

H. CHANGES, TESTS AND EXPERIMENTS REQUIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT TO 10CFR50.59.

Amendment No. 69 Deletes Appendix B from the Fort Calhoun Technical Specifications on Environmental Monitoring.

Amendment No. 70 Provides Technical Specifications changes required for Cycle 8 operation.

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< Monthly Operation Report March, 1983

Page Twelve l

II. MAINTENANCE (Significant Safety Related)

Refueling Outage Maintenance will be submitted as one package in April and May, 1983.

l T'f p W. G. Gates Manager Fort Calhoun Station t

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