ML20084M094

Monthly Operating Rept for Apr 1984

ML20084M094
Person / Time
Site:	Fort Calhoun
Issue date:	04/30/1984
From:	William Jones, Matthews T OMAHA PUBLIC POWER DISTRICT
To:	Deyoung R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
References
LIC-84-143, NUDOCS 8405150210
Download: ML20084M094 (11)

Text

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 UNIT Fort Calhoun Stag DATE NY 8 r D84 COMPLETED liY T. P. Matthews TELEPilONE (402) 536-4733 MONTH April, 1984 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)

0.0 g7 0.0 2 0.0 33 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 t:

10 0.0 26 0.0 3- 0.0 27 0.0 12 0.0 ;g 0.0 13 0.0 29 0.0 14 0.0 30 0.0 15 0.0 3, 16 0.0 INSTRUCT 10NS On this format, list the average daily unit power level in MWe Net for each day in the reporting month. Compute to the ricarest whole megawatt.

(9/77 )

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PDR ADOCK 05000285 R J) 'l PDR II

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OPERATING DATA REPORT DOCKET NO. 50-285 DATE May 8, 1984 COMPLETED BY __T. P. @ tthews TELEPilONE (402) 536-4733 OPERATING STATUS Fort Calhoun Station Notes

1. Unit Name:

2. Reporting Period: April, 1984 1500

3. Licensed Thermal Power (MWt):

4. Nameplate Rating (Gross MWe):

501 4#

5. Design El ctrical Rating (Net MWe):

6. Maximum Dependable Capacity (Gross MWe): 461

7. Maximum Dependable Capacity (Net MWe):

438

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 Any(Net MWe): N/A

10. Reasons For Restrictions,if Any: None W

This Month Yr to.Date Cumulative

!I, Hours in Reporting Period 719.0 2,903.0 92,905.0

12. Number Of Hours Reactor Was Critical 0.0 1,490.2 71,384.1

13. Reactor Reserve Shutdown flours 0.0 0.0 1,309.0

14. Hours Generator On.Line 0.0 1.489 L 70,892'.0

15. Unit Reserve Shutdown Hours 0.0 -0.0 0.0

16. Gross Thermal Energy Generated (MWH) 0.0 2,152,796.9 88,912,510.6

17. Gross Electrical Energy Generated (MWil) 0.0 690,258.0 29,007,827.0

18. Net Electrical Energy Generated (MWil) 0.0 656,536.5 27,736,405.2

19. Unit Service Factor 0.0 51.3 76.3

20. Unit Availability Factor 0.0 51.3 76.3

21. Unit Capacity FactorIUsing MDC Net) 0.0 51.6 65.1

22. Unit Capacity Factor (Using DER Net) 0.0 47.3 62.8

23. Unit Forced Outage Rate 0.0 0.0 3.5

24. Shutdowns Scheduled Oser Nest 6 MonthslType Date,and Duration of Eacht:

N/A

25. If Shut Down At End Of Report Period, Estimated Date of Startup: May 20, 1984

26. Uniss in Test StatusiPrior to Commercial Operation): N/A Forecast Achiesed INITIA L CRITICALITY INITIAL ELECTRICITY COMMERCIA L OPER ATION (9/77 )

+ .-

DOCKET NO.

50-285 UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Fort Ca:houn Statidn DATE May 8,:.984 COMPLETED BY T. P. Matthews REPORT MONTH 1rril - 19R4 TELEPHONE (402) 536-4733 l

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f .5 X Licensee E t, =L . Cause & Corrective N. Date g 3g g .3 g 5 Event u7 S."? Action to

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Report a rN O !d u

Prevent Recurrence 6

84-01 840303 S 1413 C 4 N/A XX X)D00C 1984 refueling outage camenced March 3, 1984.

4 I  : 3 Method: Exhibit G-Instructions F: Forced Reason:

for Preparation of Dara S: Scheduled A-Equipment Failure (Explain) I Manual 2-Manual Scram. Entry Sheets for Licensee B-Maintenance of Test Event Report (LER) File (NUREG-C-Refueling 3-Automatic Scram.

4-Other (Explain) 0161)

D-Regulatory Restriction E-Operator Training & License Examination 5

F-Administrative Exhibit 1 - Same Source G-Operational Error (Explain)

Pil77) Ilother (Explain)

Refueling Information

" Fort Calhoun - Unit No.1 -

Report for the nonth ending April 1984 .

~ 1. Scheduled date for next refueling shutdcran. Seotember 1985

2. ' Scheduled date for restart following refueling. November' 1985

3. Will refueling or resumption of operation thereafter -

require a technical specification change or other Yes license amendment?

a. If answer is yes, what, in general, will these be?

' Technical Specification change

b. If answer is no, has the reload fuel design ard 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 licensing action and support information. August 1985

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

6._ The number of fuel assemblies: a) in the core 133 assemblis b) in the spent fuel pool 305 c) spent fuel pool "

storage capacity 729 d) planned spent fuel pool "

storage capacity *

• may be increased via fuel pin consolation

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 Date May 1, 1984 g e

OMAHA PUBLIC POWER DISTRICT

' Fort Calhoun Station Unit No.1 April , 1984 Monthly Operations Report I. OPERATIONS

SUMMARY

Fort Calhoun has continued to perform refueling operations during April.

Major operations completed during April were:

Rim cut of the steam generator support plate; Refueling of the reactor; Steam generator testing and tube plugging; Reactor coolant pump maintenance; and Tensioning the reactor vessel head.

The preparation for startup continued as equipment and systems become available.

No safety valve or PORY challenges occurred.

A. PERFORMANCE CHARACTERISTICS LER Number Deficiency 84-002 Failure of Main Steam Safety Valve to Lift Within

'Setpoint Tolerance.84-003 Partial Loss of D.C. Power.84-004 Reactor Coolant Dose Equivalent Iodine Limit Exceeded.

B. CHANGES IN OPERATING METHODS None C. RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS None

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. Monthly Operations Rrport .

April , 1984 ' . .

Page Two D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL ,

Procedure Description SP-RC-2-1 Plugging Steam Generator Tubes (RC-28). .

This procedure did not constitute an unreviewed safety i

! question as defined by 10CFR50.59 as it involved plugging of steam generator tubes in accordance with 3

the method recommended by Combustion Engineering.

SP-SGHE-1 Pressurization Procedure for Helium Leak Testing of Steam Generator "B".

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because it only  !

provided for pressurizing a steam generator for helium .l in preparation for checking for leaking tubes. r Pressure was properly restricted in accordance with the  !

temperature of the steam generator.

SP-SGLT-1 Steam Generator Leak Test. I This procedure did not constitute an unreviewed safety  !

question as defined by 10CFR50.59 because it only 1 provided for checking for leaks in a steam generator.

No plant components were modified.  ;

System Acceptance Committee Packages for April,1984: l I- Package Description / Analysis [

h EEAR FC-83-134 ' Security Camera Calibration Window.

t i This modification provided for the installation of a  !

window in tha Electrical Shop. This modification does not involve safety related equipment so it has no  !

adverse effect on the safety analysis.

EEAR FC-83-101 Screen Wash Header Pressure Switch.

This modification replaced the temporary jumper on the screen wash header pressure switch with pemanent (

wiring. This modification does not involve safety related equipment so it has no ' adverse effect on the i safety analysis.  !

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April, 1984 i Page Three D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued) ,

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-System Acceptance Committee Packages for April,1984: (Continued)

Package Description / Analysis EEAR FC-78-43 Analysis of Failure Mode of YCV-1045. ,

) This modification changed the failure mode to open in ,

response to I.E. Bulletin 79-068, thus increasing the reliability of the steam driven auxiliary feedwater i pump. This modification has no adverse effect on the 6 safety analysis.

EEAR FC-79-147 -Trip Circuit Bypass for Auxiliary Building Fans. l This modification installed an override switch to allow  !

j the Auxiliary Building ventilation to continue should a -

spurious trip of these fans occur (due to the ,

freezestat or fire detection control circuitry). This  :

' modification has no adverse effect on the safety f analysis. i EEAR FC-81-08 Replacement of Unqualified Limit Switches on Safety e Related Valves in Auxiliary Building. l t

This modification replaced existing limit switches with i neclear grade limit switches. This modification has no adverse effect on' the safety analysis. i EEAR FC-79-152 Chemistry Lab Filters.  ;

l

This modification did not effect a safety related

, system, therefore, has no adverse effect on the safety  !

analysis. j EEAR FC-82-071 Steam Cenerator Blowdown Isolation Yalve Auto-Closure l l for "No Sample Flow" Conditions.  !

This modification improved the reliability of- the [

radiation monitors to valve off the steam generator  ;

, blowdown upon detection of high radiation. By auto-matically isolating steam generator blowdown upon detection of 'no flow' in the sample lines, there is i less chance of having unmonitored steam generator blowdown. This modification has no adverse effect on l 4 the safety analysis. l I

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Monthly Operations Report April, 1984 Page Four D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued)

System Acceptance Committee Packages for April,1984: (Continued)

Package Description / Analysis EEAR FC-83-105 Reactor Coolant Pump Gasket Leak-Off Lines.

This modification will allow better control and collection of the fluid that collects in the reactor coolant pump gasket space. This modification has no adverse effect on the safety analysis.

EEAR FC-83-11 Laundry Room Air Conditioning.

This modification did not involve a safety related system, therefore, has no adverse effect on the safety analysis.

' EEAR FC-83-150 Reroute.MSIY Leakoff Lines.

This modification rerouted the leakoffs to not only_ a  ;

non-pressurized disposal point but also to a point presently being monitored for radiation should a leak occur. Also, no phase of construction impacted any CQE equipment. This modification has no adverse effect on .

the safety analysis.  ;

EEAR FC-82-168 FH-12 Position Indication.

This modification simply replaced the existing position  ;

indication system and has no adverse effect on the safety analysis.

EEAR FC-81-99, Installation of MI Seismic Supports.

! Part 4 This modification did not effect any adjacent CQE equipment or wiring and has no adverse effect on the "

safety analysis.

- EEAR FC-82-95 Offsite Radwaste Processing.

This modification installed three tie-ins to the [

radioactive waste system. These valves will be closed with the open end blank flanged. This modification has no adverse effect on the safety analysis.

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Monthly Operaticns Report April, 1984 Page Five D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued)

System Acceptance Committee Packages for April,1984: (Continued)

Package Description / Analysis EEAR FC-83-002, FH-12 Variable Speed Drive Controller Replacement.

Part 1 This modification replaced the existing DC drive controllers with more reliable, better quality DC ,

controllers. This modification has no adverse effect on the safety analysis.

EEAR FC-83-17 Working Ledge for Equipment Hatch.

This modification did not effect safety related ,

equipment, therefore, has no adverse effect on the -

safety analysis.

EEAR FC-83-54 Solenoid Yalves for HCV-344 and 345.

This modification replaced sheet metal solenoid supports with CQE seismic supports and has no adverse effect on the safety analysis. ,

EEAR FC-83-61 Pipe Support for Turbine Plant Cooling Water Line.

This modification will stop excessive vibration on the turbine plant cooling water lines to the heater drain pumps. This modification does not involve safety related equipment so it has no adverse effect on the safety analysis.

EEAR FC-83-73 Containment Sump Pump Local Control.

This modification removed the nonfunctional local control switches for containment sump pumps. This

• modification has no adverse effect on the safety l- analysis.

EEAR FC-82-119 RC-3A Seal Removal Hazard.

l This modification only adds a valve and a flange to an existing system. This modification has no adverse effect on the safety analysis.

. . - - - - - - , .,n

Monthly Operations R; port April, 1984 ,,

Page Six

0. . CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued)

System Acceptance Committee Packages for April,1984: (Continued)

--Package Description / Analysis _

EEAR FC-83-104 Qualification of Splices on CCW, HPSI, LPSI and CS Pumps.

This modification is a one-for-one replacement of existing materials with a qualified equivalent. This modification has no adverse effect on the safety analysis. '

E. P.ESULTS OF LEAK RATE TESTS The Fort Calhoun Station is currently performing B and C penetration tests. A report will be sent out at the end of the refueling outage.

F. CHANGES IN PLANT OPERATING STAFF None G. TRAINING Training in April for Operations consisted of modifications that were installed during refueling and the associated procedure changes. Initial Operations training for two new Operators was also conducted. General Employee Training was conducted numerous times to support outage manpower requirements. Special training was conducted for crane signalmen, hydro l

testing and steam generator rim cut.

i CHANGES, TESTS AND EXPERIMENTS REQUIRING NUCLEAR REGULATORY COMMISSION

(

H.

AUTHORIZATION PURSUANT TO 10CFR50.59 Amendment No. Description Amendment 77 Revises Technical Specifications to support unit .

l operation. at full power during Cycle 9. Also revises Reactor Coolant System pressure-temperature limits for operation to 8.5 effective full power years. Also deletes the spent fuel inspection requirements.

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Monthly Op2raticns Report

' April, 1984 ,,

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'Page Seven II. MAINTENANCE (Significant Safety Related) ,

A report will be submitted at the end of the refueling outage. ,

.T. ?O-W. Gary Gates Manager Fort Calhoun Station L

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Omaha Pub!!c 310wer Disuict 1623 Hamey Omaha, Nebroska 68102 l 4021136 4000 l 3

, May 8, 1984 s

LIC-84-143 ,

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Mr. Richard C. D e 'l o u n g , Director Office of Inc."ection and Enforcement U. S. Nuclear Reguittocf Commission Washington, D.C. 23555

Reference:

Docket No. 50-235

Dear Mr. DeYoung:

' April Monthly Operating Report Please find enclosed ten (10) copies of the April Monthly Oper-ating Report for the Fort Calhoun Station Unit No. 1.

\

sincerely, L , {d;f N. C. Jones Division Mana<Jer Production Operations WCJ/TPM:jmm

• Enclosures cc: NRC Regionnl Of fice Office of Management & Program Analysis (2)

Mr. R. R. Mills - Cordoustion Engineering '

Mr. T. F. Polk - Westinghouse s Nuclear So!ety Analysid Center INPO Records Center .

NRC File s

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45 5124 Employment with Equd/doporturt ty Male / Female

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