ML20092A250

Monthly Operating Rept for May 1984

ML20092A250
Person / Time
Site:	Fort Calhoun
Issue date:	05/31/1984
From:	William Jones, Matthews T OMAHA PUBLIC POWER DISTRICT
To:	Deyoung R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
References
LIC-84-177, NUDOCS 8406190115
Download: ML20092A250 (15)

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 UNIT Fort Calhoun Statiosi DATE June 13, 1984 COMPLETED BY T. P. Matthews TELEPilONE -(402) 536-4733 MONTH Ny, 1984 DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (MWe-Net)

(MWe-Net)

I 0.0 g7 0,0 2.

0.0 gg o,o 3

0.0 gg o,o 4

0.0 20 0.0 5

0.0 21 0.0 6

0.0 22 0.0 0.0 7

73 0,0 8

0.0 24 0.0 9

0.0 25 0.0 10 0.0 26 0.0 11 0.0 27 00 12 00 28 0.0 13 0.0 29 0.0 14 0.0 30 0.0 15 0.0 3

0,0 16 0.0 INSTRUCTIONS On this format. list the average daily unit power leselin MWe-Net for each day in the reporting anonth. Cornpute to the nearest whole megawatt.

01/77) 8406190115 840531

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PDR ADOCK 05000285 R

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OPERATING DATA REPORT DOCKET NO.

50-285 DATE June 13, 1984 COMPLETED llY T. P. Matthews lELEPilONE (402) 536-4733 OPERATING STATUS

1. Unit Name:

Fort Calhoun Station Notes

2. Reporting Period:

May, 1984 1500

3. Licensed Thermal Power (MWt):

501

4. Nameplate Rating (Gross MWe):

470

5. Design Electrical Rating (Net MWe):

6. Maximum Dependable Capacity (Gross MWe): 461 438

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 Any (Net MWe):

N/A

10. Reasons For Restrictions.If Any:

None

~ This Month Yr. to Date Cumulative

!I. Hours in Reporting Period 744.0 3,647.0 93,649.0

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

13. Reactor Reserve Shutdown Hours 0.0 0.0 1,309.0

14. Hours Generator On-Line 0.0 1,489.5 70,892.0

15. Unit Resene Shutdown Hours 0.0 0.0 0.0

16. Gross Thermal Energy Generated (MWH) 0.0 2,152,796.9 88,912,510.6 0.0 690,258.0 29,007,827.0

17. Gross Electrical Energy Generated (MWH)

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

19. Unit Service Factor 0.0 40.8 75.7 0.0 40.8 75.7

20. Unit Availability Factor 0.0 41.1 64.6

21. Unit Capacity Factor (Using MDC Net) 0.0 37.7 62.3

22. Unit Capacity Factor (Using DER Net).

U.U 0.0 3.5

23. Unit Forced Outage Rate

24. Shutdowns Scheduled Oscr Next 6 Months tT.spe Date,and Duration of E cht:

25. If Shut Down At End Of Report Period. Estimated Date of Startup:

June 24, 1984

26. Units in Test Status (Prior io Commercial Operation p:

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^'hid N/A INITIAL CRITICALITY INITI AL ELECTRICITY COMMERCI A L OPER ATION (W77 i t

m DOCKET NO.

50-285 UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Fbrt Calhoun Station DATE June 13, 1984 COMPLETED BY '

T. P. Matthews REPORT MONTH May, 1984 TELEPHONE (402) 536-4733 n.

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Licensee

,E-r, Cause & Corrective No.

Date g

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Prevent Recurrence H

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o 84-01 840303 S

2157 C

4 N/A XX XXXXX 1984 refueling outage commenced March 3, 1984.

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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 iLF R) File iNUREG-D-Regulatory Restriction 4-Other (Explain) 0161)

E-Operator Training i License Examination F Adnunistrative 5

G-Operational Error (Explain)

Eshibit I - Same Source et/77)

II-Other (Explain)

Refueling Information Ebrt Calhoun - Unit No.1 Report for the month ending tiay 1984 1.

Scheduled date for next refueling shutdown.

September 1985 2.

Scheduled date for restart following refueling.

November 1985 3.

Will refueling or restmption 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 Specification change b.

If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Cbumittee 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 sutmitting proposed licensing action and support information.

Aunust 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.

6.

The number of fuel assemblies: a) in the core 133 assenblies b) in the spent fuel pool 305 c) spent fuel pool storage capacity 7?9 d) planned spent fuel pool storage capacity

• fiay be increased 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 tute June 1, 1984 Prepared by_ _

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'a OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No.1 May, 1984 Monthly Operations Report I.

OPERATIONS

SUMMARY

Fort Calhoun Station was perfonning systen startups for a return to power.

During a hydrostatic leak test of the reactor coolant system on May 16 a stean generator tube ruptured causing a primary to secondary leak. Startup has been delayed while extensive eddy-current testing is perfonned on the stean generator tubes.

Two incidents occurred on May 21. A 4160/480 volt transfonner failed causing a fire in the switchgear room.

Later, a Blair City mower threw a downed wire into the 161 KV transnission line which feeds Fort Calhoun Station causing a temporary partial loss of offsite power.

Three operators are in hot license training. They will take the NRC Reactor Operator exam in early June.

No. safety valve or PORY challenges occurred.

A.

PERFORMANCE CHARACTERISTICS None B.

CHANGES IN OPERATING METHODS None C.

RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS None D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL Procedure Description SP-RC-2-1 Plugging Steam Generator Tubes (RC-2A).

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only provide; for plugging five tubes in "A" stean generator.

Appropriate radiation protection and quality require-ments have been addressed.

y.

Monthly OpQrations Report May, 1984 Page Two D.

CHANES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued)

Procedure Description SP-SGAI-1 Steam Generator Annulus Inspection.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only involved an inspection inside stean generator "A" while the e

plant was in a refueling outage.

i SP-SGAI Stean Generator Annulus Inspections.

This procedure did not constitute an unreviewed safety i

question as defined by 10CFR50.59 as it only involved an inspection inside stean generator "B" while the plant was in a refueling outage.

SP-SGDLT-1 Steam Generator Dye Leak Test.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only provided for looking for a suspected leaking tube in "B" stean generator. The dye which was used met appropriate chenical requirenents.

Pressurization of the stean generator was in accordance with technical specifica-tions and ASME codes.

SP-VA-80 Hydrogen Purge System Test.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it provides for flow measurenent of the hydrogen purge systen, ho technical specifications are involved.

SP-UF6-1 Uranitsn Hexafluoride Storage Cylinder External Visual i

Inspection.

l This procedure did not constitute an unreviewed safety

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question as defined by 10CFR50.59 as it provides for a routine inspection of storage cylinders and area.

SP-DW-2 Salt Cleaning of Anton Bed Resin.

This procedure did not constitute an unreviewed safety l

question as defined by 10CFR50.59 since the procedure was carried out using nonnal plant practices.and procedure was adequate for work involved.

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' Monthly Operations Report May, 1984 Page Three 4

D.

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

Procedure Description SP-SL-RC-2 Steam Generator Tubesheet Sludge Lancing.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only involved renoval of sludge fran the secondary side of stean generator "A" while the plant was in a refueling ou tage.

SP-SL-RC-2 Steam Generator Tubesheet Sludge Lancing.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only involved renoval.of sludge fran the secondary side of stem; generator "B" while the plant was in a refueling outage.

SP-TG-DROP-1 Generator Drop Test.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it is not safety related equipnent.

System Acceptance Committee Packages for May,1934:

Package Description / Analysis EEAR FC-82-12 Power Receptacle.

This modification provided for the installation of two 120V power receptacles in Room 19 and does not affect any safety equipnent.

This modification has no adverse effect on the safety analysis.

EEAR FC-82-137 Reactor Stud Tensioner Repair Air Line Addition.

This modification provided for the installation of a conpressed air line to the reactor stud tensioner/ snubber repair area in Room 69 of the auxiliary building. This modification does not involve safety related equipment so it has no adverse effect on the safety analysis.

o Monthly OpGrati@ns Report May, 1984 Page Four D.~ CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL -

(Continued) l Systen Acceptance Committee Packages for May,1984:

(Continued) f Package Description / Analysis i

EEAR FC-83-121 Gaitronics for Room 23.

j This modification provided for the installation of a l

gaitronics station in Room 23. This modification does

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not involve safety related equipnent so it has no adverse effect on the safety analysis.

i EEAR FC-83-179 Clarifier Desludge Isolation Valve Installation.

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This modification provided for. the installation of an i

isolation valve to the clarifier desludge valves. This I

I modification does not involve safety related equipment so it has no adverse effect on the safety analysis.

J EEAR FC-80-10

, Drain Valve Between SI-169 and SI-170.

l This modification provided for the installation of a drain valve to expedite draining of portions of the shutdown cooling system. This modification has no adverse effect on the safety analysis.

EEAR FC-80-99 Replacanent of FCV-326 Valve.

This modification replaced the existing valve with a superior quality valve. This modification has no adverse effect on the safety analysis.

i EEAR FC-81-172 Access Platfonns for Reactor Coolant Pump Motors RC-3A, RC-38, RC-3C 'and RC-3D.

This modification provided for the fabrication and installation of platfonns to provide maintenance access to RC-3A, RC-38, RC-3C and RC-3D reactor coolant punp j:

motors.

This modification has no adverse effect on the safety analysis.

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t Monthly Operations Report May, 1984 Page Five f

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CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued) 6 Systen Acceptance Committee Packages for May,1984:

(Continued)

Package Description / Analysis EEAR FC-83-159 Steam Generator A Rim Cut No. 8 Support Plate.

(SRDC0 84-32)

This modification provided for renoving the outer rim and lugs of the No. 8 drilled tube support plates in the stean generators.

Such a design change would provide stress relief for the plates and associated tubes such that the steam generator tube denting rate can be decreased.

This modification has been reviewed by the NSSS supplier. This modification has no adverse effect on the safety analysis.

EEAR FC-83-159 Steam Generator B Rim Cut No. 8 Support Plate.

(SRDC0 84-33)

This modification provided for renoving the outer rim and lugs of the No. 8 drilled tube support lplates in the steam generators.

Such a design change would provide stress relief for the plates and associated tubes such that the stean generator tube denting rate can be decreased.

This modification has been reviewed by the NSSS supplier. This modification has no adverse effect on the safety analysis.

EEAR FC-81-99 Installation of Heated Junction Thennocouple (HJTC)

Part 14 Handling Canister Mounting Bracket and Assembly of HJTC Handling Canisters.

This modification provided for the assembly and installation of the HJTC handling canister mounting bracket and handling canisters.

This modification has no adverse effect on the safety analysis.

EEAR FC-82-96 Auxiliary Building Room 4A Ventilation.

This modification provided for additional ventilation duct in Room 4A.

This modification has no adverse effect on the safety analysis.

Monthly Operations RQport May, 1984 Page Six D.

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

Systen Acceptance Committee Packages for May,1984:

(Continued)

Package Description / Analysis EEAR FC-83-51 HCV-1103/1104/1385/1386 Valve Operators Motor Brakes.

This modification was essentially a one-for-one replacement of valve operator motor brakes on HCV-1385 and HCV-1386 and the replacenent of valve operator motor brakes operating coils on HCV-1103 and HCV-1104 in order to enhance reliability. This modification has no adverse effect on the safety analysis.

EEAR FC-78-74 HCV-1041/1042 Control Circuitry.

This modification provided for the addition of a separate control swt tch for HCV-1041 and HCV-1042.

This modification has no adverse effect on the safety analysis.

EEAR FC-83-144 PAL Door Hinges.

This modification provided for the installation and adjustment of Personnel Air Lock (PAL) door hinges.

This modification has no adverse effect on the safety analysis.

EEAR FC-83-69 Hydrogen Purge Filter Test Port.

This modification provided for the installation of a test port en the hydrogen purge filter housing (VA-82) located in Room 59 of the auxiliary building.

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

EEAR FC-82-115 Remove Concrete Interference to FW-5A Suction Piping.

This modification provided for the removal of floor concrete to widen the suction piping trenches to heater drain pumps FW-5A, SB, and SC (located in the basement of the turbine building); and then fabricated and installed a replacement structural steel floor system.

This modification did not affect safety related equipment so it has no effect on the safety analysis.

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Monthly Operations Report May, 1984 Page Seven D.

CHANES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (Continued)

Systen Acceptance Committee Packages for May,1984:

(Continued)

Package Description / Analysis EEAR FC-80-35 PORY Internals.

This modification replaced the existing PORY internals with those of a different seat design to improve leak tightness of the valves.

The operation of the valves was not changed.

This modification has no adverse effect on the safety analysis.

EEAR FC-84-74

. Fuse Protection of Certain Limit Switch Circuits.

This modification enhanced reliability of valve operation during accident conditions by allowing the valve to be operated even though position indication may be lost.

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

EEAR FC-84-004 Upgrade Lfmit Switches in Room 81.

< This modification replaced the existing limit switches with switches of superior quality.

This modification has no adverse effect on the safety analysis.

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-EEAR FC-79-81 Variable Setpoint for PORY Actuation.

3' This modification provides for the design and installation of the variable setpoint for PORY 3

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actuation to ensure Ifmiting conditions are not violated, provides for more flexibility in operating conditions thereby decreasing the chances of unnecessarily lifting the PORV's and retains the high reliability in the RCS overpressurization protection function of the PORV circuitry.

This modification has no adverse effect on the safety analysis.

EEAR FC181-59, Installation of Heated Junction Thermocouple (HJTC)

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Part 13 Probe.

This modification provided for the receipt inspection or preoperational checkout and installation of unirradiated HJTC prebes and the proper makeup and venting of the HJTC Grayloc flange in preparation for operation of the plant.

This modification has no

_I adverse effect on the safety analysis.

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Monthly Operations R; port May, 1984 Page Eight D.

CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUY COMMISSION APPROVAL (Continued)

Systen Acceptance Committee Packages for May,1984:

(Continued)

Package Description / Analysis EEAR FC-83-01 Replace Panel Mounted Indicators with Qualified In struments.

This modification provided for a one-for-one functional replacement for the existing system.

This modification has no adverse effect on the safety analysis.

EEAR FC-83-44 Differential Steam Generator Pressure Trip Module.

This modification provided for the installation of new components and modification of the reactor protective systen to provide the new Asymmetric Stean Generator Transient (ASGT) trip functions.

This modification has no adverse effect on the safety analysis.

DCR 74A-21F Steam Generator Blowdown-Condensate 1" Bypass Valve.

This modification provided for the installation of the new bypass valve to aid the plant operators to fill the piping for the condensate supply to the heater exchanger FW-44.

This modification did not involve safety related equipnent so it has no effect on the safety analysis.

EEAR FC-83-32 0'.'alification of Foxboro Transnitters.

This modification replaced the existing amplifier assenblies in Foxboro transmitters with one-for-one function replacements.

This modification has no adverse effect on the safety analysis.

EEAR FC-83-146 Relocation of PT-105.

This modification relocated Foxboro transmitter PT-105 and renumbered GEMAC transmitter PT-105 and its instrument loop to avoid confusion of two devices with the same tag. This modification has no adverse effect on the safety analysis.

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Monthly Operbtions R; port May,>1984 O Page Nine'#;

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RESULTS (F LEAK RATE TESTS y.,

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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.

CHANES IN PLANT OPERATING STAFF ne G.

TRAlllING 4

During May, general enployee training was increased to support outage requirements. Operations department 4 received Cycle 9 modifications training. NRC license candidates received increased training in preparation of an exam to be administered in June.

Advanced Technologies, Inc., was awarded a contract to support the accredidation process for training required by INPO. Three programs in the operations area are being u, pgraded in 1984.

Seven other prograns are being scheduled for upgrEdvat a later time.

't H.

CHANES,' TESTS AND EXPERIMENTS REQUIRING NUCLEAR REGULATORY COMMISSION

' AUTHORIZATION PURSUANT TO 10CFR50.59

'Ame ment No.

Description E

Amendment = 78 Incorporates administrative changes which achieve agreenent between the Fort Calhoun Technical Specifications and new regulations which became o

e'F effective January 1 1984.- This rule change affects Technical Specifications ^in the area of shift manning AA and the sane -amendment also changes' to whom the Quality Control personnel report.

This change permits Quality Control personnel to report to the Te:hnical Supervisor.

Amendnent 79 This amendment provides an up-to-date identification of the accessibility of safety related systen hydraulic snubbers (Table 2-6A).

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. Month?y Operations ' Report May, 1984 Page Ten II. MAINTENANCE (Significant Safety Related) l A report will be submitted at the end of the refueling outage.

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W. Gary Gate Manager Fort Calhoun Station i

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Omaha PubllC Power District 1623 Harney Omalia. Nebraska 68102 402/536 4000 June 12, 1984 LIC-84-177 Mr. Richard C.

DeYoung, Director Office of Inspection and Enforcement i

U.

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Nuclear Regulatory Commission Washington, D.C.

20555

Reference:

Docke t No. 50-285

Dear Mr. DeYoung:

May Monthly Operating Report

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Please find enclosed ten (10) copies of the May Monthly Operating Report for the Fort Calhoun Station Unit No. 1.

Sincerely, k' @9Pb l

W. C.l Jones Division Manager Production Operations WCJ/TPM:jmm Enclosures cc:

NRC Regional Office l

Office of Management & Program Analysis (2)

Mr.

R.

R.

Mills - Combustion Engineering Mr.

T.

F.

Polk - Westinghouse Nuclear Safety Analysis Center INPO Records Center i

NRC File 45 5124 Employment with tauat Opportunity maleiremale

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