ML20080N010

Monthly Operating Rept for Jan 1984

ML20080N010
Person / Time
Site:	Fort Calhoun
Issue date:	01/31/1984
From:	William Jones, Matthews T OMAHA PUBLIC POWER DISTRICT
To:	Deyoung R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
References
LIC-84-049, LIC-84-49, NUDOCS 8402220061
Download: ML20080N010 (13)

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AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 UNIT Fort Calhoun Station 947g February 7, 1984 COMPLETED HY T. P. Matthews TELEPHONE (402) 536-4733 MONT11 January, 1984 DAY AVERAGE DAILY POWER LEVEL DAY AVER AGL DAILY POWER LEVEL (MWe Net) iMWe Neti

, 456.2 g7 454.6 2 456.4 1g 454.5 3 455.9 19 454.6 4 458.5 20 454.5 5 458.4 21 454.7 _ _ _

6 457.8 22 454.6 7 456.4 23 _

455.2 8 454.4 28 455.0 9 454.7 25 454.9 10 455.4 26 455.1 ll 455.7 27 455.0 12 455.9 28 454.7 13 455.2 29 454.3 13 454.5 30 454.6 15 454.8 3: 454.7 16 454.9 INSTRL'CTIONS On thi: fortnat. list the average daily unit power leselin MWe Net for each day in ibe reportmg mienth. Onnpute to the nearest whole megawatt.

(9/77 )

8402220061 840131 PDR ADOCK 05000285 R PDR

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4 OPERATING DATA REPORI DOCKETNO. 50-285 DATE February 7, 1984 COMPLETED 11Y T. P. Matthews TELEPilONL (402) 5T6.4733 OPER ATING STATUS Fort Calhoun Station Notes

1. Unit Name:

2. Reporting Period: January, 1984 1500

3. Licensed Thenna! Power (MWt):

4. Nameplate Rating (Cross MWe): 501

5. Design Electrical Rating (Net MWe): 478 461

6. M:ximum Dependable Capacity (Gross Mhe):

7. Maximum Dependable Capaeny (Net MWe): 438

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

N/A

9. Power Level To which Restr:cted.If Any INet MWe): N/A _

10. Reasons For Restrictions. If Any: NONE This Month Yr. to Date Cumulatise i1. Hours in Reporting Period 744 744 90,746.0 744 744 70,637.9

12. Number Of Hours Reactor Was Critical 0.0 0.0 1,309.0

13. Reactor Reserve 5hutdown llours

14. Hours Generator On-Line 744 744 70,146.5 0.0 0.0 0.0

15. Unit Resene Shutdown Hours 1,105,887.8 1,105,887.8 _. 87,865,601.5

16. Gross Ttennal Energy Generated (MWil)

17. Gross Electrical Energy Generated (MWH) 355,5_62.0 355,562.0 28,673,131.0

18. Net Electrical Energy Generated (MWH) 338.780.8 338.780.0 27,418.648.7

19. Unit Senice Factor 100.0 100.0 77.3

20. Unit Availability Factor 100.0 100.0 77.3

21. Unit Capacity I' actor IUsing MDC Net) 104.0 104.0 65.8

22. Unit Capacity Factor (Using DER Net) 95.0 95.0 63.5

23. Unit Forced Outage Rate 0.0 0.0 3.6

24. Shutdowns Scheduled Oser Next 6 Months IType. Ilate,and Duration of Each t:

1984 refueling outage scheduled to start around March 3, 1984.

25. If Shut Down At End Of Report Period. Estimated Date of Startup: N/A 26 Units In Test Staius IPrior o Commercial OperationI: 1orecast Achiesed INITIA L CHITICALITY INITIA L ELECTRICITY COW 12RCI AL OPER ATION

('s/77 )

DOCKET NO. 50-285 q

. UNIT SHUTDOWNS AND POWER REDUCTIONS

_ , , UNIT NAME Fort Calhoun Station DATE Februarv 7, 1984 REPORT MONTH January, 1984 COMPLETED BY T. P. Matthews TELEPHONE (402) 536- 4733 n.

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-, .5 ? 3j h .Y 5 5 icensee ,Ev, Cause & Corrective No. Date g. 3g ,g g 5 Event p  %'8 gO Action to H

$5 5 j g5 g Repor = mO g Prevent Recurrent:e 6

No unit shut downs during the enth of January,1984.

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Forced Reason: Method: Exhibit G-Instructions j S: Scheduled A Equipment Failu.e(Explain) l Manual for Preparation of Data B-Maintenance of Test 2-Manual Scram. Entry Sheets for Licensee C-Refueling 3.Autonutic Scram. Event Report (LE R) File (NUREL.

D Regulatory Restriction 4-Other (Explain) 0161)

E-Operator Training & License Examination F-Administrative 5 G-Operational Eirur (Explain) Eshibit I - Same Source (9/77) 11-O her(Expliin) i

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• Refueling Information 2

Ebrt Calhoun - Unit No.1 Report for the nonth ending Januarv_1984 .

1. Scheduled date for next refueling shutdown. March 1CRd

2. Scheduled date for restart following refueling. _May 19g4

, 3. Will refueling r 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 action and support information. Tech. Soecs.- February '84

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. 'Ihe number of fuel assemblies: a) in the core 133 assemblies b) in the spent fuel pool 265 c) spent fuel pool storage capacity 483 p d) planned spent fuel pool storage capacity 728

7. - The projected date of the last refueling that can be discharged to the spent fuel pcol assuming the present licensed capacity. 1985 Prepared by a,,4n/d Date Februarv 1 1 oo M.

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OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No.1 January,1984 Monthly Operations Report I. OPERATIONS

SUMMARY

Fort Calhoun Station operated at a nominal 100% power throughout the month of January,1984, continuing its record on-line run. New fuel for Cycle 9 core load began arriving on site January 26, 1984. The installation of the new spent fuel racks was completed during January. Preparations for the 1984 refueling outage continue.

No safety valve or PORV challenges occurred.

A. PERFORMANCE CHARACTERISTICS LER Number Deficiency 83-004, Rev. 1 During perfonnance of surveillance test ST-ESF-3, F.2,

" Containment Pressure Channel Check", pressure switches A/PC-742-1 and A/PC-742-2 were found to be initiating /

actuating above the Tech. Spec. Ifmit of 5 psig.

During the time pressure switches A/PC-742-1 and A/PC-742-2 were considered inoperable, the remaining pressure switches feeding the B, C and D channels of the "A" and "B" Containment Pressure High Signal initiation taatrices were operable, available and fully capable of initiating protective actions required to mitigate the consequences of an accident. This revision depicts changes made to the Corrective Action section (Attachment 2) and Failure Data se "on (A*.tachment 3) of the original LER submittai. .83-013 During perfonnance of surveillance test ST-ESF-3, F.2,

" Containment Pressure Channel Check", pressure switches A/PC-742-1 and A/PC-742-2 were found to be initiating /

actuating above the Tech. Spec. limit of 5 psig.

During the time pressure switches A/PC-742-1 and A/PC-742-2 were considered inoperable, the remaining i pressure switches feeding the B, D and D channels of the "A" and "B" Containment Pressure High Signal (CPHS) initiation matrices were operable, available and fully capable of initiating designed protective acti~ons required to mitigate the consequences of an accident.

B. CHANGES IN OPERATING METHODS None a

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- Monthly Operations Report Page Two January,1984 C. RESULTS OF SURVEILLANCE TESTS AND INSPECTIONS Mone D. CHANGES, TESTS AND EXPERIMEH" CARRIED OUT WITHOUT COMMISSI0'l APPROVAL Procedure _

Description SP-FAUD-1 Fuel Assembly Uplift Condition Detection. i 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 fual assembly uplift condition did not exist.

SP-ATCOR-1 Compare Electrical Drawings with As Built Condition.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 since it only involved work on a non-safety related system.

SP-CAPSULE-1 Separation of Surveillance Capsule Wall Assembly.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only provided for disassembly of a surveillance capsule once it was removed from the reactor vessel.

SP-CAPSULE-2 Reactor Yessel Surveillance Capsule Installation.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only provided for the safe and orderly insertion of a surveillance capsule assembly into the reactor vessel during refueling operation. The use of surveillance capsules to monitor the effect neutron flux has on the reactor vessel wall is a Technical Specification requirement.

SP-CTPC-1 Core Thermal Power Calculation.

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

consisted of recording data from existing plant i nstrumentation.

Mdnthly Operations Report Page Three January,1984 D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (continued)

Procedure Description SP-ECT-2 Eddy Current Testing of Shutdown Cooling Heat Exchange Tubes.

This procedure did not constitute an unreviewed safety  !-

question as defined by 10CFR50.59 as it only provided for eddy current testing of tubes in a shutdown cooling heat exchanger. The test was performed during refueling shutdown when the reactor core was off loaded and the shutdown cooling heat exchanger was not required to be ir service.

SP-FE-10 Inspection of Spent Fuel Using CE Curb Mounted Equipment.

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This procedure did not constitute an unreviewed safety question _as defined by 10CFR50.59 since all fuel handling was in compliance with approved procedures and Technical Specification limitations. The existing safety analysis applied to this work.

SP-GCASK-1 Shipment of CNSI Cask.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because this procedure only provides for loading a cask with a surveillance capsule. The cask was handled with a single failure proof crane and appropria'a radiation protection measures were utilized.

SP-GCASK-1 Shipment of CNSI Cask.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because this procedure only provides for shipment of radioactive material in an approved cask.

SP-NI-6 Nuclear Instrumentation Decalibration end Heat Rate Optimization Test.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because all data was obtained during typical plant operating conditions covered by accident analyses and Technical Specifications.

Monthly OpGrations Report Page Four-January,1984 D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL

'(continued)

Procedure Description SP-PRCPT-1 Post Refueling Core Physics Testing.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because all testing was perfonned in accordance with Technical Specifications and was consistent with startup during previous cycles.

SP-RCS-LE-1 Visual Inspection of Class I Piping in 10 Year ISI.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 since it only provided guidelines for perfonning a visual leak examination of RCS piping hydrostatically tested under a surveillance test.

SP-RRC-3 Reactivity Computer and Reactor Physics Constants Adjustments.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because the procedure simply consisted of recording data from existing plant instrumentation.

SP-SHAP-1 Shape Annealing Factor Verification.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 because all reactor power and axial shape maneuvers were performed within limits described by the Station's Technical Specifications and safety analysis report.

SP-UF6-2 Off Loading of UF6 Containers in Storage Areas.

This procedure did not constitute an unreviewed safety question as defined by 10CFR50.59 as it only provided a safe and orderly procedure for off-loading UF6 storage cylinders into the UF6 storage area which has been approved by the NRC.

- Mo'nthly Operations Report Page Five January,1984 D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (continued)

Package _ Description / Analysis SP-WDS-16 Spent Resin Disposal.

This procedure did not constitute an unreviewed safety question as defined by 10CFRSO.59 as it only provided an alternate method for sampling a waste gas decay tank. Technical Specifications dealing with radioactive gas release have been addressed.

System Acceptance Committee Packages for January,1984:

Package Description / Analysis DCR 74A-27 Main Steam Isolation Valves.

This modification completed a manufactuer's request to modify the main steam isolation valves to prevent stress created upon power closure. This modification has no adverse ef fect on the safety analysis.

DCR 74A-78 Bantam Crana Seismic Tiedown.

This modification installed hold downs for the bantam crane on top of the containment. This modification has ne adverse effect on the safety analysis.

DCR 77-11 Request for a Bathroo.n in Warehouse.

This modification installed a bathroom in the warehouse and is not safety related. This modification has no adverse effect on the safety analysis.

DCR 77-106 Installation of FM Radio System.

This modification installed an emergency FM radio communications system. All equipment is powered by convenience outlets and is not safety related. This modification has no adverse effect on the safety analysis.

EEAR FC-83-95 Circuit Noise on LC-101X and LC-101Y.

This modification reduced circuit noise only. This modification has no adverse effect on the safety analysis.

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. Monthly Operations R: port Page Six January,1984 D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (continued)

Package Description / Analysis EEAR FC-83-98 PIC-1538 Contact Rearrangement.

This modification involved a high pressure alam indicator and rearranging the contacts did not change its operation. This modification has no adverse effect on the safety analysis.

EEAR FC-83-102 Access Platfoms for Inlet Cells to Intake Structure.

This modification provided access platforms for inlet cells to the intake structure. This modification has no adverse effect on the safety analysis <

DCR 74A-29 Automatic Gas Analyzer.

This modification installed an automatic waste gas analyzer that is no longer used. This modification has no adverse effect on the safety analysis.

OCR 748-01 FP.C-269X Modification.

This modification changed a response based on a square root to a linear response and no operational aspects were changed. This modification has no adverse effect on the safety analysis.

DCR 75A-36 SIRWT Level RAS Switches.

This modification improved the repeatability of the SIRWT RAS level swit;hes. This modification has no adverse effect on the safety analysis.

DCR 76-105 Steam Generator Blowdown Flow Transmitters.

This modification installed blowdown flow transmitters which improved the control system due to better indication of the flow through the system. Thi s modification has no adverse effect on the safety analysi s.

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- Monthly Operations Report Page Seven January,1984 D. CHANGES, TESTS AND EXPERIMENTS CARRIED OUT WITHOUT COMMISSION APPROVAL (continued)

Package Description / Analysis DCR 78-36 Noise Spikes in Plant Instrumentation.

This modification only installed noise spike  !'

suppressing devices across the relay coils of specific val ves. Operation of the valves or associated system was not effected. This modification has no adverse effect on the safety analysis.

EEAR FC-79-111 Instrument and Control Shop Modification.

This modification expanded the Instrument and Control Shop only and is not safety related. This modification has no adverse effect on the safety analysis.

EEAR FC-80-111 Waste Holdup Tank Instrument Line.

This modification is intended to provide a recircula-tion path so that the gas space may be circulated through a charcoal filter. This would provide a method of removing noble gases from the gas space so the tank could be opened to the atmosphere for inspection. This modification has no adverse effect on the safety analysis.

EEAR FC-80-138 Storeroom Power Cable.

This modification rerouted a power cable in the storeroom in conduit. This modification has no adverse effect on the safety analysis.

E. RESULTS OF LEAK RATE TESTS Procedure Results ST-CONT-2, F.2 PAL 60 psig test - 3800 cc's (700 cc's greater than previous test).

F. CHANGES IN PLANT OPERATING STAFF None

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. . . . j Monthly Oparations R: port i Page Eight January,1984 G. TRAINING Training for January was conducted as scheduled in the areas of operator requalification (licensed and non-licensed operators), fire brigade, maintenance, crane operator and the emergency plan. General employee initial and requalification training program schedule was increased to accommodate the additional manpower required for the upcoming Fort Calhoun Station refueling outage. Operator training was conducted on the ERF computer being installed in the control room and Technical Support Center.

H. CHANGES, TESTS AND EXPERIliENTS REQUIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT TO 10CFR50.59 Package Description Amendment No. 76 The amendment incorporates administrative changes which correct terminology in the basis section concerning pressurizer operability, clarify the basis section for the diesel generator fuel oil inventory, clarify the basis section of shock suppressors (snubbers) specifications, clarify the scope of the inservice inspection program, correct 4

references to DNB parameters and environmental sampling data, remove reference to an offsite organization figure which was deleted in a prior amendment, changes the title of a Safety Audit and Review Committee member, and changes the titles of other OPPD support staff members. The amendment also increases the audit frequency of the Emergency Plan, Site Security Plan, and Safeguards Contingency Plan from at least once per two years to at least once every twelve months.

II. MAINTENANCE (Significant Safety Related)

None l

W. Gary Gates lianager Fort Calhoun Station i _

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Omaha Pubilt Power District 1623 Harney Omaha, Nebraska 68102 402/536 4000 February 14, 1984 LIC-84-049 Mr. Richard C. DeYoung, Director Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Reference:

Docket No. 50-285

Dear Mr. DeYoung:

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

Sincerely, W. C.

'{ones 0 Divisign Manager Production Operations WCJ/TPM:jmm Enclosures cc: NRC Regional Office Office of Management & Program Analysis (2)

Mr. R. R. Mills - Combustion Engineering Mr. T. F. Polk - Westinghouse Nuclear Safety Analysis Center INPO Records Center NRC File l

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