ML20091K156

Monthly Operating Rept for Dec 1991 for Fort Calhoun Station Unit 1

ML20091K156
Person / Time
Site:	Fort Calhoun
Issue date:	12/31/1991
From:	Cavanaugh G, Gates W OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LIC-92-007R, LIC-92-7R, NUDOCS 9201220301
Download: ML20091K156 (8)

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Omaha Public Power District 444 South 16th Stract Mall Omaha, Nebraska CJ3102-2247 402/63G2000 January 14, 1992 LIC-92-007R U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Mail Station P1-137 Washington, DC 20555

Reference:

Docket No. 50-285 Gent!cmen:

SUBJECT:

DecemberMonthlyOperatingReport(MOR)

Enclosed is the December 1991 MOR for Fort Calhoun Station (FCS) Unit No. 1 ss required by FCS Technical Specification Section 5.9.1.

If you should have any questions, please contact me.

Sincerely,

/

W.G. Gates [A Division Ma ager Nuclear Operations WGG/sel Enclosures c: LeBoeuf, Lamb, Leiby & MacRae R. D. Martin, NRC Regional Administrator, Region IV R. P. Mullikin, NRC Senior Resident Inspector D. K. Sent(.ll, Combustion Engineering R. J. Simon, W stinghouse Office of Management & Program Analysis (2)

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, AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 UNIT Fort Calhoun Station DATE January 13, 1992 COMPLETED BY E. R. Cavanaugh TELEPHONE (402)636-2474 MONTH December 1991 DAY -  ! AVERAGE DAILY POWER LEVEL ' DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 487 17 486 2 486 18 486 3' 486- __

19 486 4 486 20 486 5 486 21 486 6 487 22 486 7- 487 23 486 487 24 _, 486 9 487- ~25 486

~10 487 26 486 11- - 487- 27 486 12 487 28 486 11 3 486 29 486 14- 486 30 486 151 486 31 485-116 486

' INSTRUCTIONS On.this form, list the average daily unit power level in MWe-Net for each day-in'the reporting month. Compute-to the nearest whole megawatt.

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, . . . - - . . - - . - - . . -- ~

S' OPERATING DATA REPORT ,

DOCKET NO. 50-285 UNIT Fort Calhoun Station ~

DATE January 13, 1992  !

COMPLETED BY G.-R. Cavanaugh TELEPHONF (402)636-2474 8

=0PERATING STATUS 1.,' Unit Names Fort Calhoun Station Notes i

2. . Reporting Period: December 1991 J 1500-

3. -Licensed Thermal Power (MWt :

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4.: NameplateRating(GrossMWe: -

502

5. DesignElectricalRating(NtT1We): 478 6.- Maximum Dependable Capacity-(Gross MWe): 502

7. MaximumDependableCapacity(NotMWe):. 478

,8. If_ changes _ occur in Capacity Ratings (Item Numbers 3 through 7) Since Last Report, Give Reasons:

NA ,

9. PowerLeveltoWhichRestricted,IfAny(NetMWe):-NA

10. Reasons for Restrictions,_If Any:_NAA '

l This Month Yr-to-Date Cumulative l

11. Hours in Reporting Period ~ 744.0 8,760.0 160,130.0

12. Number of Hours 1 Reactor was Critical -744.0 8,030'T _

124,818.7

13. ' Reactor Reserve' Shutdown Hours 0.0 0.0 1,309.5 114. Hours Generator.0n-Line- 744.0 7,947.2 123,377.1

15, Unit Reserve Shutdown Hours 0.0 0.0 0.0 16,; Gross Thermal Energy _ Generated (MWH) 1,111,552.0 10,340,236.4 161,623,725.7-17.-GrossElectr_icalEnergy: Generated (MWH) 378,520.0 3,426,000.0 53,176,126.2 18.- Net' Electrical Energy Cenerated-(MWH) 361,687.9 3,248,975.1- _ 50,733,751.4

19. Unit.' Service Factor 100.0 -90.7 77.0

20. -Unit i Availability Factor-100.0 90.7 77.0 121. Unit Capacity Factor (Using MDC-Net) 101.7 77.6- ~~

69.0

'22.. Unit Capacity Factor (Using DER Net) 10T.7 77.6 67.1 23.- Unit Forced Outage Rate 0.0 9.3 3.9 124.--Shutdowns.ScheduledOverNext6 Months (Type,Date,andDurationof;Each):

' Refueling Outage scheduled for February 1, 1992, and will last

-approximately three months.

25.. If Shut Down at End of- Report _ Period,_ Estimated Date of Startup: NA

- 26.. Units In Test: Status (Prior to Commercial Operation): Forcast- Achieved INITIAL CRITICALITY INITIAL ELECTRICITY N/A

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COMMERCIAL OPERATION

, Refueling Information Fort Calhoun - Unit No. 1 Report for the month ending December 1991 4

- 1. Scheduled date for next refueling shutdown. February 1. 1992

2. Scheduled date for restart following refueling, aprJl_22u1992

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?

Incorporate specific-requirements resulting from reload safety analysis.

. If answer is no, has the reload fuel design and core configuration been reviewed by your Plant Safety Review Committee to determine whether any unreviewed safety questions are associated with the core reload. N/A

c. If no such resiew has taken place, when is it scheduled? N/A

4. Scheduled date(s) for submitting proposed licensing action and support information. Submitted November 27.

1991

5. Important licensing considerations a::sociated 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. New fuel supplier lipw LOCA Analysis

6. The number of fuel assemblies: a) in the core 133 3 A;semblies b) in the spent fuel pool 477 Assemblies c) spent fuel pool storage capacity 729 Assemblies d) planned spent fuel pool Planned to be increased storage capacity with higher density spent fuel racks.

7. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity. 1995

• o- Capability of full core offload of 133 assemblies lost. Reracking to be performed between the 1993 and 1995 Refueling Outages.

Prepared'by kM'11 b kce Date i,/,s /s 2

UNIT SHUTDOWNS AND P')WER REDUCTIONS DOCKET NO. 50-285- - -

UNIT NMtE Frrt Calhoun Station:

DATE January 9. 1992 COMPLETED BY G. R. Cavanauch -

TELEPHONE 402 - 636-2474 REPORT MONTH Decer.ber 1991 l

2 Licensee System Component Cause & Corrective No2 Date Type' Duration Reason Method of Shuting Event Code' Code' Action to (Hours) Prevent Recurrence Down Reactor 8 Report #

l 'Ibere were no unit shu.Jowns or power reductions in Decernber 1991.

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3 8 I 2 Reason: Metini: Exhibit G - Insturctions F: Forced A-Equipnnnt Failure (Explain) 1-Manual for Preparation of Data S: Scheduled 2-Manual Scram. Entry Sheets for Licensee B-Maintenance or Test C-Refueling 3-Automatic Scram. Event Report (LER) File D-Regulatory Restriction 4-Other (Explain) (NUREG-0161)

E-Operator Training & License Examination '

F-Administrative G-Operational Error (Explain) Exhibit 1 - Same Source (9/77) H-Other (Explain)

i

- . OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No. 1 December 1991 Monthly Operating Report I. OPERATIONS

SUMMARY

Fort Calhoun Station operated at a nominal 100% power throughout the month of December 1991.

On December 1, 1991, the Process Radiation Monitrrs for the Radioactive Waste Processing Building (RM-041,-042, & 043) were found to have zero sample flow due to a closed exhaust damper in the building ventilation system. This required entry-into Technical Specification LC0 2.9.1 (2) which requires - grab samples each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for gaseous activity and continuous sampling for particulate and iodine activity on the Rad Waste Building exhaust stack. The damper apparently failed closed due to the loss -tf the control air in the building. The control air system self-contained air compressor had tripped off and not restarted following a momentary . loss of off-site (161kV) power. The air compressor was restarted and sample flow was re-established to the monitors shortly after discovery. This allowed clearing the LCO.

On-December 4, 1991, a four-hour notification was made to the NRC due to a failure to adequately test the equalizing valve on the containment inner Personnel Air Lock (PAL) door. The e tested per Technical Specification 3.5(qualizing volve had not been leak 3) s measure the inner door was declared inoperable and the outer PAL door was immediately danger tagged closed to prevent a breach of Containment integrity.- ~ Preparations for testing the equalizing valve included fabricating a test assembly and obtaining a waiver of compliance from the NRC to allow the outer PAL door to be opened. On December 6, 1991 the NRC granted Omaha Public Power District (OPPD) a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> waiver to allow testing of the inner PAL door equalizing valve. The valve was ,

~

successfully tested.on December 7, 1991.

On. December 10,1991, the last shipment of new Westinghouse fuel was received and inspected for the upcoming refueling outage. Currently all new fuel assemblies have been accepted by OPPD exccpt for one (R-001).

The assembly in question has a small surface defect on one fuel pin.

On December 16, 1991 a one-hour notification was made to the NRC based on discovery.thatportIonsofthetubingusedtotesttheContainmentPAL were not seismically qualified; thus, the PAL was outside the design bas?

of the plant. The non-qualified test equipment has been removed and new seismically rated tubing and isolation valves will be-installed January

-1992.

Unknown leakage from the RCS had been increasing over the last few months.

During the month of December, efforts to identify the leak path found that the source of leakage was the High Pressure Safety Injection (HPSI) -

Charging system crosstie valve HCV-2988. To prevent any further leakage-from this path, the Charging - HPSI manual isolation valve (CH-li,1) was closed. This resulted in a reduction in RCS unknown leakage to normal values.

1 MonthlyOperatingReport(Continued)

LIC-92-007R Page 2 On December 19, 1991, the 161kV Offsite power feed to the plant was taken out of service to repair a power pole arm which had been damaged by a winter storm earlier in the year. Repairs were completed by OPPD's Electric Operations Division and power was restored in approximately six hours.

The following hRC inspections took place during December 1991:

IER No. 91-24 Routine Resident Inspection IER No. 91-26 Special Inspection (Containment Integrity)

The following LERs were submitted during December 1991:

jJR NO. Descrintion 91-24 ESF Actuation after Pulling Fuses 91-25 S1 Pipe Supports Outside Design Basis 91-26 SR0 without NRC Physical 91-27 RCDT Sampling (WD-1060) 91-28 Unmonitored Release on Loss of 161kV A. SAFETY VALVES OR PORV CHALLENGES OR FAILURES WHICH OCCURRED NONE B. RESULTS OF LEAK RATE TESTS Total RCS leak rate increased from 0.475 gpm at the first of the month to 0.712 gpm on December 18, when the major source of leakage was identified. Charging flow was leaking through HCV-2988 to the 2500 psig HPSI header and then through the "B" Safety Injection Tank leakage cooler to the Reactor Coolant Drain Tank. HCV-2988 is the cross connect valve between the charging header and the high pressure HPSI header. The valve is a solenoid operated valve that is normally open to provide a flow path for hot leg injection and long term core cool ing. The valve is scheduled to be rsplaced during the 1992 refueling outage.

HCV-2988 was isolated by closing CH-191. This reduced the leakage to .137 gpm. With CH-191 closed, the total RCS leakage remained near this amount the rest of the month.

9 MonthlyOperatingReport(Continued)

LIC-92-007R Pagc 3 C. CHANGES, TESTS AND EXPERIMENTS REQUIRING NUCLEAR REGULA10RY COMMISSION AUTHORIZATION PURSUANT TO 10CFR50.59 Amendment No. Description NONE-D. SIGNIFICANT SAFETY RELATED MAINTENANCE FOR THE MONTH OF DECEMBER 1991

- Adjusted impeller lift and obtained new baseline data on three RawWaterPumps(AC-10A,B,C)

-Charging Replacedthegasbag)assemblyontheSuctionAccumulatorfor Pump (CH-18 .

Straightened the top cap and installed a new diaphragm on the fresh air supply damper (YCV-8710) for Diasel Generator No. 2.

- On 125V DC Battery Charger No.1, the silicon controlled rectifier (SCR) was found to be degraded and a replacement was I installed. Also replaced a faulty amplifier module and re-soldered a loose connection on the SCR firing module.

Installed the new Breaker for the Vacuum Deaerator Pump

. (DW-46A)

I Replaced ferro-resonance choke and filter components for Inverter"C"BypassTransformer(EE-4Q) .

t Replaced.a defective coil for the Containment High Radiation )

- SignalLockoutRelayChannelB(86B/CRHS)

Specific . tubes for Component Cooling Water Heat Exchanger AC-10 were cleaned to execute performance testing of various individual tubes, .

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