ML20096D250

Monthly Operating Rept for Apr 1992 for Fort Calhoun Station,Unit 1

ML20096D250
Person / Time
Site:	Fort Calhoun
Issue date:	04/30/1992
From:	Cavanaugh G, Gates W OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LIC-92-179R, NUDOCS 9205150170
Download: ML20096D250 (9)

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Omaha Pub'ic Power District 444 South 16th Street Mall Omaha, Nebr,.ska 68102-2247 402/636-2000 May 14, 1992 LIC-92-179R (hI m 5 9fclear Regulatory Commission i ATTb: ament Control Desk y I' e dion Pl-137

.p M u gton, DC 20555 Y _:. >

Reference:

Docket No. 50-285 Gentlemen:

SUBJECT:

April 1992 Monthly Operating Report (MOR)

Enclosed is the April 1992 MOR for Fort Calhoun Station (FCS) Unit No. I as required by FCS Technical Specification Section 5.9.1.

If you should have any questions, please contact me.

Sincerely,

, /f&*. N . h

• W. G. Gates Division Manager Nuclear Operations WGG/sel Er.closvees c: LeBoeuf, Lamb, Leiby & MacRae R. D. Martin, NRC Regional Administrator, Region IV R. P. Hullikin, NRC Senior Resident Inspector D. L. Wigginton, NRC Senior Project Manager S. D. Bloom, NRC Project Engineer R. T. Pearce, Combustion Engineering R. J. Simon, Westinghouse Office of Management & Program Analysis (2)

IHP0 Records Center American Nuclear Insurers 9205150170 920430 q PDR ADOCK 05000285 R PDR /j } t l tfQQf F e

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. OPERATING DATA REPORT DOCKET NO. 50-285 UNIT FORT CALHOUN STATION DATE MXi 11 1992 COMPLETED BY G. R. CAVATXUGH OPERATING STATUS TELEPHONE TT07)T3TZid74 l

1.-Unit Name: FORT CALHOUN STATION

2. Reporting Period: APRIL 1992 NOTES l

3. Licensed Thermal Power (MWt): 1500

4. Nameplate Rating - (Gross MWe): 502

5. Design Elec. Rating (Net MWe): 478~

6. Max. Dep. Capacity (Gross MWe): 502 7'. Max. Dep.' Capacity (Net MWe): 478

8. If changes occur in Capacity Ratings (3 through 7) since last report, give reasons:

'9. Power Level to which restricted, if any:(Net MWe):

10. Reasons for restrictions, if any:

THIS MONTH YR-TO-DATE CUMULATIVE

11. Hours in Reporting Period........... 719.0 2903.0 163033.0 '

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12. Number of Hours Reactor was Critical .0 751.0- 125565~.~T

13. Reactor Reserve Shutdown Hours...... .0~ .o~ ~ 1309.5

14. Hours Generator On-line............. .0 746.0 124123.1

15. Unit Reserve Shutdown Hours......... .0 .0 .0

16. Gross-Thermal Energy Generated (MWH) .0 958856.1- 162582581.8

17. Gross Elec. Energy Generated (MWH).. .0 323236.0 53499362.2

-18.-Net Elec. Energy Generated (MWH).... .0 306776!6_ 51040528.0

19. Unit-Service Factor................. .0 25.7 76.1 20.' Unit Availability Factor............ .0 25.7 76.1

.21. - Unit Capacity: Factor-(using MDC Net) .0 22.1 68.1 22.. Unit' Capacity Factor (using DER Net) .0 22.1 66.3

23. Unit Forced Outage Rate............. .0 .0 3.9 H2 4. Shutdowns scheduled over next 6 months (type, date, and duration of each):

THE' THIRTEENTH REFUELING OUTAGE CONCLUDED AND THE PLANT WENT ON-LINE MAY 3, 1992. NO' OUTAGES ARE SCHEDULED OVER THE NEXT SIX MONTHS.

25. If. shut down at end of report period, estimated date of startup: 05/03/92

26. Units in testLstatus-(prior to comm.-oper.): Forcast Achieved INITIAL CRITICALITY INITIAL ELECTRICITY N/A __

COMMERCIAL-OPERATION

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AVERAGE DAILY UNIT 1 POWER LEVEL DOCKET NO. 50-285 UNIT FORT CALHOUN-STATION.

DATE MhY ____ 11'lY92 COMPLETED BY G.-R. CAVANAUGH-TELEPHONE (402)636c2474 MONTH APRIL 1992 DAY' AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL'

-(MWe-Net) (MWe-Net) 1 0- 17 0 2 0 18 0

3. 0 19 0 4 0 20 0

-5 -0 21 0 6 0 22 0 7 0 23 0 8- 0 24 0 0 25 0 10- 0 26 0 11 0- 27 0 12 0 28 0 4

4 13' 0- 29 0

'14 ' 0 30. 0 15 0 31

- 16 ' 0

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

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UNIT. SHUTDOWNS ~AND POWER REDUCTIONS DOCKET NO. 50-285 UNIT NAME Fort Calhouft St..

DATE Nay 11c 1992 CONPLETED BY G. R. Cavanauch

. TELEPHONE (402) 636-2474 REPORT NONTH April 1992 i No. Date Type'- Duration Reason' Method of Ucensee 'Symem . Compones Cause & Corrective .

(Hours) . Shuning Event ' Code

• Code' Action to

'J Down Reactor' Report # ' Preven Recurrence 9241 02101/92 5 2157.0 C 1 XX . . XXXXXX on February 1,1992, the 13th Fat Calhoun Station .

Refueling Outage cornmenced.

e 1 3 s 4 F: Forced Reason: Method: Exhibit G - Instructions S: Scheduled A-Equipmes Failure (Explain) 1-Manual for Preparation of Data B-Maizaenance or Test 2-Manual Scram. Emry Sheets for Ucensee C-Refueling . 3-Automatic Scram. Evem Report (LER) File (NUREG4161)

D-Regulatory Restriction 4Oher (Explain)

E-Operstor Training & Ueense Examinaticus s

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

H-Other (Explain)

(9/77)

Refueling Information Fort Calhoun - Unit No. 1 Report for the month ending enril 1992

1. Scheduled dated for next refueling shutdown. Refuelina ou"aae beaan on February . 1992

2. Scheduled date for restart following refueling. May 3. 1992

3. Will refue. ling or resumption of operations thereafter require a technical specification char.ge or other license amendment? __ Ye s

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

Incorporate specific requirements resulting from reload safety analysis,

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

c. If no such review 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 associated with refueling, e.g., new or different fuel design or supplier, unrevlewed design or performance analysis methods, significant changes in fuel design, new operating procedures. New fuel supplier New LOCA analysis

6. The number of fuel assemblies: a) in the core _133 Assemblies in the spent 529 Assemblies b) fuel pool c) spent fuel pool storage capacity 729 Assemblies d) planned spent Planned to be fuel pool increased with higher storage capacity density spent fuel I racks.

7. The projected date of the last refuelina 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 MNb Date 0 M -i t

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. l OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No. 1 APRIL 1992 Monthly Operating Report I. OPERATIONS

SUMMARY

Fort Calhoun Station (FCS) remained shutdown for its thirteenth Refueling and Maintenance Outage. Major activities that occurred during April included the following:

Reactor Core Mapping and fuel Reload Verification Engineered Safeguards Surveillance Testing Draining and Refilling of Steam Generators Blacklighting of the Condenser Replacement of Station Battery No. 2 Reestablishment of the RCS Pressure Boundary by Reinstalling the o mssurizer Hanway and the Reactor Vessel Head Fifteen of the twenty-eight Rhodium Incore Nuclear Instrumentation Detectors in the reactor were replaced. The replacements were necessary due to normal deviation of the detectors, and in one case, a detector failure. During that process, a flashlight end cap was accidentally dropped through a tool access hole on the Upper Guide Structure (UGS) lift rig into the Reactor core. Removal of the cap required draining the Reactor cavity and removing the UGS lift rig. Operations located the cap lodged in the bottom of the shroud of CEA No. 3. Operations removed the cap by vacuuming it with the suction hose of the cavity filter.

On April 12, 1992, with the plant in an off-normal electrical lineup, 480 VAC feeder breaker IB3A tripped. This resulted in a loss of Shutdown Cooling Flow indication and control power to FCV-326. Upon loss of control power, FCV-326 failed open. The pump providing shutdown cooling flow was appropriately secured by Operations to protect it from possible runout. The o restored power,perating crew determined the cause of the loss of power,placed the FC valve and restarted the pump within seven minutes. This event was reported under Licensee Event Report 92-15.

On April 23, 1992 the RCS was taken water solid for the performance of the cold hydrostatic (200 psia) leak test. Performance of this test is completed prior to plant heatup to detect any gross leakage. No significant leakage was detected.

On April 26, 1992 plant heatup commenced to 395 F in preparation for plant startup. The hot hydrostatic test was conducted at 395 F and 2250 psia.

The test yielded several minor valve and fitting leaks. Fort Calhoun Station maintenance personnel repaired these leaks. A small flange leak on RC-142 (Pressurizer Code Safety Valve) was also discovered during this test. The leak was Furmanite sealed prior to power operation. With the hydrostatic test successfully completed, plant heatup to Mode 3 (Hot Shutdown: Tavg > 515 F, and 4% shutdown margin) was accomplished on April 28, 1992.

Surveillance testing required prior to Mode 2 was successfully performed in preparation for taking the reactor critical in early May 1992.

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9 Mohthly Operating Report '

LIC-92-179R Page 2 The following NRC inspections took place during April 1992:

IER No. Title 92-03 Fuel integrity and Reactor Subcriticality 92-09 Monthly Resident Inspection 92-10 Closcout Items Inspection The following LERs were submitted during April 1992:

LER No. LER Date Description 92-09 04/01/92 VIAS During Fuse Replacement in 86B/CRHS 92-11 04/20/92 Unacceptable Valve Arrangement for Service Air System -Containment Penetration H-74 92-12 04/24/92 Non-conservative Steam Generator Differential Pressurc Trip Setpoints A. SAFETY VALVES OR PORV CHALLENGES OR FAILURES WHICH OCCURRED None.

B, RESULTS OF LEAK RATE TESTS Due to the 1992 Refueling Outage, only two RCS leakrate tests were performed in April 1992. The first test was performad on April 29 1992 and the leakrate was 0.156 gpm total (0.125 gpm known and 0.031 -

gpm unknown . The second leakrate test was performed under Surveillance) Test OP-ST-RC-0001 and the results ware 0.258 gpm total (0.068 gpm known and 0.190 gpm unknown).

C. CHANGES, TESTS AND EXPERIMENTS REQUIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT TO 10CFR50.59 Amendment No. Description 143 This amendment makes changes to the Technical Specifications to revise the negative limit for- the Moderator Temperature Coefficient required for the Cycle-14 Reload.

144 This amendment makes changes to the Technical Specifications to incorporate the latest NRC approved revisions to the Core Operating Limits Report (COLR) required for the Cycle 14 Reload. The chances are

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administrative in nature.

Monthly Operating Report LIC-92-179R Page 3 Amendment No. Description 145 This amendment makes changes to the Technical Specifications by implementing Generic letter 90-09 concerning Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Action.

D. SIGNIFICANT SAFETY RELATED MAINTENANCE FOR THE MONTH OF APRIL 1992 Significant safety related work performed during April 1992 included:

Rebuilt and reinstalled "B" raw water pump (AC-10B).

Calibrations and repairs were performed on the following bus-tie breakers:

BT-184A BT-1B4C Removed CA-555 service air supply header, in board isolation valve)(containmentand associated piping and replaced with a blank flange.

A cracked cam follower was discovered on a SBN type switch (to be discussed further in LER 92-017). As a result, an inspection of safety significant SBM control switches throughout the plant was performed.

Repaired / replaced the following SBM control switches in the Control Room:

C/TS-120 (Test switch for Channel "C" of DSS)

CS/1A13 Control switch for 4.16 kV incoming breaker 1A13)

CS/lA44 Control switch for 4.16 kV incoming breaker lA44)

CSl/lA-D (Control switch for D1 breaker LAD 1)

HC-25B (Control switch for HCV-258)

The following SBM type switches we:_ replaced in breakers / cubicles in the plant:

1A33 (Feeder for Bus lA3) 1A3-12 Feeder for Transformer TIB-3B) 1A3-16 Feeder for Auxiliary feedwater Pump FW-6) 1A3-5 ( eeder for Reactor Coolant Pump RC-3C) 1A4-1 (DG-2 Feed to Bus lA4) 1A4-10 Breaker for Transformer TlB-4A) 1A4-ll Feeder for Raw Water Pump AC 10B) 1A4-14 Feeder for LPSI Pump SI-1B) 1A4-15 Feeder for Lighting Transformer TIC-4A) 1A4-16 Feeder for Reactor Coolant Pump RC-30) 1A4-18 345 kV Standby Feed to Bus lA4) 1A4-20 161 kV Normal Feed to Bus lA4)

Monthly 0

LIC-92-17heratingReport:

R Page 4 1A4-3 fFeeder for Circu;cting Water Pump CW-lC) 1 A4-4 I 1A4-5 i Feeder for Feed-Pump FW-4C llFeederforHeaterDrainPumpFW-5C) 1A4-8 l Feeder for Trsnsformer -

TIB)F1

-1A4-6 I

- 1A4-12, Feeder (Feeder for for RawCudensate Pump F=-2C))

Water Pump-AC-100 Replaced rotor retaining bolts and cooling fan blades for DG-2.

Refurbished and tested the following motor operated valves (MOV) as part_of_the M0V testing program:

.HCV-1041C Main Steam Bypass Valve

'tiCV-1042C Main Steam Bypass Valve HCV-1384 ( ain and Auxiliary feed ater cross-connect valve)

HCV-150 Pressurizer RC-4 relief isolation valve)

HCV-151 HCV-308- Pressurizer Chargi g- Pump RC-4 reliefdischarge CH-1A/Blc isolationto valve)HPSI header isolation valve HCV-311 HPSI t RC Loop 1B Isolation Valve HCV-315 HPSI: to RC Loop 1A Isolation Valve HCV-317: HPSI to RC Loop 2A Isolation Valve HCV-320 HPSI to RC Loop 2B Isolation Valve HCV-490A-(Comp.CoolingHeatExchangerAC-1 CCWInletValve)

Replaced / repaired the following "86" type (lockout) relays:

86A/CPHS (Channel "A" Containment Pressure High Signal Lockout Relay) 86A/PPLS (Channel "A" Pressurizer Low Pressure Signal- Lockout Relay) 86A/ VIAS (Channel "A" Safety Injection Actuation Signal

-Lockout Relay) 86A/ VIAS (Channe l "A" Ventilation Isolation Actuation Signal Lockout Relay) 86B/CRHS (Chaniel "B" Containment -High Radiation -Signal Lockout Relay 86Bl/ VIAS (Cha)nnel "B" Ventilation Isolation Actuation Signal.

Lockout Relay) l

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