ML20113G664

Monthly Operating Rept for March 1992 for Fort Calhoun Station

ML20113G664
Person / Time
Site:	Fort Calhoun
Issue date:	03/31/1992
From:	Edwards M, Gates W OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LIC-92-154R, NUDOCS 9204150285
Download: ML20113G664 (8)

Text

s' Omaha Public Power District 444 South 16th Street Mall Omaha, Nebraska 68102 2247 402/636-2000 April 14, 1992 LIC-92-154R U. 3. Nuclear Regulatory Commission ATTN: Document Control Desk Hail Station Pl-137 Washington, DC 20555

Reference:

Docket No. 50-285 Gentlemen:

SUBJECT:

March 1992 Honthly Operating Report (MOR)

Enclosed is the March 1992 HOR 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,

&. N EDO W. G. Gates Division Manager Nuclear Operations WGG/sel Enclosures c: LeBoeuf Lam S.D. Bloom,b,Leiby&MacRae NRC Pro ect Engineer R. D. Martin, NRC Re ional Administrator, Region IV R. P. Mullikin, NRC e-ior Resident Inspector R. T. Pearce, Combusti i Engineering R. J. Simon, Westinghouse Office of Management & Program Ans. lysis (2)

INP0 Records Center American Nuclear Insurers t

9204150285 92 P PDR ADOCK 05 85 h I R PDR j

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OPERATING DATA REPORT DOCKET NO. 50-285 ~~

UNIT FONT tKLilDOff STXTTOD DATE APRIL 08_,1992 COMPLETED BY M. L. EDWARDS OPERATING STATUS TELEPHONE T4021Z6362245'l

l. Unit Name: FORT CALHOUN STATION

2. Reporting Period: MARCH 19"2- NOTES

3. Licensed Thermal Power (MWt): 1500

4. Nameplate Rating (Gross MWe) $D 2-

5. Design Elec. Rating (Net MWe): 470

6. Max. Dep. Capacity (Gross MWe): 502

7. Max._Dep. Capacity (Net MWe): -~T78'

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

N/A _

9. Power Level to which restricted, if any (Net MWe): N/A

10. Reasons for restrictions, if any:

N/A THIS MONTH YR-TO-DATE CUMULATIVE

_ _ _ _ _ _ _ _ _ . __________ __________ a

11. Hours in Reporting Period........... 744.0 2184.0 162314.0

12. Number of Hours Reactor was Critical .0~ 7 5f."o- 1735T977'

13. Reactor Reserve Shutdown Hours...... .6 . 0~ ~~~~~1309.5

14. Hourr Generator On-line............. .6 7T670 124123.1

15. Unit Heserve Shutdown Hours......... .0 .0 .0 IT2$Yis81.8

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16. Gross Thermal Energy Generated (MWH) 76 - 9~5'8 s i 6 ~ l~

17.-Gross Elec. Energy Generated (MWH).. . 6' 3 fi2T6To~

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~5T4##162T2-

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18. Net Elec. Energy Generated (MWH).... .0 3067767f 5104~d528.0' -

19. Unit Service' Factor................. .0 3172~ 76T5~

20. Unit Availability Factor............ .6 34.Y -7f.~s"

21. Unit Capacity Factor (using MDC Net) .6 2VIT~ 6 67~4~

22. Unit Capacity Factor (using DER Net) .0 2Y!4 66T6'

23. Unit Forced Outage Rate............. _

.0 . 6~ 3T9'

24. Shutdowns scneduled over next 6 months (type, date, and duration of each):-

THE PLANT IS CURRENTLY SHUTDOWN FOR THE 13TH REFUELING OUTAGE. THE OUT-AGE-IS SCHEDULED TO BE_C_OMPLETED_ON_AP_RIL_26,_1992.

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25. If shut down at end of' report period, estimated date of startup: 04/_26/92

-26. Units in test status (prior to comm. oper.): Forcast Achieved INITIAL CRITICALITY ~

INITIAL ELECTRICITY N/A _

COMMERCIAL OPERATION _

AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 -~

UNIT YDNT CXf!HDUiFsiKTIi5N DATE XPRfL ~ 08 1742 COMPLETED BY R7 T.~s6WKRD'S

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TELEPHONE H~0 2 ) 636;Y451 ,

MONTH MARCH 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 1 9 0 25 0-I

.10 0 26 0 l 11 ~- 0 27 0 o

l 12 0 28 0 1

13 0 29 0 14 0 30 0 15 0 31 0 16 0

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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UNIT SHUTDOWNS Alm POWER REDUCTIONS. DOCKET No. 50-T!85 UNIT NAME For" Calhoun St.

DATE Apr' 1 8. 1992

! CONPLETED BY N. L. Edwards-l TELEPHONE (402) 636-2451 '

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REPORT MONTH March 1992 j- .

t j'~ No. Date Type' Duration Reaere' Method of Ixeneen Syssem Cerrpones Cause & Corrective ,

(linurs) Shutting . Event code' Code' Actice to  !

4 Down Reactor' Report # Prevent Recurrence t

, 9241 02/01/92 5 1438.0 C 1 XX X:OOOOC On Felmsery 1,1992, the 138 Fort cad >=a station Refuelmg Outage ' s l commenced.

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g F: Fo ced Reasen: Met!ud: Embebit G - Imeneneas t

[' ' 5: Scheduled A-Equipmera Failure (Explain) . 1-Manual for Prepersoon of Data n

B-Maissenance or Test 2-Mennel Screm. Entry Sheets for Licensee [

C-Refueling 3-Automatic scrast Eves Report (1.ER) File (NURFG4161) 1 D-Regulatory Restrictsan 4Oher (Explais) 1 E4perator Training & License Exansnation 6 3 . F-Administrative 7 j Goperational Error (Explain) Exhht 1 - Some Source t HOber (Erplein)  ;

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,' ', Refueling Information Fort Calhoun - Unit No. 1 Report for the month ending March 1992

1. Scheduled dated for next refueling shutdewn. Refuelina outaae hagan on Februarv . 1992

2. Scheduled date for restart following refueling. April 26. 1992

3. Will refueling or resumption of operations thereafte 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.

b. 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 review has taken place, when is it scheduled? N/A

4. Scheduled date(s) for submittin proposed licensing action and support in ormation.

gittedNcvember27.

5. Important licensing considerations associated with refueling, e.a., new or different fuel design or supplier, unreviewed design or performance analysis methods, significant chan design, new operating procedures, ges in fuel New fuel supplier New LOCA analysis 6, The number of feel assemblies: a) in the core 133 Assemi es in the spent 529 Assemi es -

b) fuel pool c) spent fuel pool storage capacity 729 Assemblier d) planned spent Planned to be ,

fuel pool increased wit' pher storage capacity density spe.. 'uel racks.

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

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

Prepared by_ M dl h Date 4 *P- 9t -

OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No. 1 MARCH 1992 Honthly Operating Report I. OPERATIONS

SUMMARY

In March 1992, Fort Calhoun Station remained shutdown for its thirteenth Refueling Outene. Both steam generators were drained and backfilled with nitrogen to pu( them in a dry layup condition. Circuit trip checking and bus maintenance was conducted on: 345 KV, non-vital 4160 V bus IAl vital 4160 V bus lA3, vital 120 VAC inverter A, station battery No. I and,DC bus No. 1. Diesel generator No. I was taken out of service for its refueling outage inspection. The circulating water system was taken out of service for ultrasonic testing of condenser tubes and valve work. All four reactor coolant pump motors were run for post maintenance testing after annual preventative maintenance was completed. Thermal shield work and reactor vessel inspection were also successfully completed.

The raw water system was taken out of service to facilitate piping and valve replacement. This system indirectly provides cooling to the s)ent fuel pool. In preparation, the temperature of the spent fuel pool, w11ch contained the off-loaded reactor core, was lowered to 68*F before the pumps were secured. The s and reached a peak of 108' pent fuel F before pool repair temperature work was completed. was closely monitored The compnnent cooling water (CCW) system was also taken out of service to aerform the Ten Year Inservice Inspection Hydrostatic Test. The 1

lydrostatic outside test was split2 intothe containment, three CCWphases:

surge tank, ) andpiping and components

3) piping inside containment. Only pha)ses1 and 2 of the hydrostatic test required isolation of cooling for the spent fuel pool. The spent fuel pool temperature was lowered as far as possible before the CCW pumps were secured. The spent fuel pool temperature was closely monitored as the test progressed. In phase 1, the spent fuel pool reached a peak temperature of Il5'F. In phase 2, the spent fuel pool reached a peak temperature of 90*f. All three phases of the hydrostatic test were performed satisfactorily. .

Ultrasonic testing was successfully completed on 50 spent fuel assemblies that will be returned to the reactor core. A defective fuel pin was found on fuel assembly N-08 and the pin was replaced with a rod made from stainless steel.

In response to operating experience information, a special " vortex" test was aerformed at midloop to determine how far valve FCV-326 shutdown cool ng flow control valve) could open before an operating low (pressure safety injection (LPSI) pump vortexed and lost suction.

On March 28, 1992 the reactor refueling cavity was filled and the upper guide structure removed in preparation for fuel reloading. Fuel reloading was successf ully completed in two and one half days, and core mapping and alignment checks were in progress at month's end.

Monthly Operating Report LIC-92-154R Page 2 The following NRC inspections took place during March 1992:

KUA ltth 92-05 Residents' Monthly Inspection 92-06 Inservice Inspection Progrtms 92-07 Radiation Protection Programs 92-08 Boric Acio Corrosion Prevention Programs The following LERs were submitted during March 1992:

LER No. DESCRIPTION 92-05 Unplanned Reactor Protective System Actuation 92-06, Rev. 1 Inoperable Alarm Function on Radioactive Waste Building Stack Monitors 92-07 Inadvertent Isolation of Radiation Monitors During Containment Purge 92-08 Safety Injection Relief Valve Setpoints Greater than Qualified System Design Pressures A. SAFETY VALVES OR PORV CHALLENGES OR FAILURES WHICH OCCURRED None B.. RESULTS OF LEAK RATE TESTS Reactor coolant system leak rate tetting was not performed during March 1992 since the plant is in Refueling Shutdown (Mode 5).

C. CHANGES, TESTS AND EXPERIMENTS REQUIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT T0 10 CFR 50.59

-Amendment No. Description 141 The amendment revises the Technical Specifications, removing the cycle specific operating limits .and instituting the Core Operating Limits Report in accordance with Generic Letter 88-16.

142 The amendment makes changes to Technical Specification 3.3(tor reac vessel specimen withdrawal schedules.1)c by implementing L

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Monthly Operating Report '

LIC 92-154R Page 3 D. SIGNIFICANT SAFETY RELATED MAINTENANCE FOR THE MONTH OF HARCH .1992 Significant safety related maintenance performed during March 1992 included:

Calibration and installation of an ammeter for breaker unit BT-183A.

Correction of a fault in the hand control switch (HC/SI-2A) on high pressure safety injection (HPSI) pump SI-2A.

Repairs and/or maintenance on the following valves / operators:

HCV-1335 0 Main feedwater isolation valve operator to stesm generator B)

HCV-1386-0 Main feedwater isolation valve operator to sttam generator A)

HCV-314-0 HPSI isolation valve operator to loop 1A HCV-329 0 LPSI isolation valve operator to loop 18 HCV-333-0 LPSI isolation valve operator to loop 2B HCV-335-0 Shutdown cooling heat exchanger inlet valve operator to eader HCV-347-0 (Shutdo)wn cooling from loop 2 containment isolation valve operator)

HCV-485-0 (Outlet to shutdown cooling heat exchanger 48 isolation valve operator LCV-218-2-0 (Volume ) control tank outlet low level isolation valve operator)

LCV-218-3-0 (Charging pump suction from safety injection refueling water tank valve operator)

PCV-742E, PCV-742F, PCV-742G and PCV-742H (radiation monitoring cabinet ~

isolation valves)

RC-142 (Pressurizer RC-4 relief valve)

Replacement of fan blades for diesel generator DG-1 Jacket water radiator JW-3-1.

Repair of a broken weld in steam generator RC-2A at the auxiliary feed water nozzle thermal liner.

Repair of a crack in the reactor coolant pump RC-3D case to cover-inner gasket leak-off line.

Repair and/or mainterance actions on 480V GE breakers IB3A 3, IB3A-6, 183B, 1838-5 and 183C.

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