ML20045H048

Monthly Operating Rept for June 1993 for Fcs,Unit 1.W/930714 Ltr

ML20045H048
Person / Time
Site:	Fort Calhoun
Issue date:	06/30/1993
From:	Gates W, Howman M OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LIC-93-0182, LIC-93-182, NUDOCS 9307160288
Download: ML20045H048 (8)

Text

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Omat a Public Power District l P.O. Box 399 Hwy.75. No th of Ft.Calhoun Fort Calhoun, NE 680230399 j 402/636-2000 July 14, 1993 LIC-93-0182 I

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station PI-137 Washington, DC 20555

Reference:

Docket No. 50-285 Gentlemen:

SUBJECT:

June 1993 Monthly Operating Report (HOR)

Enclosed is the June 1993 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, b, 5.

W. G. Gates Vice President  ;

WGG/pjc Enclosures c: LeBoeuf, Lamb, Leiby & MacRae J. L. Milhoan, NRC Regional Administrator, Region IV S. D. Bloom, NRC Project Manager R. P. Mullikin, NRC Senior Resident Inspector R. T. Pearce, Combustion Engineering R. J. Simon, Westinghouse Office of Management & Program Analysis (2)

INPO Records Center I

f 9307160288 930630 F / /

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. OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No. 1 JUNE 1993 Monthly Operating Report

1. OPERATIONS

SUMMARY

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The station began the month at 100% power. On June 14, 1993, moderator I temperature coefficient testing was conducted successfully at approximately 95% power. Subsequently, full power operation resumed until a plant trip occurred at 1322 hours0.0153 days <br />0.367 hours <br />0.00219 weeks <br />5.03021e-4 months <br /> on June 24. The cause of the trip was 1 determined to be inadvertent jarring of a 345KV fault relay when a cabinet door was opened during maintenance in the switchyard. This relay trip 1 f caused opening of the generator output breakers a turbine trip, and a l reactor trip. Post-trip actions included troubleshooting and testing of 1 non-safety related electrical components and main feedwater heater relief valves. Further details of this event will be reported in LER 93-011.

The reactor was brought critical at 1412 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.37266e-4 months <br /> on June 26. Reactor power was slowly raised and the turbine generator was synchronized to the grid at 1115 hours0.0129 days <br />0.31 hours <br />0.00184 weeks <br />4.242575e-4 months <br /> on June 27. Power was held at 30% for chemistry ,

considerations. At 0606 hours0.00701 days <br />0.168 hours <br />0.001 weeks <br />2.30583e-4 months <br /> on June 30, chemistry limits were met and a i power ascension commenced. Power was raised to 70% over the next several  !

hours, where it remained until July 1 for the testing of the feedwater l heater relief valves mentioned above. {

On June 29, the annual Emergency Plan Exercise was conducted; this exercise was integrated into the overall Federal Radiological Monitoring and Assessment Center drill.

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The fellowing NRC inspections were completed during this reporting period:

IER No. DescriDtion .

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93-06 Residents' Routine Inspection .'

Fire Protection Inspection 93-09 93-10 Confirmatory Measurements Inspection 93-14 Maintenance Inspection The following LERs were submitted during this reporting period:

LER No. Description t

93-007 Unplanned Emergency Diesel Start and Reactor Trip Signal 93-008 Failure to Implement Surveillance Requirement for Stack Flow Indicator 93-009 Automatic Start of Backup Charging Pumps and Isolation of Letdown F1ow l

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. MONTHLY OPERATING REPORT  ;

JUNE 1993 PAGE TWO

2. SAFETY VALVES OR PORY CHALLENGES OR FAILURES WHICH OCCURRED j No primary safety or power operated relief valve challenges or failures occurred. Although the plant tripped on loss of load on June 24 primary pressure only reached approximately 2166 psia, which -is well below the PORV or primary safety valve setpoints.

3. RESULTS OF LEAK RATE TESTS The reactor coolant system (RCS) leak rate was less than 0.200 gpm for most of June. On June 5, the leak rate increased due to packing leakage from Charging Pump CH-1A. The pump was repacked and returned to service with no further indication of increased leakage, Except for the increase on June 5, the. leak rate remained steady until June 24 when the plant tripped. Leak rate test results have been erratic since then due to changes in reactor power level involving boration and dilution of the RCS. However, there is no indication of a change in the previously observed trend.

4. L A.tLGES. TESTS AND EXPERIMENTS RE0VIRING NUCLEAR REGULATORY COMMISSION A.lH0 llZATION PURSUANT TO 10CFR50.59 Amendment No. Description l

NONE

5. SIGNIFICANT SAFETY RELATED MAINTENANCE FOR THE MONTH OF JUNE 1993

• Replaced rotating assembly on Raw Water Pump AC-10A.

• Replaced the discharge pressure controller root valve for Raw Water Pump.

AC-10A.

• Cleaned the raw water side of Component Cooling Heat Exchanger AC-1A.

Performed abrasive blast and applied an epoxy coat to the tube sheets, sides and end bell interiors.  ;

• Rebuilt Component Cooling Water (CCW) Inlet Valve HCV-489A and CCW Outlet Valve HCV-4898, and replaced Solenoid HCV-489A-20 and Solenoid HCV-489B-20.

• Replaced the plunger packing and front ca) gaskets on Charging Ps 'H -

1A and repaired the Suction Accumulator C1-26A. 3

• Replaced Level Controller LC-283 on Packing Cooling Pump CH-IA-1.

• Repaired loose connection and replaced ammeter on'125V AC Battery Charger No. 2.

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6. OPERATING DATA REPORT Attachment I L'

. MONTHLY OPERATING REPORT JUNE 1993 PAGE THREE

7. - AVERAGE DAILY UNIT POWER LEVEL Attachment II-

8. UNIT SHUTDOWNS AND POWER REDUCTIONS Attachment III

9. REFUELING INFORMATION. FORT CALHOUN STATION UNIT NO. 1 Attachment IV i

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. ATTACHMENT I 1 OPERATING DATA REPORT DOCKET NO. 50-285 FBRT dKf;HUUN 5fXffdD UNIT DATE 3DLY 08.199'3 COMPLETED BY M. A. HDWMXN OPERATING STATUS TELEPHONE jyj-533-65~39 .

1. Unit Name:

n FORT CALHOUN STATION

2. Reporting peFlod: JON5~TY53] NOTES ,

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): 475[

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 '

-11. Hours in Reporting Period........... 720.0 - 4343.0 173257.0

12. Number of Hours Reactor was Critical 671.1 4f3316 134743.9'

13. Reactor Reserve Shutdown Hours......

.0 .0 1309.5

14. Hours Generator On-line............. 649.4 4103.6 133166.8

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

'16. Gross Thermal Energy Generated (MWH) 883603.7 33f5'721.1 175185197.i~

17. Gross Elec. Energy Generated (MWH).. 21TidD7 V

~f81f53]io .~d~ 57726180.2

18. Net Elec. Energy Generated (MWH)..... 280104.3 ~1YFF390.8- 55070277.7

-19. Unit Service Factor................. 90.2 76.9

20. Unit Availability Factor............ 94.5~

90.2 94.5 76T9~

-21. Unit Capacity Factor (using MDC Net) 81.4'

~~ ~

86.T 6Di~d-

22. Unit Capacity Factor (using DER Net) 81.4- 837~7~ 677Y~

-23. Unit Forced Outage Rate............. 9.8 1. T

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4.3~

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

REFUELING OUTAGE SCHEDULED TO BEGIN ON SEPTEMBER 183 _1993 AND LAST APPROXIMATELY_56_ DAYS. -

25, If shut down at end of report period, estimated date of startup:

-26. Units in test status (prior to comm. oper.): Forecast Achieved INITIAL CRITICALITY INITIAL ELECTRICITY N/A ~~~

COMMERCIAL' OPERATION t

e e w e-' -gi- --w- w -

ATTACHMENT II AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50_-2__8_5 _ ___ ___ _

UNIT FORT CALHOUN STATION DATE ylfLY _03]i993 COMPLETED'BY M. A. HOWMAN TELEPHONE 402-533-6919 MONTH JUNE 1993 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 485 17 477 '

2 484 18 476 3 486 19 478 4 487 20 480 5 487 21 479 6 _

486 22 477 7 .

485 23 475 8 484 24 260 9 482 25 -0 10 _

481. _

26 0 11 479 27 20 12 450 28 100 13 449 29 '101 14 449 30 214 15 478 31 N/A 16 481 l 1

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.. 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, i

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ATTACHNENT III ' DOCKET NO. 50-285

_ UNIT SHlfTDOWNS AND POWER REDUCTIONS UNIT NAME Fort Calhoen St.

. DATE July 8.1993 CONPLETED BY N. A. Howsan.

TELEPHONE f402) 533-6939 REPORT NONTH June 1993 . .

No. Date Type t Duration Reason 2 Method of Ucensee - System Component Cause & Corrective

. (Hours) Shutting Event Code' Code s Action to Down Reactor 3 Report # Prevent Recurrence 3 06/24 S 3 F 70.6 B 3 93-011 EA RELAYX During switchyard mairmenance, a relay was inadvertaraly actuated.

Corrective actions will be discussed in LER 93411.

-1 2- 3- 4 F: Forced - Reason: Method: Exhibit G - Instructons

' 5: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparstion of Data B-Maintenance or Test 2-Manual Scram. Entry Sheets for Licensee C-Refueling 3-Automatic Scram. Event Repost (LER) Fde (NUREG416I)

D-Regulatory Restriction - 4Cher (Explain)

E-Operator Training & License Examination F-Administrative 5 ~

G-Operational Error (Explain) Exhibit l - Same Source HOber(Explain)

-(9/77)

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. . . - - - . . - . . . - . . . - = . - - - _ _ . - _ _ - - - .- _ _ - _----__ - -

Attachment IV Refueling Information Fort Calhoun - Unit No. 1 Report for the month ending June 1993

1. Scheduled dated for next refueling shutdown. September 1993

2. Scheduled date for restart following refueling. November 1993

3. Will refueling or resumption of operations thereafter require a technical specification change or other license amendment? Yes

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

Administrative i controls revision regarding- reload analysis methodology changes.

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 submitting proposed licensing action and support information. June 1993

5. Important licensing considerations associated with refueling, e.g., new or different fuel design  :

or supplier, unreviewed design or performance analysis methods,'significant chan ,

design, new operating procedures. ges in fuel None Planned

6. The number of fuel assemblies: a) in the core 133 Assemblies 529 Assemblies b) fuel poolin the spent-c) spent fuel pool storage capacity 729 Assemblies d) planned spent Planned to be fuel pool increased with high storage capacity density spent fuel-racks. '

7. The 3rojected date of the last refueling that can be disclarged to the spent fuel pool assuming the present licensed capacity. 1995*

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

Prepared by Id&#h 6e kcu Date 7/B[93 f /

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