LIC-97-0002, Monthly Operating Rept for Dec 1996 for Fort Calhoun Station Unit 1

Monthly Operating Rept for Dec 1996 for Fort Calhoun Station Unit 1

ML20133M557
Person / Time
Site:	Fort Calhoun
Issue date:	12/31/1996
From:	Edwards M, Tills J OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LIC-97-0002, LIC-97-2, NUDOCS 9701220426
Download: ML20133M557 (8)

Text

{{#Wiki_filter:. l Omaha Public PowerDistrict 444 South 16th Street Mall Omaha NE 68102-2247 , January 15, 1997 LIC-97-0002 l U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, D.C. 20555 1

Reference:

Docket No. 50-285 l l 1

SUBJECT:

December 1996 Monthly Operating Report (MOR) 2 Enclosed please find the December 1996 MOR for Fort Calhoun Station (FCS) Unit No.1 as required by FCS Technical Specification 5.9.1. . If you should have any questions, please contact me. 1 Sincerely, A i J. W. Tills Manager - Nuclear Licensing JWT/mle 1 Enclosures l c: Winston & Strawn L. J. Callan, NRC Regional Administrator, Region IV L. R. Wharton, NRC Project Manager ' W. C. Walker, NRC Senior Resident Inspector R. J. Simon, Westinghouse INP0 Records Center g qf l

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9701220426 961231 PDR ADOCK 05000285 R PDR i 45.5134 Employment with Equal Opportunity

2 LIC-97-0002 j Enclosure j Page 1 1 OMAHA PUBLIC POWER DISTRICT

Fort Calhoun Station Unit No. 1

) . j December 1996

Monthly Operating Report i

i 1. OPERATIONS

SUMMARY

i l Following the 1996 refueling outage, from December 1 through December 6, l the Fort Calhoun Station (FCS) increased power to 97%. On December 6 at

0902 hours, power was reduced to 92% to complete Moderator Temperature

! Coefficient testing. This testing was completed on December 8 at 1109 hours. 1 j On December 8 at 1440 hours, power was reduced to 85% to facilitate l turbine control valve testing. The testing was satisfactorily completed j at 1800 hours and power ascension commenced. i Unreliable emergency response facility ([tt ; computer operation caused i reactor power to be maintained between 95% and 98% between December 9 and December 16. This was done to maintain Peak Linear Heat Rate (PLHR) 4 at less than 90% of the limit (15.5 KW/ft as defined in Technical { Specification 2.10.4 and the Core Operating Limits Report). Technical Specification 2.10.4(c)i would have required power to be reduced to 80% (rodded) or 85% (unrodded) if the ERF computer had failed while at a ~ ! higher power level. Hardware problems were detected when the "C" host was running as the active computer. Circuit boards were reseated, power l was cycled and the "D" host was selected as the active computer. This i resolved the problem and on December 16 at 0910 hours, the ERF computer tested satisfactorily and reactor power was increased to 100%. i On December 31 at 1100 hours FCS entered Abnormal Operating Procedure i A0P-05, Emergency Shutdown, due to a non-isolatable steam leak on a i drain line going to the condensor. At 1128 hours, a Notification Of i Unusual Event (NOUE) was declared per EAL 11.6, " Plant Conditions { Warrant Increased Awareness By Plant Staff Or Government Authorities."

Appropriate NRC, state and local authorities were notified of the NOVE.

At 1330 hours, the turbine was manually tripped and the leak was isolated. The reactor remained critical at approximately 10% power during the repairs. At 1333 hours, A0P-05 was exited and at 1340 hours the NOVE was terminated. The leak was repaired and the turbine / generator was placed on line at 1821 hours. Dr. December 31 at 2400 hours, FCS was at 30% power waiting to clear steam generator chemistry holds prior to increasing power."

l LIC-97-0002 Enclosure Page 2

2. SAFETY VALVES OR PORV CHALLENGES OR FAILURES WHICH OCCURRED During the month of December, no power operated relief valves (PORV) or primary system safety valve challenges or failures occurred.

3. RESULTS OF LEAK RATE TESTS The reactor coolant system (RCS) leak rate was steady throughout the month continuing the trend of minimal RCS leakage following the 1996 refueling outage. December daily leak rates wefe constant at approximately 0.2 gpm except for increases caused by a charging pump packing leak. The leak rate returned to normal following completion of each charging pump packing replacement. No degrading trends were noted this month and the RCS continues to operate with minimal leakage.

4. CHANGES. TESTS AND EXPERIMENTS RE0VIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT TO 10CFR50.59 Amendment No. Description Amendment 179 Amendment 179 revises the Technical l Specifications to increase the amount of '

trisodium phosphate (TSP) dodecahydrate located in the containment sump storage baskets. l 1

5. SIGNIFICANT SAFETY RELATED MAINTENANCE FOR THE MONTH OF DECEMBER 1996 Replaced a 2 micron filter in CH-17A (Purification Filter "A")

with a 1 micron filter per ECN 94-568 Replaced Breaker MCC-3B1-E2R (Battery Room Fan VA-71A) per ECN 96-444 Removed gag on MS-282 (Main Steam Line "B" Relief Valve) Repaired compressor on VA-46A (Control Room Air Conditioning Unit) Replaced optics and master bulbs in YIT-6288A (Toxic Gas Monitor, Chlorine Gas) Replaced 5 General Electric CR120A relays per ECN 95-374 1

1 i LIC-97-0002 Enclosure Page 3 l

6. OPERATING DATA REPORT Attachment I l

7. AVERAGE DAILY UNIT POWER LEVEL

Attachment II l 8. UNIT SHUTDOWNS AND POWER REDUCTIONS i

Attachment III

9. REFUELING INFORMATION. FORT CALHOUN STATION UNIT N0. 1

Attachment IV 4

 - . . . - .       . .   ~.~      --..-         -    .-..-- -._- ---._.- -. . - - . - . - . . - _ - . . - . _
              .        .                            ATTACHMENT I OPERATING DATA REPORT l

i DOCKET NO. 50-285 UNIT FORT.CALHOUN STATION l DATE . JANUARY _07_,1997 COMPLETED BY M. L. EDWARDS OPERATING STATUS TELEPHONE 402-533-692i-

l. Unit Name: FORT CALHOUN STATION

2. Reporting Period: DEf?3pER 1996 NOTES

3. Licensed Thermal Power (MWt): 1500

4. Nameplate Rating (Gross MWe): 50f ~

5. Design Elec. Rating (Net MWe): 475[

6. Max. Dep. Capacity (Gross MWe): 50_2

7. Max. Dep. Capacity (Net MWe): 478

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........... 744.0 8784.0 203978.0

12. Number of Hours Reactor was Critical 744.0 6983.6 160691.6

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

14. Hours Generator On-line............. T39.1 ~ 6887.3 158867.8

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

16. Gross Thermal Energy Generated (MWH) 1064592.8 9817571.7 212503880.0

17. Gross Elec. Energy Generated (MWH).. 3Ti226.0 3284507.9 70218233.1

18. Net Elec. Energy Generated (MWH).... 344367.7 3128752.5 66986121.3

19. Unit Service Factor................. 99.3 78.4 77.9

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20. Unit Availability Factor............ 99.3~~ 78. 4 77.9

21. Unit Capacity Factor (using MDC Net) 96.8 74.5 70.9

22. Unit Capacity Factor (using DER Not) 96.8 74.5 69.3

23. Unit Forced Outage Rate............. .0 5.3 4.0

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

N/A

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

26. Units in test status (prior to comm. oper.): Forecast Achieved i INITIAL CRITICALITY l

INITIAL ELECTRICITY N/A I COMMERCIAL OPERATION _

ATTACHMENT II AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 l UNIT FORT.CALHOUN ST_ATION DATE JANUARY O_L 1997 COMPLETED BY M. L. EDWARDS TELEPHONE To2 ST3-69T9 MONTH DECEMBER 1996 l DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL I (MWe-Net) (MWe-Net) 1 321 17 486 2 411 18 488 3 465 19 488 4 471 20 488 5 471 21 ___ 488 6 458 22 488 1 7 444 23 488 8 435 24 488 9 468 25 488 l 10 470 26 488 11 470 27 488 12 470 28 488 13 472 29 488 14 475 30 488 15 476 31 261 16 481 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.

ATTACHMENT III DOCKET NO. 50-285 UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Fort Calhoun St. DATE January 9. 1997 COMPLETED BY M. L. Edwards TELEPHONE (402) 533-6929 REPORT MONTH December 1996 iNo? L Date) Type 1:. Duration 1 iReason .'- Method of.; L Licenseei lSystem ! 1 Component 1 5 Cause & Corrective ~

                                                                    '                                                   .; Code51
                                           )(Hours))                            j Shutting :    l Event r   E. Code"                                         : Action' to -

1Down2 . Report No.: i: Prevent Recurrence t

Reactor' :

96-07 961231 F 4.9 A 1 N/A CC PIPEXX On December 31, 1996 at 1100 hours, a non-isolatable steam leak on a drain line going to the condenser developed. The turbine was manually tripped and the leak was isolated. The reactor remained critical at approximately 10% power 4 during the repairs. Following repairs, the generator was placed online at 1821 hours. 1 2 3 4 F: Forced Reason: - Method: - Exhibit F - Instructions S: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparation of Data B-Maintenance or Test 2-Manual Scram Entry Sheets for Licensee C-Refueling 3-Automatic Scram Event Report (LER) File (NUREG-0161) D-Regulatory Restriction 4-Other (Explain) E-Operator Training & License Examination F-Administrative 5 H-Other (Explain) Exhibit H - Same Source (9/77) P S

'o - Attachment IV Refueling Information Fort Calhoun Station Unit No.1 Report for the month ending: December 31. 1996

1. Scheduled date for next refueling shutdown.

• March 23, 1998 I

2. Scheduled date for restart following refueling. May 2, 1998 j  !

3. Will refueling or resumption of operations No thereafter require a technical specification change or other license amendment?

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

b. If answer is no, has the reload fuel No design and core configuration been l i

reviewed by your Plant Safety Review l Committee to determine whether any l unreviewed safety questions are l associated with the core reload? l

c. If no such review has taken place, when Prior to May 2, 1998  ;

is it scheduled? l

4. Scheduled date(s) for submitting proposed No submittal required i licensing action and support information. , j

5. Important licensing considerations associated None i with refueling, e.g., new or different fuel ,

design or supplier, unreviewed design or l performance analysis methods, significant changes in fuel design, new operating  ; , procedures.  ! )

6. The number of fuel assemblies: i a) in the core 133 Assemblies b) in the spent fuel pool 662 Assemblies c) spent fuel pool storage capacity 1083 Assemblies i

7. The projected date of the last refueling that can be discharged to the spent fuel pool 2007 Outage assuming the present lic,ensed capacity.

Prepar by: _. 1 2a/ Date: //f/96

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