LIC-96-0004, Monthly Operating Rept for Dec 1995 for Fort Calhoun Station

Monthly Operating Rept for Dec 1995 for Fort Calhoun Station

ML20096C793
Person / Time
Site:	Fort Calhoun
Issue date:	12/31/1995
From:	Lippy D, Tira Patterson OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LIC-96-0004, LIC-96-4, NUDOCS 9601180283
Download: ML20096C793 (8)

Text

~

Y..

.J r.7 -;- -

. . .; . . .. .. n. . ;- .. . .-

Omaha Public Power District 444 South 16th Street Mall Omaha, Nebraska 68102-2247 402/636-2000 January 15, 1996 LIC-96-0004 l U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, D.C. 20555

Reference:

Docket No. 50-285 )

l I

SUBJECT:

December 1995 Monthly Operating Report (MOR)

Enclosed please find the December 1995 MOR for Fort Calhoun Station (FCS) l Unit No. 1 as required by FCS Technical Specification 5.9.1.

If you should have any questions, please contact me.

Sincerely, T. L. Patterson Division Manager Nuclear Operations TLP/dll Enclosures 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

180071a.. T. Pearce,

a. Simon, Combustion Engineering westinghouse t INP0 Records Center <, M DO O 0285 R PDR

.t LIC-96-0004 Enclosure Page 1 OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No. 1 DECEMBER 1995  ;

Monthly Operating Report

1. OPERATIONS

SUMMARY

During the month of December,1995, Fort Calhoun Station (FCS) operated at i a nominal 100% power with the exception of a one-day power reduction to  !

99.2% for placing an Ion Exchanger for the Reactor Coolant in service. I Normal plant maintenance, surveillance, equipment rotation activities and scheduled on-line modifications were performed during the month.

Monitoring of a minor Control Element Drive Mechanism (CEDM) mechanical seal leak continued.  !

On December 4,1995, a one hour non-emergency NRC notification was made as a result of the determination that the plant had been outside of its l design basis for maintaining an adequate quantity of Trisodium Phosphate 1 (TSP) in tM Containment Building to neutralize the sump water to a pH of a 7.0. The TSP is stored in the basement of the building and is designed to neutralize the boric acid which would be injected to the Reactor l Coolant System (RCS) and containment during a Loss-of-Coolant-Accident (LOCA). The amount of TSP in the FCS containment is sufficient to neutralize the sump water to a pH a 7.0 for current boric acid concentrations in the RCS, Safety Injection Tanks, Boric Acid Storage Tanks and the Safety Injection Refueling Water Tank. Corrective actions are being taken as reported in Licensee Event Report (LER)95-008.

On December 7,1995, the plant Fire Brigade was alerted and assembled to respond to smoke in the warehouse. The smoke was determined to be caused by an overheated motor on an oscillating fan. No fire suppression system or equipment discharge was required.

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.

.- . . _ - - - - -- . - . .- ~ - - -. . . .

,, eb i .

LIC-96-0004 Enclosure

. Page 2 i

3. RESULTS OF LEAK RATE TESTS Although above normal, the December 1995 RCS leak rate was steady at approximately 0.30 gpm throughout the month. This leak rate remained relatively steady following the reactor trip and resultant surveillance testing of the CEDMs on August 26, 1995.

The major contributor to the increase in RCS leakage has been classified as "Known" leakage. This leakage is being collected in the Reactor Coolant Drain Tank (RCDT). The leakage source for "Known" leakage has been attributed to CEDM #15. The "Known" leak rate has decreased slightly over the last several months to approximately 0.20 gpm. The remainder of the leakage has been classified as " Unknown" leakage.

In response to increasing containment activity, a containment entry was made on December 21, 1995 to inspect RCS components for leakage. One or more of the reactor head vent system isolation valves were found to be leaking through. Head vent system valve leakage is currently considered a minor contributor (estimated at less than 0.06 gpm) to both the "Known" and " Unknown" leak rate totals.

l l

4. CHANGES. TESTS AND EXPERIMENTS RE0VIRING NUCLEAR REGULATORY COMMISSION i AUTHORIZATION PURSUANT TO 10CFR50.59 Amendment No. Description 172 This amendment revised the Technical Specification on the chemical and volume control system (CVCS) to reformat and clarify the requirements and make them more consistent with the requirements of the Combustion Engineering Standard Technical Specifications (STS) as  ;

presented in NUREG-0212, Revision 2.

5. SIGNIFICANT SAFETY RELATED MAINTENANCE FOR THE MONTH OF DECEMBER 1995

- Refurbished Raw Water Pump Motor AC-10C-M Rebuilt Charging Pump CH-1C Replaced a broken yoke sleeve on the suction valve for Charging Pump CH-1C Replaced the regulator internals on the secondary air start SA-143 for Diesel Generator #1

LIC-96-0004 Enclosure Page 3 i

6. OPERATING DATA REPORT  ;

Attachment I

.i

7. AVERAGE DAILY UNIT POWER LEVEL  ;

l Attachment II 9

8. UNIT SHUTDOWNS'AND POWER RF. DUCTIONS Attachment III

9. REFUELING INFORMATION. FORT CALHOUN STATION UNIT NO. l' Attachment IV I

h l

l l

i i

l I

s

-+ - , w , e * -. -- - , - -=

.., l ATTACHMENT I OPERATING DATA REPORT DOCKET NO. 50-285 UNIT FORT CALHOUN STATION -

DATE JANUARY 04,1396 ~

COMPLETED BY D. L. LIPPY OPERATING STATUS TELEPHONE (402) 57I23843 _

l

1. Unit Name: FORT CALHOUN STATION I

2. Reporting Period: DECEMBER 1995 NOTES  :

1

3. Licensed Thermal Power (MWt): 1500 j

4. Nameplate Rating (Gross MWe): 502

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

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

i 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 Le"el 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 8760.0 195194.0

12. Number of Hours Reactor was Critical 744.0 7290.1 153708.0

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

14. Hours Generator On-line............. 744.0 7206.2 151980.5

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

16. Gross Thermal Energy Generated (MWH) 1113287.3 F0537607.8 202686308.3'

17. Gross Elec. Energy Generated (MWH).. 379432.0 3528843.0 66933725.2

~ii65576.5

~

18. Net Elec. Energy Generated (MWH).... 362705.4 ~3'3 8 5 7 3 6 8 . 8

19. Unit Service Factor................. 100.0 82.3 77.9

20. Unit Availability Factor............ 100.0 82.3 77.9

21. Unit Capacity Factor (using MDC Net) 102.0 80.4_ 70.7_

22. Unit Capacity Factor (using DER Net) 102.0 80.4 69.1

23. Unit Forced Outage Rate............. .0 3.7 4.0

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

A MAINTENANCE OUTAGE IS SCHEDULED TO OCCUR FROM MARCH 16-23, 1996 TO RE-PAIR / REPLACE DEGRADING CEDM MECHANICAL SEALS.

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 __

"' ATTACHMENT II AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 UNIT FORT CALHOUN STATION DATE JANUARY _ 04,1996 COMPLETED BY D. L. LIPPY TELEPHONE (402) 533 T8f3 MONTH DECEMBER 1995 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 1 488 17 488 2 488 18 488 3 488 19 487 4 487 20 487 5 488 21 487-l 6 487 22 487 7 487 23 488 8 487 24 488 9 487 25 488 __

10 488 26 488 11 488 27 488 12 488 28 487 13 488 29 486 14 488 30 488 15 487 31 487 16 488 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.

l l

3*

ATTACHMENT III DOCKET NO. 50-285 UNIT SHUTDOWNS AND POWER REDUCTIONS UNIT NAME Fort Calhoun St. 4'*

DATE January 9. 1996 COMPLETED BY D. L. LiDDY TELEPHONE (402) 533-6843 . j REPORT MONTH December 1995 sush;;- ~10sts? [TM' O ationi

~ '~

' h$s$s$

?lleilWili Sidensee; [SNtel Campement [6$eilCSrrective]M

• I w,

j(Neurs);t

..ll(ef.i.. 1Eventj ' (Co$ ( Ce@ * "g. g" ..

'dlActiektei" '

w;;;~7 g;5huttingt- Reperte as Prevent securrence

> ':t gewal.,.i f. i.les:.

m... , .: e. 4.

NONE t

s f

I 1 2 3 4 F: Forced Reason: Method: Exhibit F - Instructions S: Scheduled A-Equipment Failure (Explain) 1-Manual for Preparation of Data  !

8-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) i i

op. ': l

. Attachment IV l Refueling Information Fort Calhoun Station - Unit No. 1 Report for the month ending December 31. 1995

1. Scheduled date for next refueling shutdown. September 21. 1996

2. Scheduled date for restart following refueling. November 2. 1996

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 Enrichment limit of spent l these be? fuel racks is to be  !

increased to at least 4.5  !

w/o from 4.2 w/o. This is  !

necessary based upon the I oreliminary Cycle 17 core j Dattern development.  !

I

b. If answer is no, has the reload fuel design )

and core configuration been reviewed by your l Plant Safety Review Committee to determine l whether any unreviewed safety questions are  !

associated with the core reload. N/A l l

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

4. Scheduled date(s) for submitting proposed January 1996 (for soent licensing action and support information. fuel rack enrichment limit j chanae) 1 1

5. Important licensing considerations associated with refueling, e.g., new or different fuel design or supplier, unreviewed design or l performance analysis methods, significant changes l in fuel design, new operating procedures. N/A

6. The number of fuel assemblies:

a) in the core 133 Assemblies b) in the spent fuel pool 618 Assemblies c) spent fuel pool storage capacity 1083 Assemblies

7. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity. 2007 Outaae Prepared by M M l __ Date i9%