LIC-96-0174, Monthly Operating Rept for Oct 1996 for FCS Unit 1

Monthly Operating Rept for Oct 1996 for FCS Unit 1

ML20129K316
Person / Time
Site:	Fort Calhoun
Issue date:	10/31/1996
From:	Edwards M, Tira Patterson OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LIC-96-0174, LIC-96-174, NUDOCS 9611200218
Download: ML20129K316 (10)

Text

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$998 Omaha PublicPowerDistrict 444 South 16th Street Mall Omaha NE 68102-2247 November 15, 1996 LIC-96-0174 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, D.C. 20555 -

Reference:

Docket No. 50-285

SUBJECT:

October 1996 Monthly Operating Report (MOR)

Enclosed please find the October 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.

Sincerely,

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T. L. Patterson Df vision Manager Nucicar Operations TLP/mle Enclosures 200074 c: Winston & Strawn L. J. Callan, NRC Regional Administrator, Region IV L. R. Wharton, NRC Project Manaaer ~ . ..

I W. C. Walker, NRC Senior Resident Inspector J R. J. Simon, Westinghouse /

INP0 Records Center 9611200218 961031 PDR R

ADOCK 05000285

. ppR 45.5124 Employment with EqualOpportunity

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LIC-96-0174 Enclosure Page 1 l

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OMAHA PUBLIC POWER DISTRICT  !

Fort Calhoun Station Unit No. 1 OCTOBER 1996 l Monthly Operating Report

1. OPERATIONS

SUMMARY

Fort Calhoun Station (FCS) Unit No.1 operated at a nominal 45% power level until October 4th at 1529 hours0.0177 days <br />0.425 hours <br />0.00253 weeks <br />5.817845e-4 months <br /> when power was reduced in preparation for the scheduled refueling outage. The reduced power level i prior to the outage was a planned process, which successfully reduced reactor coolant system (RCS) radionuclide concentrations and associated outage dose. The generator was taken offline at 2207 hours0.0255 days <br />0.613 hours <br />0.00365 weeks <br />8.397635e-4 months <br /> and on October 5th at 0339 hours0.00392 days <br />0.0942 hours <br />5.605159e-4 weeks <br />1.289895e-4 months <br />, the reactor was made subcritical. Scheduled

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l maintenance and modifications are being completed and all twenty-eight incore nuclear detector strings have been replaced. The following l events occurred in October. '

On October 5th at 1339 hours0.0155 days <br />0.372 hours <br />0.00221 weeks <br />5.094895e-4 months <br />, FCS was performing a plant cooldown to 400 F per procedure OP-3A when steam generator low pressure trip signals were received on two of the four reactor protective system (RPS) ,

': anels. The reactor trip signal occurred as a result of not l

...otalling the bypass keys prior to reaching the low pressure trip setpoint. A four-hour notification was made to the NRC pursuant to 10 CFR 50.72(b)(2)(ii). This event is described in Licensee Event Report ,

(LER)96-007.

A discrepancy between FCS procedure 01-FH-1, " Fuel Handling Equipment i Operation.~ and the Design Basis Control Room Habitability Radiological Consequences Calculation was identified on October 8th. A four-hour  !

l non-emergency notification was made to the NRC pursuant to 10 CFR l 50.72(b)(2)(iii)(D). An Operations Memorandum was generated to ensure '

both control room filtration units ye placed in the filtered air makeup 1 mode and that at least one of the unita is in service during irradiated fuel movements. Appropriate controls tave been implemented throughout I the current refueling outage. This event is described in LER 96-003. )

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LIC-96-0174 Enclosure Page 2 i

i The steam generator primary side manways were removed on October 12th causing radiation levels in containment to rise above the containment radiation high signal setpoint. A ventilation isolation actuation signal (VIAS) resulted which ensured that the release remained below the limits specified in 10 CFR 20. A four-hour non-emergency notification was made to the NRC pursuant to 10 CFR 50.72(b)(2)(ii).

This event is described in LER 96-008.

On October 17th. Wyle Labs notified OPPD that the "As-Found" lift pressure of Pressurizer Safety Valves (PSV) RC-141 and RC-142 appeared i to be outside of the specified lift setting acceptance criterion. This l event is described in LER 96-009. l l

On October 18th, a four-hour non-emergency notification was made to the NRC pursuant to 10 CFR 50.72(b)(2)(iii)(C). A breach of containment closure was identified during core offload. Refueling operations were I stopped and maintenance was dispatched to close the flowpath. Refueling operations were not permitted to resume until containment was walked down to verify closure. This event is described in LER-96-010.

On October 30th, a four-hour non-emergency notification was made to the l NRC pursuant to 10 CFR 50.72(b)(2)(iii)(C) when a breach of containment closure was identified during fuel reload. The path identified was through an open secondary manway on Steam Generator "A" through a 8 inch

! needle valve that was 1/16th inch open. This event is described in LER- 1 96-011.

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

Surveillance testing of PORV valves PCV-102-1 and PCV-102-2 was l completed satisfactorily. (See previous section for discussion on Valves RC-141 and RC-142.)

, 3. RESULTS OF LEAK RATE TESTS Leak rate tests conducted through October 4th were steady at slightly above 0.2 gpm. Leak rate tests conducted on October 5th and 6th were slightly higher as a result of shutting the plant down to begin the l refueling outage. No leak rate tests were conducted after October 6th.

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~LIC-96-0174 Enclosure l Page 3

4. CHANGES. TESTS AND EXPERIMENTS REOUIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT TO 10CFR50.59 7

Amendment No. Descriotion  ;

l 1 176 Amendment 176 revise.c Sections 2.18, 3.14 '3.3 L and 5.10 of the Technical Specifications. The j

! Amendment relocates snubber operability l requirements to the Updated Safety Analysis

! Report and incorporates snubber examination and l testing requirements into Section 3.3 for ASME Section XI requirements. l l \

177 Amendment 177 modifies Paragraph 2.B(2) of  !

Facility Operating License No. DPR-40 to I reference 10 CFR Part 40 allowing the use of source materials, in the form of depleted or natural uranium, as reactor fuel.

178 Amendment.178 changes Section 4.3.2 to allow the use of zircaloy or ZIRLOS fuel cladding and the use of depleted uranium as reactor fuel  ;

material. Two new Westinghouse Topical Reports are also being added to Section 5.9.5.

5. SIGNIFICANT SAFETY RELATED MAINTENANCE FOR THE MONTH OF OCTOBER 1996 Replaced eleven (11) General Electric (GE) Type SBM switches and two (2)

GE CR120A relays.

Replaced sixteen (16) Agastat relays.

Replaced sight glass on Diesel Generator Fuei 011 Day Tank FO-2-1. i Rebuilt shuttle valves for Containment Purge Exhaust Isolation Valve

! PCV-742B and Containment Purge Air Inlet Outboard Isolation Valve PCV-

7420.

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LIC-96-0174 l Enclosure l Page 4 Replaced Breaker Unit (cubicle and breaker: 52/183C) 183C-183C.

Replaced RMS-9 Trip Unit On Breaker 183C-2.

Rebuilt Component Cooling Water Pump AC-3B. -

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Repaired chains and pulley on Containment Equipment Access Hatch AE-1.

Performed a variety of maintenance tasks on Diesel Generator No.1.

Performed dynamic MOVAT testing and replaced torque switch on Steam Generator RC-2B, Isolation Valve HCV-1385.

Replaced grease in Low Pressure Safety Injection (LPSI) to RC Loop 1B Isolation Valve HCV-327 and LPSI to RC Loop 1A Isolation Valve HCV-329.

. Rewelded support for Main Steam Line B Relief Valve MS-281.

Removed Pressurizer Safety Valves RC-141 and RC-142 and sent them to Wyle Labs for setpoint surveillance testing.

Retorqued two bolts on Steam Generator RC-2B Steam Separator Cannister.

Rebuilt LPSI Pump Motor SI-1A-M.

Replaced flange studs and nuts on Safety Injection Refueling Water Tank.

Outlet Check Valve SI-140.

Replaced diaphragm on Safety Injection Leakage Cooler Accumulator SI-7D.

6. OPERATING DATA REPORT Attachment I

7. AVERAGE DAILY UNIT POWER LEVEL l

Ati.achment II l

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LIC-96-0174

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8. UNIT SHUTDOWNS AND POWER REDUCTIONS Attachment III

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

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l ATTACHMENT I

, OPERATING DATA REPORT s l l DOCKET NO. 50-285 l UNIT FORT _CALHOUN STATION j DATE NOVEMBER _ 08,1996  ;

COMPLETED BY M. L. EDWARDS OPERATING STATUS TELEPHONE (402) 533-6929 1

l. Unit Name: FORT CALHOUN STATION

2. Reporting Period: OCTOBER 1996 NOTES l I

3._ Licensed Thermal Power (MWt): 1500 l

4. Nameplate Rating (Gross MWe): 502  ;

5. Design Elec. Rating (Net MWe): 47_8] I

6. Max. Dep. Capacity _(Gross MWe): 502 l 7. Max. Dep. Capacity (Net MWe): 478 1

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 j 10. Reasons for restrictions, if any:

N/A 1

THIS MONTH YR-TO-DATE CUMULATIVE l 11. Hours in Reporting Period........... 745.0 7320.0 202514.0

12. Number of Hours Reactor was Critical 99.6 6102.4 159810.4

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

14. Hours Generator On-line............. 94.1 --6056.0 158036.5 i 15.' Unit Reserve Shutdown Hours......... .0 .0 .0 l l 16. Gross Thermal Energy Generated (MWH) 62573.9 8707871.3 211394179.6 l 17. Gross Elec. Energy Generated (MWH).. 18716.0 2911913.9 ~

69845639.1 l 18. Net Elec. Energy Generated (MWH).... 16944.0 277 4 63~8. 5 66632007.3 i

19.~ Unit Service Factor................. 12.6 82.7 78.0 i

20. Unit Availability Factor............ 12.6 82.7 78.0

21. Unit Capacity Factor (using MDC Net) 4.8 79.3 71.0 i i 22. Unit Capacity Factor (using DER Net) 4.8 79.3 69.5 ]

23. Unit Forced Outage Rate............. .0 6.0 4.1 )

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

REFUELING' OUTAGE COMMENCED ON OCTOBER 5, 1996 WITH A SCHEDULED DURATION OF 44 DAYS.

25. If shut down at end of report period, estimated date of startup: 11/17/96 t

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

INITIAL ELECTRICITY N/A COMMERCIAL OPERATION l

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, ATTACHMENT II l , AVERAGE DAILY UNIT POWER LEVEL DOCKET NO. 50-285 j' UNIT TORT CALHOUN STATION l DATE NOVEMBER 08,1996 l COMPLETED BY M. L. EDWARDS l TELEPHONE (402) 533-6929 MONTH OCTOBER 1996 I

l DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net)

• l l- 188 17 0 2 186 18 0 1

3 186 19 0 i

4 146 20 0 5 0 21 0 -

6 0 22 0 -)

i 7 0 23 0 )

1 I

8 0 24 0 9 0 25 0 10 0 26 0 _ j 11 0 27 0 l  !

l 12 0 28 0 l

13 0 29 0 j 14 0 30 0 15 0 31 0 l 16 0 INSTRUCTIONS i On this form, list the average daily unit power level in MWe-Net for each i-day in the reporting month. Compute to the nearest whole megawatt.

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

DATE November 8. 1996 [

COMPLETED BY M. L. Edwards- l' TELEPHONE (402) 533-6929' REPORT MONTH October 1996 I M XDAte) iType}'" ~(DArat on..iRegson . Method $fl Ucensee $ysierhj f Comoonent ECAbseEConechve / @^

.. )(Hours); iShuthng [Eventi. [Codey, (Code $: Action;tol "

w . , t

.~ C

~ bpm R-  ;

1 Reactor35 i

l 96-06 961004 S 650.9 C 1 N/A ZZ ZZZZZZ A power reduction commenced on October 4. 1996 in preparation for a scheduled refueling outage. The power reduction began at 1529 hours0.0177 days <br />0.425 hours <br />0.00253 weeks <br />5.817845e-4 months <br /> '

and the generator was taken offline -:

at 2207 hours0.0255 days <br />0.613 hours <br />0.00365 weeks <br />8.397635e-4 months <br />. On October 5. 1996 '

at 0339 hours0.00392 days <br />0.0942 hours <br />5.605159e-4 weeks <br />1.289895e-4 months <br />, the reactor was made subcritical . The reduced power level prior to the outage was.a  !

planned process, which successfully

reduced reactor coolant system i (RCS) radionuclide concentrations  !

and associated outage dose. j l

1 2 3 4 i 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 i H-Other (Explain) Exhibit H --Same Source (9/77)

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L Attachment IV

. .- Refueling Information Fort Calhoun Station Unit No. 1 Report for the month ending: October 31. 1996  :

1. Scheduled date for next refueling shutdown. Shutdown - October 5, 1996 ,

2. Scheduled date for restart following refueling, November 17, 1996

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

l l a. -If answer is yes, what, in general, will 1) Allow use of source 'adterial these be? for fuel assemblies.

2) Allow use of ZIRLOS fuel cladding. .

b. If answer is no, has. the reload fuel l design and core configuration been N/A  !

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

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

a

4. Scheduled date(s) for submitting proposed 1) Amendment 177 was received on )

licensing action and support information. October 7, 1996.  !

2) Amendment 178 was received on November 1, 1996.

l S. Important licensing considerations associated with refueling, e.g., new or different fuel For Cycle 17, the new fuel batch design or supplier, unreviewed design or will contain mid-grid design changes l.

performance analysis methods, significant changes to mitigate future grid-to-rod in fuel design, new operating procedures. fretting failures.

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6. The number of fuel assemblies: )

a) in the core 133 Assemblies  ;

j. b) in the spent fuel pool 618 Assemblies 4 c) spent fuel pool storage capacity 1083 Assemblies l

l 7. The projected date of the last refueling that can i be discharged to the spent fuel pool assuming the 2007 Outage i present licensed capacity.

Prepared by: 'NdN- fV Date: / ///f /%

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