ML20104B610

Monthly Operating Rept for Aug 1992 for Fort Calhoun Station Unit 1

ML20104B610
Person / Time
Site:	Fort Calhoun
Issue date:	08/31/1992
From:	Cavanaugh G, Gates W OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LIC-92-297R, NUDOCS 9209150424
Download: ML20104B610 (8)

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Omaha Public Power District 444 South 16th Street Ma!I Omaha. NeDraska G8102 2247 402/636-2000 September 14, 1992 LIC-92-297R V. S. Nuclear Regulatory Commission ATIN: Document Control Desk Mail Station Pl-137 Washington, DC 20555

Reference:

Docket No. 50-285 Gentlemen:

SUBJECT:

August 1992 Monthly Operating Report (MOR)

Enclosed is the August 1992 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.

Sinterely, 4'.,) }{Xw W. G. Cates Division Manager Nuclear Operations WGG/grc Enclosures c:

LeBoeuf, Lamb, Leiby & MacRae J. L. Milhoan, NRC Regional Administrator, Region IV R. P. Mullikin, NRC Senior Resident inspector S. D. Bloom, NRC Act':g Project Manager R. T. Pearce, Combustion Engineering R. J. Simon, Westinghouse Office of Management & Program Analysis (2)

INP0 Records Center American Nuclear Insurers i

9209150424 920831 PDR ADOCK 05000285 s

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AVERAGE DAILY UNIT POWER LEVEL i

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DOCKET NO.

50-285 j

UNIT I' ORT CALHOUN SfXf1dII DATE 5tFTER~BER 08~T9Y2

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COMPLETED BY T. R. d%VKRKlJda TELEPHONE (W2 ) f3TZ2T77 MONTH AUGUST 1992 DAY AVERAGE DAILY POWER LEVEL DAY AVERAGE DAILY POWER LEVEL i

(MWe-Net)

(MWe-Net) l 4

1 471 17 477 1

2 472 18 475 l

3 4-*

19 473 i

l 4

470

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472 5:

353 21-472 l

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6 14 22 23 i

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7 302 23 0

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8 459 24 0

1 9

462 25 0

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10 461 26 0

l 11 468 27 0

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12 470 28 0-i l-13 472 29 0

j 14-474 30 0

j 15

-475 31 0

16 476 INSTRUCTIONS-2 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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OPERATING DATA REPORT DOCKET NO.

50-285 UNIT FUNFUXENOUN STATION DATE SEPTEMBER 08 1992 COMPLETED BY ii.~N. CAVANAUGH OPERATING STATUS TELEPHONE TTif2) 636 247T

1. Unit Names FORT CALHOUN STATION

2. Reporting Period:

AUGUdT 1992~

NOTES

3. Licensed Thermal Power (MWt): 1500 4

e. 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):

478

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

NA

9. Power Level to which restricted, if any (Net MWe): NA

10. Reasons For restrictions, if any:

NA l-THIS MONTH YR-TO-DATE CUMULATIVE

11. Hours in Reporting Period...........

~ 744.0 5855.0 165985.0-

12. Number of Hours Reactor was Critical 505.8 iF64.7 12T78774~

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

0 0

13697!f

- 14o Hours Generator On-line.............

487.9 2874.2 12 6~251. T

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

0'-

. 0' 0

16. Cross Thermal Energy Generated-(MWH) 700842.0 3785203.6 165408929.3

17. Gross E3ec. Energy Generated (MWH)..

271608.0 1259591.0 5443571777

18. Net Elec. Energy Generated (MWH)....

219931.7 119436877 51928120.1

19. Unit Service Factor.................

65.6 497T 76.1

20. Unit Availability Factor............

65.6 49.1~

7671-

21. Unit Capacity Factor (using MDC Net) 61.8 42.7 6'8. 0

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' 23. Unit Capacity Factor (using DER Net) 61.8_

42.7 66.2

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

34.4.

21.1 4.4

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

NONE.

25. If shut down at end of report period, estimated date of startup: 09/05/92

26. Units in test status (prior to-comm. oper.):

Forecast Achieved L

INITIAL CRITICALITY INITIAL ELECTRICITY N/A COMMERCIAL OPERATION

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

DATE September 9. 1992 COMPLETED BY G. R. Cavanauch TELEPHONE (402) 636-2474 REPORT MONTH August 1992 No.

Date Type' Duratica Ream m' Medm.J of Licenace System Cary ment Canne & Correctise (Hours)

Shutting Evers Cnde' Code' Actum >>

Down Reactor' Repor*#

Prevers Recurrerte 9244 920M05 F

17.8 A

4 N.me EJ BRK On August 5,1992, uhde the plans was operating er 1(W)% pawer, dunng a rmwmai preventat.ve mairder.fm e setmey a terning edit was traced ks the main switch of AI 41 A (IIS VDC BUS-1 Panet). He switch was found to be ins M the piuch acd had visual indicatiema of heat damage. Due tre the adverse consequern:es of a p=4ential loss (f DC peer at 100% pewer. the decision was made km do the repairs off-ime. A power reductim aas inivisted at 1250 hours0.0145 days <br />0.347 hours <br />0.00207 weeks <br />4.75625e-4 months <br /> on August 5, and the turbine was taken c4T4ine at 3

2250 hours0.026 days <br />0.625 hours <br />0.00372 weeks <br />8.56125e-4 months <br />. Reactor power wm then reduced to apprmmastriy 1 % before install'ng a junger to keep the DC penel energued while repairs were made. He gener hw was put back cre!% on Aug+.se 6. at 1639 hours0.019 days <br />0.455 hours <br />0.00271 weeks <br />6.236395e-4 months <br />.

924)7 920822 F

238.4 A

3 LER-92424 EC IX Failure of a power surply in the Electrwbydraulic Contnd System caused the turbine corarol valv s km partially close.. His caused the Reackw Coolars System pre sure so increase. Ahlmmgh the pre **ure did mit reach the setpoire, RC-142 lined prestnrurely to relieve pressure. He valve l

rescated but pressure drtyped >> the reeckw trip setroira Sw Dermai l

Margin /tew Pressure armi en automanc reachw tr p occuned. De plant rsmained shut &mn for the remaimler of the tr=mth.

1 See LER-924T.'8 f(v further octaile.

I 2

3 4

F: Forced Reamm:

Metimmi:

Exhibit G - Instroms 5: Scheduled A.Equipmers Failure (Explain) 1-Manuct nic Preparatum cf Data B Maintenance or Test 2 Manual Scram.

Erary Sheets f.= Licensee C-Refbehng 3-Automatic Scrsra.

Event Regert (LER) File (NUREG416I)

D-Regulatory Remrictkm LOther (Explain)

IAyrator Training & License Exa.runatkm F-Administrative Es.hibit 1 - Same Sese G4)peradonal Erruc (Explaia)

II4Xber t' Explain)

(9/77)

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Refueling Information fort Calhoun - Unit No. 1 Report for the month ending Auaust 1992

1. Scheduled dated for next refueling shutdown.

Sentenber 1993

2. Scheduled date for restart follos ng refueling.

November 1993

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

_Jes

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 question: 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 c' mes in fuel design, new operating procedurt New fuel supplier New l.CCA analysis

6. The number of fuel assemblies:

a) in the core q33Assembl<es b) fuel pool n: the spent 29 Assembl' es

(

c) spent fuel pool storage capacity 729 Asiemblies d) planned' spent Planned to be fuel pool increased with higher storago capacity density-spent-fuel racks.

7. The )rojected date of the last refueling that can be disciarged to the spent fuel pool assuming the present licensed capacity.

1995*

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

Prepared by MM

'Date~

ci./'L,/9-7

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4 OMAHA PUBLIC POWER DISTRICT Fort Calhoun Station Unit No. 1 AUGUST 1992 Monthly Operating Report I.

OPERATIONS

SUMMARY

Fort Calhoun Statio, (FCS) operated at nominal full power August I through August 5.

During a normal preventive maintenance activity, a burning odor was noted within one of the Control Room panels.

The odor was traced to the main switch of Al-41A (dications of heat da) mage.

125 VDC BUS-1 Panel.

The switch was hot to the touch and had visual in Due to the adverse consequences of a potential loss of DC power at 100% power, the decision was made to do the repairs off-line.

A power reduction was initiated at 1250 hours0.0145 days <br />0.347 hours <br />0.00207 weeks <br />4.75625e-4 months <br /> on August 5,1992, and the turbine was taken off line at 2250 hours0.026 days <br />0.625 hours <br />0.00372 weeks <br />8.56125e-4 months <br />.

Reactor ~ power was then reduced to approximately 1% before installing a jumper to keep the DC panel energized while repairs were made.

Following repairs, the turbine-generator was synchronized to the grid at 1639 hours0.019 days <br />0.455 hours <br />0.00271 weeks <br />6.236395e-4 months <br /> on August 6, 1992.

During synchronization of the turbine generator on Augest 6, tne 3451-5 Breaker (one of two breakers connecting FCS to the 345 kV grid) would not close.

An investigation revealed.3 defective coils.

Repairs were successfully completed and the breaker was reclosed at 1540 hours0.0178 days <br />0.428 hours <br />0.00255 weeks <br />5.8597e-4 months <br /> on August 13, 1992.

Full power operation continued until 0152 hours0.00176 days <br />0.0422 hours <br />2.513228e-4 weeks <br />5.7836e-5 months <br /> on August 22, 1992.

Failure of a power supply in the Electro-Hydraulic Control caused the turbine control valves to close from approximately(EHC) System 40% open to 22% ope.e.

The primary to secondary load mismatch caused reactor coolant system pressure to increase to approximately 2398 asia setpoint is <2400 At approximately 2398 psia,1C-142(reactor trip code safety valve, psia).and relieved pressure. point appro set prematurely The valve reseated, but pressure dro d to the reactor trip setpoint f or Thermal Margin / Low Pressure (T

) and an automatic reactor trip occurred.

RC-14t is the same valve ch lifted prematurely and did not properly reseat during the July 3,1992 trit (see July 1992 Monthly Operating Report).

The -plant was maintained < n hot shutdown while in situ testing of RC-141 and RC-142 was ccmpleted.

Results of this testing indicate that RC-142 was lifting ap roximately 100 psia early. In situ testing was also conducted for RC-141 the other code safety valve with.similar results.

temperature i)s a significant contribctor to the setpoint and consequently,OPPD previous setpoints may have been inappropriately low.

After concludirg the in situ testing, the plant commenced a cooldown, initiating shutdown cooling at 1059 hours0.0123 days <br />0.294 hours <br />0.00175 weeks <br />4.029495e-4 months <br /> on August 27, 1992.- RC-141 and RC+142 were removed on August 28, 1992 and shipped off-site to Wyle Laboratories for inspection and additional testing. Further details of the testing and the root cause will be detailed in LER-92-028.

The pressurizer manway was removed and RCS level was reduced to midloop for Reactor Coolant Pump RC-3D seal replacement on August 30, 1992.

. Attachment LIC-92-297R Page 2 The

.Pety valves were returned to the site on August 31, 1992 and were reirstailed September 1,1992.

The failed EHC System power supply was re,oved.

The EHC System was modified to receive its power directly from the power bus within Panel Al-50 (where the EHC is located). Panel Al-50 receives its power from the Permanent Magnet Generator with back-up from 1

the Station Ic0 VAC.

j The following NRC Inspections took place during August 1992:

IER No.

11tli 92-15 Monthly Resident inspection 92-17 E0P Follow-up Inspection 92-21 August 22, 1992 Flant Shutdown Special Inspection The following LERs were submitted during August 1992:

LER No.

LER Data Descriq11gJ1 92-022 08/03/92 Inadequately Sized Heater Drain Pump Cables92-023 08 03/92 Reactor Trip Due to inverter

/

"alfunction and Subsequent Pressurizer Safety Valve Leak 92-024 03/17/92 Failure to Comply with Linear Heat Rate Technical Specifications During Alarm Inoperability-92-025 08/21/92 Inadvertent Manual Start of Emergency Diesel Generatcr at the local Control Panel 92-026 08/24/92 Incore Detector Alarm Limits Non-Conservative for Monitoring Peak Linear Heat Rate A.

SAFETY VALVES OR PORY CHALLENGES OR FAILURES WHICH OCCURRED Code Safety Valve RC-142 lifted below its expected set 22, 1992. poi nt,

resulting in the automatic reactor trip on August

.The valve reseated after lifting.

For further details, see Section I,_

Operations Summary.

8.

RESULTS OF LEAK RATE TESTS RCS leakage during the month of August, 1992 was generally low.

During perinds of relative stability, the total RCS leakage averaged

'less than 0.1 gpm.

This leakage was composed of approximately one-half "Known leakage to the Reactor Coolant Drain Tank and one-half " Unknown" leakage.

" Unknown" leakage is the arithmetic difference between Known leakage and Total leakage.

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Attachment LIC-92-297R Page 3 On August 22, 1992, the plant experien:ed an automatic reactor trip as noted in Section 1.

This and the subsequent cooldown and depressurization resulted in anomalous leakrate test results on August 22 and August 26. With the plant in the cold thutdown mode, no leakrate tests could be performed after August 26.

C.

CHANGES, TESTS AND EXPERIMENTS REQUIRING NUCLEAR REGULATORY COMMISSION AUTHORIZATION PURSUANT T0 10CFR50.59 Amendment No.

Qq.scription 146 This amendment revises the Technical Specifications to increase the maximum allowable

!etpoints from 11% to +3%/-2% for the MSSVs and to specify lift settings for all MSSVs and for the e.wo pressurizer safety valves.

147 This amendment makes changes to Technical Specification 2.7,

" Electrical Systems",

to-correct inconsistencies and to provide further guidance on equipment necessary for the 161 KV power supply.

Additionally, foradministrative changes are incorporated Technical Specification 2.7 and Table 2-10.

D.

SIGNIFICANT SAFETY RELATED MAINTENANCE FOR THE fiONTH OF AUGUST 1992 A1TS-120 (Test switch for Channel

'A' of the diversified scram system) did not operate correctly and was replaced.

AC-189 and AC-190 discharge check valves for spent fuel pool circulation pumps AC(-5A and AC-58) were replaced.

Damage due to overheating was discovered to the main switch of Al-41A (125 VOC BUS-1 panel). A temporary modification instructed electrical maintenance to remove the main switch and install jumpers in its place.

Packing was replaced on Charging Pump CH-18.

Repairs were performed on llCV-2805B (raw water strainer AC-12B, backwash control valve).

Repairs were performed on component cooling heat exchanger AC-10 component cooling water inlet ard outlet valves HCV-492A and HCV-492B.

New supports were welded in place and new fittings installed into primary containment penetrations M 73 and M-84.

Toxic gas monitors YlT-6288A and YIT-6288B were repaired due - to evidence of water intrusion.

A 52/STA switch was installed into IA3-18 (spare #eeder in the 4160V switch gear).

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