ML20099G632

Monthly Operating Rept for Jul 1992 for FCS Unit 1

ML20099G632
Person / Time
Site:	Fort Calhoun
Issue date:	07/31/1992
From:	Edwards M OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20099G620	List: ML20099G617 ML20099G632 ML20099G638
References
NUDOCS 9208170138
Download: ML20099G632 (8)

Text

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OMAHA PUBLIC POWER DISTRICT fort Calhoun Station Unit No.1 JULY 1992 Monthly Operating Report OPERATIONS

SUMMARY

Fort Calhoun Station (FCS) operated at full power Ju13 On July 2, a design deficiency was discovered whereby a single failui:: in the wiring of Heater Drain Pumps FW-5A/B/C could cause a fire in both switchgear rooms.

Consequently, a one-hour notification to the NRC was made pursuant to 10 CFR 50.72.

As a compensatory measure, hourly fire watches were instituted. To correct the potential problem, cable upgrade and rerouting is currently in progress, At 2335 hours0.027 days <br />0.649 hours <br />0.00386 weeks <br />8.884675e-4 months <br /> on July 3, during replacement of a degraded circuit board in non-safety related inverter No. 2, power was momentarily lost to the instrument bus that supplies power to the turbine electrohydraulic control system, causing tne turbine control valves to close. The closing of the turbine controT valves and resulting large mismatch between reactor power and steam demand caused a sharp increase in reactor coolant system (RCS) temperature and oressure.

At 2336 hours0.027 days <br />0.649 hours <br />0.00386 weeks <br />8.88848e-4 months <br /> the reactor protection system automatically tr'ipped the reactor due to high pressurizer pressure.

Subsequently, primary safety valve RC-142 failed open resulting in high pressure in the pressurizer quench tank. The pressure in the quench tank caused the tank's rupture disk to burst at 2355 hours0.0273 days <br />0.654 hours <br />0.00389 weeks <br />8.960775e-4 months <br />, effecting the loss of approximately 21,500 gallons of RCS water to the containment building sump.

An ALERT emergency classification was declared at 2352 hours0.0272 days <br />0.653 hours <br />0.00389 weeks <br />8.94936e-4 months <br /> on July 3, due to a challenge to a fission product barrier (i.e., the RCS pressure boundary /depressurization

)

The plant was stabilized and a controlled, expeditious cooldown was commenced to bring the plant to cold shutdown.

At Ot. O hours on July 4, the event was downgraded to a Notification of Unusual Event (NOUE).

The plant was placed on shutdown L

cooling at 1312 hours0.0152 days <br />0.364 hours <br />0.00217 weeks <br />4.99216e-4 months <br />, and at 1825 hours0.0211 days <br />0.507 hours <br />0.00302 weeks <br />6.944125e-4 months <br /> FCS entered cold shutdown.

At 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br /> the plant downgraded from the NOVE, terminating the emergency.

Additional details of this event are contained in Licensee Event Report No.92-023.

OPPD's recovery organization developed a 28 point Recovery / Restart Action Plan.

Both primary safety valves were removed on July 7 and sent to an offsite laboratory for testing and recalibration.

The root cause of the RC-142 safety valve failure was determined to be an adjusting bolt nut which apparently vibrated loose during the original pressure transient and allowed the adjusting bolt to back out.

The change of position of the adjusting bolt effectively lowered the safety valve setpoint to approximately 1923 psia, causing the valve to lift during the normal post-trip repressurization.

Both non-safety related inverters were modified to provide an additional testing power source to the non-safety buses, so that the inverters can be removed from service and tested prior to their return to service. Another modification added a turbine trip whenever turbine control valve No. I closes, with provisions to bypass this trip under low power conditions.

Also, positive mechanical locking devices for both pressurizer safety valve adjusting bolts were added.

@ AN R

i l

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y s

' Monthly Operating Report LIC-92 287R Page 2 OPERATIONS

SUMMARY

(continued)

Actions specified in the Recovery / Restart Action Plan were completed and reactor criticality was achieved at 2052 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.80786e-4 months <br /> on July 22. The generator was. synchronized to the grid at 0610 hours0.00706 days <br />0.169 hours <br />0.00101 weeks <br />2.32105e-4 months <br /> on July 23.

Power was maintained at 30% for chemistry holds until July 25, when power was raised to 90%.

Currently, FCS is operating at 100% power.

The following NRC Inspections took place during July:

IER No:

Title 92-14 Monthly Resident Inspection 92-18 Augmented Inspection Team 92-16 Generic Letter 88-17 The following LERs were submitted during July:

If.B Description 92-20 Failure to Obtain Grab Sample During Radiation Monitor Inoperability 92-21 Failure to Initiate a Fire Watch for an Inoperable Fire Door A.

SAFETY VALVES OR P0P.V CHALLENGES OR FAILURES WHICH OCCURRED

~FCS experienced the previously described event which challenged both PORVs and one safety valve-(RC-142). Both PORVs opened as designed on the loss of load event and both reseated properly. _ One of the two primary safety valves (RC-142) opened during the initial pressure transient of the event, closed, then reopened but did not properly re:::at.

Surveillance testing was performed before plant' restart to demonstrate operability.

B.

RESULTS OF LEAK RATE TESTS Due to the July 3 reactor trip, only 13 RCS leak rate tests were performed in July. Reactor power and Xenon concentration were changing during many of the leak rate te,ts, reducing the accuracy of the results. During the periods of relative stability, the total RCS leak rate was about 0.12 gpm, composed ef' approximately 0.03 to 0.06 gpm of "Known" leakage to the reacter coolant drain tank with the balance being " Unknown" leakage.

Early in July,harging-pump discharge valvehigher than normal leak rates were attrib CH-192 was closed and RCS leakage returned (CH-192) packing. Therefore, from the "B" c

to normal until the July 3 plant trip. After power and Xenon levels stabilized following startup on

' July 29,' the RCS leak rate tests again yielded normal values.

l Monthly Operation Report LIC-92-287R Page 3 C.

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

Description None D.

SIGNIFICANT SAFETY RELATED MAINTENANCE FOR THE MONTH OF JULY The following transmitters were inspected for water / foreign material intrusion:

A/DPT-ll4X (RCS Loop 18 Cold Leg Differential Pressure)

C/DPT-ll4X (RCS Loop 1 Hot le and 1A Cold leg Differential Pressure Transmitter)g B/PT-120 (Pressurizer RC-4 Pressure Transmitter)

C/PT-120 (Pressurizer RC-4 Pressure Transmitter)

Tne following transmitters were inspected and calibrations performed:

PT-105 (Wide Range Pressure Transmitter for RC-4)

C/DPT-ll4W (RCS Loop 1 Hot Leg and IB Cold leg Differential Pressure)

A/LT-911 (Steam Generator RC-2A Wide Range level Transmitter)

A/PT-913 (Steam Generator RC-2A Wide Range Pressure Transmitter)

LT-2904X (Safety injection Tank SI-6A Wide Range level Transmitter)

FT-328 (Low Pressure Safety Injection to RC Loop IB Flow Transmitter)

FC-313 (High Pressure Safety Injection to RC Loop 1B Flow Transmitter)

LT-595 & LT-600 (Containment Sump Level Transmitters)

PT-130 & PT-131 (Pressurizer Quench Tank RC-5 Narrow Range Pressure Transmitters)

LT-132 (Pressurizer Quench Tank Level Transmitter)

PT-3194 (Reactor Coolant Pump RC-3B Gasket Housing Pressurizer Transmitter)

Monthly Operating Report LIC-92-287R Page 4 D.

SIGNIFICANT SAFETY RELATED MAINTENANCE FOR THE MONTH OF JULY (continued)

Inspected and tested ilCV-150 and HCV-151 (Power Operated Relief Isolation Valves for Pressurizer RC-4).

Removed, refurbished and reinstalled RC-141 and RC-142 (Pressurizer RC-4 Code Safety Valves).

Repaired leaking flange gasket on RC-142.

Replaced rupture disc on RC-5 (Pressurizer Quench Tank).

Inspected limit switches and meggered the motor of HCV-314 (HPSI to RC Loop 1A Isolation Valve) and HCV-327 (LPSI to RC-Loop 1B Isolttion Valve).

Performed setpoint and leakage test of AC-336 Charging Pump CH-1A 011 Cooler Component Cooling Water Inlet Relief Val (vQ and AC-337 (Charging Pump CH-1B Oil Cooler Component Cooling Water Inlet Relief Valve).

Valve) and install (ed an approved replacement. Component Cooling Heat Exchanger AC-1B R Removed HCV-2881B Obtained resistance readings for each control element drive mechanism clutch coil.

Replaced 52/STA and 52/)HH switches for 4160V cubicle / breaker IA4-9 (Feeder for Transformer TIB-48.

Replaced 94-4/1045 relay (Auxiliary Feed Water Pump FW-10). Time Delay Relay in Auto Start Ci 1045A and Steam Feed for Re laced packing cooling water pump and motor for CH-1B (Charging Pump 1B.

Replaced FW-1444 (Auxiliary Feed Water to RC-2B Header Relief Valve).

1 1

OPERATING DATA REPORT DOCKET NO.

50-285 UNIT FORT CALHOUN STATION DATE XUdtT5T' 66~f9~92 COMPLETED BY M.

L.

EDWARDS LOPERATING STATUS TELEPHONE TT6~ ) 636-7T51 2

1. Unit' Names FORT CALHOUN STATION

2. Reporting-Period JULY 1972-~

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): ^~f012]

l7.- Max. Dep. Capacity (Net MWe):

478

8. !f 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 ll'--Hours in Reporting Period...........

744.0 5111.0 165241.0

-12.: Number of. Hours Reactor was Critical 290.7 2458.9 127277.l'

13. Reactor Rese rve Shutdown Hours......

.0

.0 13097~5~

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

2 Ell. 4 88~6TT 12576~37 x15. Unit Reserve Shutdown: Hours.........

.0

.0

.0 16.-Gross Thermal Energy Generated (MWH)-

334189.7 f0eT3T1Ta' 164708087.3

~'T764109.2 ~

17. Gross Elec.' Energy-Generated (MWH)..

108894.0 1077983.0 5

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

102660.1 974437.0 51708188.4

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

37T5 46TT 7T.T

20.; Unit Availability-Factor............

37.8 46.7' 76.1,

21. Unit-Capacity Factor (using MDC Net)

28. I 39.9 68.1 R22. Unit Capacity Factor (using DER Net) 28.9-39.9 66.2

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

6_2.2 17.7

4. 2'

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:

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

Forcast Achieved INITIAL CRITICALITY INITIAL ELECTRICITY N/A COMMERCIAL OPERATION

!! o '.:

.1 AVERAGE DAILY UNIT POWER LEVEL DOCKET NO.

50-285 -

UNIT YdRY~CXLIf00N' STK' TION DATE' XUGlisT.

OT~15if2 COMPLETED BY-M.

L.

EDWARDS TELEPHONE TT0f) ir36-2Tn MONTH JULY-1992

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

(MWe-Net)

-1.

476 17 0

2 477 18 0

3 469 19 0

4 0-20 0

5

-0 21 0

6 0

22 0 0 23 48 8

0-24 91 9

' 0 --

25 138 10--

0 26 357 0 27 428 12-

-0 28 427

<13 0'

29 430

.14 30 466 15 0

31 470 16-0-

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.

Refueling Information Fort Calhoun - Unit No. I

~ Report for the month ending July 1992

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?

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 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.

J1ng 1993 i

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 New fuel supplier New LOCA anaivsis

6. -The number of fuel assemblies:

a) in the core 133 Assembl<es b) in the spent 529 Assembl-es fuel pool c) spent fuel pool storage capacity 729 Assemblies d) planned spent-Planned to be fuel pool increased with higher storage capacity density spent fuel racks.

7. The projected date of the last refueling that can be discharged to the spent fuel pool assuming the

.present licensed capacity.

1995*

O Capability of full core offload of 133 asser'411es lost. Reracking to be performed between

-the.1993 and 1995 Refueling Outages.

Prepared by 14 MWR Date

! +7 2-

UNIT SHUTDOWNS AND POWER REDUCTIONS' DOCKET NO. 50-285 UNIT NAME Fort C9houn St,'

DATE August 6. 1992 COMPLETED BY M. L. Edwards TELEPHONE (402) 636-2451 REPORT MONTH July 1992 No.

Date Type' Duratkm Reason' Method of Licensee System Comporwin Cause & Corrective (Hers)

Shutting Evera Code' Code' Action to Down Reactos' Report i Prevera Recurrence 924)5 920703 F

462.6 B

3 92423 ED GENERA On July 3.1992, at 2336. wlule the plant was cyerating at 100% power, the Reactor Pregection System automatically tripped the reactor due to high pressuriter pressure. The event was initiated as a resuh of maintenance on a non-safety related inverter. D=2 ring replacemers of a degraded circuit tmard, power was momersarily lost to the instrument bus that inspplies power to the Tudsne Electruhydraulic Control System, resuhing 'm cloatre of the turbine control valves. A subsequent failure of a pressurizer caie nefety valve (RC-142) resuhed in high pressuie in the pressurizer quench tank that blew the tank's rupture disk and resuhed in the kw of approximately 21,300 galkms of contaminated water to the containmets building surnr.

The consequences of the evers are bmnded by the Fort Calben Station Updated Safety Analysis Report.

The root cause of the nementary loss of power to the instrument bus was determined to be the inability to isolate and test *he non-eafety related inverters after maintenance without potentially losing power to the.

respective 120V AC instrument buses. The runt cause of the malfunction of RC-142 was determined to be the adjusting boh nut that loosened eraf a!!omed the art pressure adjusting bolt to back out.

Corrective actions included: a nextification to enhance the abiInty to test the non-safety related inverters off-line, the addition of a positive mechanical hxking device for the pressurizer safety valve adjusting bolts and coneletion of a comprehensive Recovery / Restart Actkm Plan.

I 2

3 4

F: Forced Reason:

Method:

Exhibit G - Instructions S: Scheduled A-Equ'rpmera Failure (Explain) 1-Manual for Preparution of Data B-Maintenance or Test 2-Manual Screm.

Erary Sheets for Licensee C-Refueling 3-Automatic Scram.

Evera Report (LER) File (NUREG416I)

D Regulatory Restriction 44)ther (Expla' )

m E-Operator Training A License Examination F-Administrative G4hwrational Error (Explain)

Exhibit 1 - Same Source H-Other (Explain)

(9/77) m