05000341/LER-1990-001, :on 900108,blown Fuse in Testability Cabinet H21-P083 Caused Entry Into Tech Spec 3.0.3

:on 900108,blown Fuse in Testability Cabinet H21-P083 Caused Entry Into Tech Spec 3.0.3

ML20058A370
Person / Time
Site:	Fermi
Issue date:	10/19/1990
From:	Orser W, Pendergast J DETROIT EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-90-0160, CON-NRC-90-160 LER-90-001, LER-90-1, NUDOCS 9010260037
Download: ML20058A370 (7)

LER-1990-001, on 900108,blown Fuse in Testability Cabinet H21-P083 Caused Entry Into Tech Spec 3.0.3

Event date:
Report date:
3411990001R00 - NRC Website

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s e V ce resident Detroit r-u I S O n LT S,se[ M k,"g'.7 le M 10CFR50.73 Nucleat ne me oe no -

October 19, 1990 NRC-90-0160 U. S. Nuclear Regulatory Commission i

Attention: Document Control Desk Washington, D.C.

20555 1

Reference: Fermi 2 NRC Docket No. 50-341 l

NRC License No.-NPF-43 i

Subject:

Licensee Event Report (LER) No. 90-001-01 l

Please find enclosed LER No."90-001-01, dated October 19, 1990,.

for a reportable event that occurred on January 1, 1990.

A copy of this LER is also being sent to the Regional Administrator, USNRC Region III.

If you have any questions, please, contact Joseph Pendergast,-

)

Compliance Engineer, at (313) 586-1682.

Sincerely l

Enclosure: NRC Forms 366, 366A cc:

A. B. Davis J. R. Eckert R. W. DeFayette W. G. Rogers J. F. Stang Wayne County Emergency Management Division 9010260037 901019 PDR ADOCK 050 41

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On January 8,1990, at approximately 1410 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.36505e-4 months <br />, annunciator ( ANN) 2DS alarmed for Division II Emergency Core Cooling System testability logic / power failure. The operations personnel investigated and discovered that fuse B21-F2B, was blown. The fuse was replaced at 1442 hours0.0167 days <br />0.401 hours <br />0.00238 weeks <br />5.48681e-4 months <br />, but blew again approximately ten minutes later.

Technical Specification 3 0 3 was entered due to various instrument trip units affected and the inability to place them in the required tripped conditions without causing actuations/isolations.

An Unusual Event was declared at 1510 hours0.0175 days <br />0.419 hours <br />0.0025 weeks <br />5.74555e-4 months <br />.

A third fuse was installed during the course of troubleshooting with no subsequent problems noted. Channel Functional surveillances were performed to verify the operability of the trip units involved in this event. All of the trip units were found to be operable and the Unusual Event was terminated at 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br />.

Reactor Power had been reduced to approximately 40 percent.

Subsequent investigation revealed a failed capacitor on a trip init I

which shorted the power supply causing the fuses to blow. F ai'.ure I

analysis of the suspect capacitor on the trip unit was completed.

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The analysis did not reveal a root cause. Fermi 2 Engineering stsff I l

concluded that the failure of the capacitor is random in nature.

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Initial Plant Conditions

Operational Condition: 1 (Power Operation)

Reactor Power: 100 percent Reactor Pressure: 1010 psig Reactor Temperature: 530 degrees Fahrenheit Description of the Event:

On January 8,1990, at approximately 1410 hours0.0163 days <br />0.392 hours <br />0.00233 weeks <br />5.36505e-4 months <br />, annunciator ( ANN) 2D5 alarmed for Division II Emergency Core Cooling System (B) testability logic / power failure. The operations personnel investigated and discovered that fuse (FU) B21-F28, which is the power fuse for card file Z2 in testability cabinet (PL) H21-P083 was blown. The fuse was replaced at 1442 hours0.0167 days <br />0.401 hours <br />0.00238 weeks <br />5.48681e-4 months <br />, but blew again approximately ten minutes later.

Dt:e to the instrument trip units (ET) affected and the a*ssociated parameters affected for Reactor Steam Dama Pressure (PI), Drywell Pressure (PI), Reactor Vessel Water Levcl (LI), and Reactor Vessel Pressure (PI), Technical Specificatica 3 0 3 was entered and an Jnusual Event was declared at 1510 hours0.0175 days <br />0.419 hours <br />0.0025 weeks <br />5.74555e-4 months <br />. The affected trip units could not be placed in the trip condition without causing actuation / isolation of the affected systems. The affected systems, which included the Emergency Core Cooling Systems, Alternate Rod Insertion (ROD) (ARI), Anticipated Transient Without Scram Systems

( ATt'S)(JC) and the Safety Relief Valves (RV) (SRV) Low-Low l

Setpoint, were declared inoperable and a reactor shutdown commenced.

No problems were found within the H21-P083 Z2 cardfile or trip units in troubleshooting immediately following the event. A third fuse was instal'ed during the course of troubleshooting with no subsequent problems ncted. Channel Functional surveillances were performed to verify the operability of the trip units involved in this event. All of the trip units were found to be operable and the Unusual Event was terminated at 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br />.

Reactor Power had been reduced to approximately 40 percent.

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Cause of Event

Subsequent investigation revealed a failed C25 capacitor on B21N694D l which caused a short circuit on panel H21-Po83 Z2 power supply.

I Failure analysis of the C25 capacitor on B21-N694D was completed by 1

i the vendor. The analysis did not reveal a cause for the failure of l

the C25 capacitor. The failure analysis did state there had been a 1

dielectric breakdown, as had been expectet by Fermi 2 engineers.

It I was therefore concluded that this type of failure was random in I

nature. The capacitor burned open, clearing the fault and allowing _ l Z2 to be re-energized.

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Analysis of Event

This event rendered selected Division II instrumentation inoperable.

This instrumentation provides actuation input to both Divisions of the ECCS, one Division (Div) of ATWS, ARI and SRV Low-Low set logic circuits and trip circuits for the Main Turbine (TA) and Feedwater Turbines (SJ). The following lists the instrumentation parameter that was declared inoperable and the affected system / function.

Parameter System / Function (1) Reactor Steam Dome Pressure o

Div I/II Core Spray System (BG) (CS) valve permissive o

Div I/II Low Pressure Coolant Injection (BM) (LPCI) valve-permissive (2) Drywell Pressure o

Automatic Depressurization System logic B j

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Div I/II CS o

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Main Turbine Trip Level 8 (Hight.evel) o Feed Water Turbine Trip Level 8 (High Level)

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ARI o

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SRV Low-Low set and scram Pressure Interlock There were no safety consequences to this event.

It is within-the bounds of a loss of the Division II DC battery (BAT) supply which is an analyzed event as discussed in Chapter 15, Section 15 of the Updated Final Safety Analysis Report (UFSAR).

As described in the UFSAR, the loss of an Engineered Safety Feature (ESP) Div. II DC power system would remove the HPCI system should an incident occur in which the vessel did not depressurize. The HPCI system is, however, backed up by the Automatic Depressurization System (ADS), which has its' power supply from the ESF Div. I DC system.

Thus, in the event of the loss of all ESP Div. II DC power, the ADS would depressurize the reactor, and the LPCI and CS systems would

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Thus, the Division I instrumentation was available to perform the same functions as the failed instrumentation in that is was capable of initiating both divisions of ECCS equipment.

Corrective Actions

Infrared checks of the cabinet were made to determine if high I

temperature components (from high current) were present.. No I

abnormalities were identified. Monitoring of the power supply I

outputs was performed to determine if excessive noise or ripple i

spikes were present. Nothing considered excessiva was found. The I

cards ir, the cabinet were visually checked and the suspect I

capacitors (CAP) were seasured for DC resistance, capacitance and I

dissipation factor (in-circuit) to determine if there were any l

failed capacitors present on the card's power supply input.

l Troubleshooting revealed a failed capacitor position (C25) on Trip I

Unit B21-N694D. Evidence of overheating (cracked / burned) and I

abnormal readings were observed. This evidence indicated that the l

physical cause for the fuse blowing was a short circuit through I

the capacitor. This short physically damaged the capacitor, thus I

clearing the original fault condition.

1 Failure analysis of the (C25) capacitor on B21-N694D was completed byl the vendor. The analysis did not reveal a cause for the failure of l

3 the capacitor. The failure analysis did state that a dielectric l

breakdown had occurred. But the cause for this dielectric breakdown I could not be established.

It can be concluded that the failure of I

i the affected capacitor is random in nature. A review of the Nuclear l Plant Reliability Data System for capacitor failures on Rosemount I

trip units revealed that two previous capacitor failures had been I

reported with only one being reported 'directly as a C25 capacitor.

I Both were on model Rosemount 710 DU Trip Units, i

Fuse analysis showed that both fuses blew from a similar over-currentl condition.

No other deficiencies were noted.

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Previous Similar Events

Licensee Event Report 87-012 " Inoperable High Pressure Cooling Injection and Reactor Core Isolation Cooling Due to Blown Power Supply Fuse" described a similar event.

Failed Component Data:

Kemet, Capacitor, Molded Epoxy Case, Monolithic Ceramic Chip, Radial Leads, C062K105K5X5CA.

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