NRC-90-0021, LER 90-001-00:on 900108,annunciator 2D5 Alarmed for Div II ECCS Testability Logic/Power Failure.Caused by Blown Fuse. Fuse Installed During Course of Troubleshooting W/No Subsequent Problems noted.W/900207 Ltr

LER 90-001-00:on 900108,annunciator 2D5 Alarmed for Div II ECCS Testability Logic/Power Failure.Caused by Blown Fuse. Fuse Installed During Course of Troubleshooting W/No Subsequent Problems noted.W/900207 Ltr

ML20011E384
Person / Time
Site:	Fermi
Issue date:	02/07/1990
From:	Orser W, Pendergast J DETROIT EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-90-0021, CON-NRC-90-21 LER-90-001, LER-90-1, NUDOCS 9002130321
Download: ML20011E384 (6)

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February 7, 1990 NRC-90-0021 ,

.U. S. Nuclear Regulatory Commi ss ion j

. Attention:. Document Control Desk Washington, D.C. 20555

Reference:

Fermi 2 NRC Docket No. 50-341

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Facility Operating License No. NPF-43

Subject:

Licensee Event Report (LER) No. 90-001-00.- .

Please find enclosed LER No. 90 001-00, dated February 7,  ;

1990, for a reportable event that occurred on January 8, 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 at (313) 586-1682.

Sincerely, W -

Enclosure:

NRC Forms 366, 366A  ;

cc: A. B. Davis- <

J.-R. Eckert .!

R. W. Defayette/W. L. Axelson W. G. Rogers J. F. Stang Wayne County Emergency Management Division q

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'It 14. Blown Fuse an Testability Cabinet H21-P083 Caused Entry Into Tuchnical Specification 3.0.3 EVENT DAf t (4) LIn $sunauta ISI REPORT DATE til oTHt h , ACatlTIES INVotVED 10.

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]* 01 5 0 11 91 0 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 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 thit, 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.

Several potential causes are being investigated. Determination of final corrective actions will depend upon the results of the on-going investigation. A Supplement to this LER will be submitted after the investigation is completed.

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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 Energency Core Cooling System (B) testability logic / power failure. The operations personnel investigated and discovered that fuse (FU) B21-F2B, which is the power fuse for card file Z2 in testability' cabinet (PL) N21-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.

Due to the instrument trip units (ET) affected and the associated parameters affected for Reactor Steam Dome Pressure (PI), Drywell Pressure (PI), Reactor Vessel Water Level (LI), and Reactor Vessel Pressure-(PI), Technical Specification.3 0 3 was entered and an Unusual 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 Energency Core Cooling Systems, Alternate Rod Insertion (ROD) (ARI), Anticipated Transient Without Scram Systems (JC) and the Safety Relief Valves (RV) (SRV) Low-Low 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 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 i 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 arproximately 40 percent.

Cause of Event:

Troubleshooting activities are underway to determine the root cause.

Thnse activities under review include failure of a capacitor within be trip unit, possible hot components or junction points (infrared wacks), large amount of noise or ripple from the power supply and analysis of the blown fuses. An action plan was drawn up and all of these potential causes have been or are being investigated.

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i Analysis'of Event: .;

I 1his event rendered selected Division II instrumentation inoperable.- g This instrumentation provides actuation input to both Divisions of the i 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 l- l (1) Reactor Steam Dome Pressure o Div I/II Core Spray System .i (BG) (CS) valve permissive o Div I/II. Low Pressure Coolant (

Injection (BN) (LPCI) valve permissive (2) Drywell Pressure o Automatic Depressurization '

System logic B l- o Div II LPCI Loop Select logic  ;

o Div I/II CS .;

o Div I/II LPCI o Div I/II Residual Heat Removal' I I

(RHR) (BO) i o Div I/II Torus Cooling (BT) I o Reactor Core Isolation Cooling l

(RCIC) (BN) o High Pressure Coolant (HPCI)

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(3) Reactor Vessel Water Level o (High Level) 1 o Feed Water Turbine Trip Level 8 (High Level)

(4) Reactor Vessel Pressure o ARI o ATWS o 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 ve sel 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 provide adequate core cooling.

Thus, the' Division I instrumantation 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:

Investigation is continuing into possible root causes. Specifically, infrared checks of the cabinet were made to determine if high temperature components (from high current) were present and no abnormalities were identified. Honitoring of the power supply outputs was performed to determine if excessive noise or ripple spikes were

, present and nothing considered excessive was found. The cards in the cabinet are being visually checked and the suspect-capacitors (CAP) are being seasured for DC resistance, capacitance and dissipation factor (in-circuit)'to determine if there are any failed capacitors present on the card's power supply input.

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0l 0 15 of 15 van ha .w =c e muuim Troubleshooting to date has revealed a failed capacitor in the suspect position (C25) on Trip Unit B21-N694D. Evidence of overheating (cracked / burned) and abnormal readings were observed. 1his evidence indicates that the physical cause for the fuse blowing was from a short circuit through the capacitor. This short evidently cleared itself by physically damaging the capacitor and clearing the original fault conditioh. Investigation as to the cause of the capacitor failure is ongoing at this time.

Fuse analysis has preliminarily shown that both fuses failed from a similar over-current condition.

Determination of final corrective actions will depend upon the results of the on-going investigation. This investigation is tentatively scheduled to be completed by March 30, 1990. A Supplement to this LER will be submitted after the investigation is completed.

Previous Similar Eventat 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 Datat Data will be provided upon determination of the root cause.

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