05000272/LER-1986-001, :on 860116,reactor Trip Occurred Following Inadvertent Deenergization of Vital Bus 1A.Caused by Direct Shock to Relay.Failed Power Supply Replaced.Svc Qualification of Relays Being Investigated

:on 860116,reactor Trip Occurred Following Inadvertent Deenergization of Vital Bus 1A.Caused by Direct Shock to Relay.Failed Power Supply Replaced.Svc Qualification of Relays Being Investigated

ML20205J549
Person / Time
Site:	Salem
Issue date:	02/14/1986
From:	Rupp J, Zupko J Public Service Enterprise Group
To:	NRC OFFICE OF ADMINISTRATION (ADM)
References
LER-86-001, LER-86-1, NUDOCS 8602260069
Download: ML20205J549 (5)

LER-1986-001, on 860116,reactor Trip Occurred Following Inadvertent Deenergization of Vital Bus 1A.Caused by Direct Shock to Relay.Failed Power Supply Replaced.Svc Qualification of Relays Being Investigated

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv), System Actuation
2721986001R00 - NRC Website

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On January 16, 1986, a reactor trip occurred immediately following the inadvertent de-energization of the 1A Vital Bus.

The initiating event was the actuation of the bus differential relay which, by design, caused the 1A Vital Bus to be de-energized.

Investigation revealed that the relay actuation was caused by a direct shock, resulting from the harder than normal closing of the 1A Diesel Generator output circuit breaker cabinet door by an operator who was tagging out the diesel for routine maintenance activities.

The loss of the bus'resulted in the loss of one of the two power supplies for the rod control power and logic cabinets.

Ilo w e v e r, the remaining power supply failed due to an internal fault, resulting in the loss of all power to the logic cards.

This, in turn, caused the control rods in shutdown banks C and D to drop into the core.

The Reactor Protection System then responded as designed to initiate a rsactor trip on high negative flux rate.

The failed power supply was replaced and satisfactorily tested.

Although the seismic qualification of the differential relays is not in question, their service qualification is being investigated; i.e.,

the relays contain " telephone contacts" which are vulnerable to actuation upon shock loading.

In the interim, operators were made aware of this incident and signs were placed on the cabinet doors instructing personnel to close the doors easily.

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salen Generating Station DOCKET NUMBER LER NUMBER FAGE Unit 1 05000272 86-001-00 2 0F 4 FLANT AND SYSTEM IDENTIFICATIONl Westinghouse - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as [XX].

IDENTIFICATION OF OCCURRENCE:

Reactor Trip From 100% on High Negative Flux Rate Event Date:

01/16/86 Report Date: 02/14/86 This report was initiated by Incident Report No.86-024 CONDITIONS FRIOR TO OCCUERENCE:

/ Mode 1 - Rx Power 100 % - Unit Load 1115 MWe DESCRIPTION OF OCCUERENCE:

0n January 16, 1986, during routine power operation, IA 4KV Vital Bus [EB] de-energized.

This was followed almost immediately, at 0640 hours0.00741 days <br />0.178 hours <br />0.00106 weeks <br />2.4352e-4 months <br />, by a reactor trip.

lA Emergency Diesel Generator [EK]

was being tagged out for routine relay maintenance activities, which was the only evolution in progress at the time of the trip.

The Unit was stabilized in Mode 3 (Hot Standby), and at 0748 hours0.00866 days <br />0.208 hours <br />0.00124 weeks <br />2.84614e-4 months <br />, in accordance with the requirements of the Code of Federal Regulations, 10CFR 50.7 2(b)( 2)(ii), the Commission was notified of the automatic actuation of the Reactor Protection System [JC].

AFPARENT CAUSE OF OCCUERENCE:

The initiating event was the actuation of the 1 A 4KV Vital Bus differential relay which, by design, caused the 1A Vital Bus to be de-energized.

Investigation revealed that the relay actuation was not caused by an actual electrical fault on the bus, but instead was the result of a direct shock to the relay.

This shock was the result of a harder than normal closing of the 1A Diesel Cenerator output circuit breaker cabinet door by the operator performing the diesel tagout.

The " normal" power supplies for the rod control power and logic cabinets are fed from the rod drive motor-generator sets, while 1A Vital Bus feeds the " backup" power supplies.

The outputs of these power supplies are auctioneered, with the highest being automatically selected to carry the load.

Troubleshooting revealed that the " normal" negative twenty-four (-24) volt power supply on Rod Control Power Cabinet SCD had failed in such a manner tha t, when loaded, its output was only minus 1.4 volts.

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salen Generating Station DOCKET NUMBER LER NUMBER FAGE Unit 1 05000272 86-001-00 3 0F 4 AFFARENT CAUSE OF OCCURRENCE:'(cont'd)

However, when not loaded (when the rod control power and logic circuits were being supplied by the " backup" power supply) the

" normal" power supply voltage was reading normal.

Therefore, at the time of occurrence, cabinet 'SCD was being supplied by the " backup" power supply.

When the 1A Vital Bus de-energized, the " backup" power supply f ailed to zero, and the " normal" supply, attempting to pick up the load, went to minus 1.4 volts.

This resulted in a loss of all negative twenty-four ( -24) volt supply to the logic cards, causing the control rods in i*nhutdown banks C and D to drop into the core.

As a result, all four'(4) power range negative flux rate trips were initiated resulting in a reactor trip.

ANALYSIS OF OCCURRENCE:

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The reactor trip, on negative flux r a t e',

ensures that the minimum Departure from Nucleate Boiling Ratio'(DNBR) is maintained above 1.30 for multiple control rod drop accidenta.

The loss of power to Rod Control Power Cabinet SCD resulted in the dropping of eight (8) rods, and the Reactor Protection System responded as designed to initiate a reactor trip on high negative flux rate.

The plant is designed for four-hundred (400) trips from full power; therefore, this transient was within the design of the plant.

t As stated, the operstion of the vital bus differential relay was caused by an impact or shock load due to a hard closure ck the cubicle door (the relay is mounted on the door).

An impact load is not considered as part of the seismic criteria for thesrelay.

Seismic loading is a result of ground motion which is a low frequency vibration, whereas an impact load results in a high frequency response.

Therefore, the seismic qualification of the relay is not in question.

Concerning the loss of the 1A Vital Bus, sufficient redundant equipment remained available from the IB and 1C Vital Busses to provide the minimum required safeguards equipment to mitigate the consequences of a design basis event.

Therefore, this event involved no undue risk to the health or safety of the public.

However, because of the automatic actuatl'on of the Reactor Protection System, the event. is reportable in accordance with the Code of Federal Regulations,J10CFR 50.73(a)(2)(iv)..

CORRECTIVE ACTION _;

The failed power supply in the SCD cabinet was replaced and satisfactorily tested.

Because of the peculiar way in which this power supply failed (supplying normal output voltage while unloaded), a "non-urgent" failure alarm was not received; therefore, the failure of this power supply went undetected until the loss of the backup power supply.

Following a similar event which occurred in August of 1983 (see Unit 2 LEH 83-041/03L), testing of the rod control cabinet power supplies'was instituted on a refueling outage basis.

During the last refueling outage, all power supplies were tested with satisfactory results.

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e LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salen Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 05000272 86-001-00 4 0F 4

CORRECTIVE ACTION

(cont'd)

Because of this recent occurrence, PSE&G Engineering is now investigating possible methods of verifying the operability of the power supplies in the rod control cabinets during normal operation.

This item is being addressed by SORC Open Item number 86-003B-01.

Concerning the inadvertent actuation of the bus differential relay, testing of the relays associated with the remaining vital busses verified that they too were subject to similar actuations upon hard closure of the cabinet doors.

Although the seismic qualification of these relays is not in question, PSE&G Engineering is invertigating their service qualification; i.e.,

these relays contain " telephone contacts" which are vulnerable to actuation upon shock loading.

This item is being addressed by SORC Open Item number 86-0038-02.

In the interim, operators were made aware of this incident via the Operations Department Newsletter.

Additionally, signs were placed on the cabinet doors instructing personnel to close the doors easily.

FAILURE _DAT_Al Lambda Electronics D.

C.

Power Supply Type LM-262

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General Manager-Salem Operations JLR:tna SORC Mtg 86-008

O PSEG Public Service Electric and Gas Company P.O. Box E Hancocks Bridge, N+w Jersey 08038 Salem Generating Station February 14, 1986 U. S.

Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Dear Sirs SALEN GENERATING STATION LICENSE NO. DPR-70 DOCKET NO. 50-272 UNIT NO. 1 LICENSEE EVENT REPORT 86-001-00 This Licensee Event Report is being submitted pursuant to the requirements of 10CFR 50.73(a)(2)(iv).

This report is required within thirty (30) days of discovery.

Sincerely yours, J.

M.

Zupko', Jr.

General Manager-Salem Operations l

JLRaama C

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The Energy People

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