ML18058B083

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LER 92-034-01:on 920701,reactor Automatically Tripped on Loss of Load Signal,Resulting in Turbine Generator Trip & Loss of Signal.Caused by Momentary Loss of Power to Turbine Sys computers.Post-trip Review completed.W/920915 Ltr
ML18058B083
Person / Time
Site: Palisades Entergy icon.png
Issue date: 09/15/1992
From: Roberts W, Slade G
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-034, LER-92-34, NUDOCS 9209220405
Download: ML18058B083 (6)
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Text

..

consumers Power GB Slade General Manager *

~-., POWERINli .

. MICBlliAN"S PROliRESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Coven, Ml 49043

  • September 15, 1992 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

. DOCKET 50-255 - LICENSE DPR PALISADES PLANT . , LICENSEE EVENT REPORT

  • 92-034 REACTOR TRIP CAUSED BY LOSS OF LOAD SIGNAL RESULTING FROM A.

FAILURE OF THE TURBINE CONTROL SYSTEM - SUPPLEMENTAL REPORT Li~ensee Event Report ("LER) 92-034-01.is attached.* This supplement changes the most probable cause 6f the failure of the turbine control system from

. loose cable connections to the data processing units, to a momentary loss of power to the.control sjste~s computers .. This event wa~*originally reported to the NRC in accordance with 10 CFR 50.73(a)(2)(iv) as an event that resulted in the automatic actuation of an erigineered safety feature.

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Gerald B Slade General Manager CC Administrator, Region III~ USNRC NRC Resident Inspector Palisades Attachment

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9209220405 9209i5 PDR ADOCK 05000255 S . PDR .

... A CMS NCRGY COMPANY

NRC Form 388. U.S. NUCLEAR REGULATORY COMMISSION 19*83) APPROVED OMB NO. 3160-01~

EXPIRES: 8/31 /86

  • LICENSEE EVENT REPORT (LERI FACILITY NAME 111 DOCKET NUMBER 121 PAGE.131 Palisades Plant 0 I 5 I 0 I 0 I 0 j '2 I 5 I 5 1 I OF 0 1** 5 TITLE 141 REACTOR TRIP CAUSED BY LOSS OF LOAD SIGNAL RESULTING FROM A FAILURE OF THE TURBINE

-,i\ITi< SVS'T'F.M EVENT DATE 161 LER NUMBER 181 REPORT DATE 181 OTHER FACILITIES INVOLVED 181 SEQUENTIAL *REVISION* FACILITY NAMES YEAR ....

MONTH DAY YEAR NUMBER .  :* .> NUr.ilBER MONTH DAY YEAR N/A 0 I .I I I I .I 6 *O 0 0 011 011 9 2 912 o I 314 al i oI 9 ~ s9I2 N/A 0 l l l l I I 6 0 0 0 THIS.REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR I: fCh<<k OM or,,,0,. of 1M following} 1111

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  • 60.73(1)(2Jlxl b91ow and !n Text, NRC Form 388Al LICENSEE CONTACT FOR THIS LER 1121 NAME TELEPHONE NUMBER William L. Roberts, Staff Licensing Engineer COMPLETE ONE Ll°NE FOR EACH COMPONENT FAILURE.DESCRIBED IN THIS REPORT 113)

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to 1 81 hours9.375e-4 days <br />0.0225 hours <br />1.339286e-4 weeks <br />3.08205e-5 months <br />' the pl ant was stable' operating at 100% power with all system~ in a normal full power alignment, when the reactor automatically tripped on a loss of load signal. A turbine gener~tor trip was generated by the turbine generator digital electrohydraulic (DEH) control system, resulting in a loss of load signal to the reactdr protection system (RPS). The plant was taken to a hot shutdown condition after the event. All safety systems responded as designed. The apparent cause of the turbine generator trip was a momentary loss of power to the turbine DEH system computer*s which ca_use.d a turbine trip. The turbine trip resulted in a reactor trip due to lo~s of load. Corrective acti-0ns taken in response to the July 24, 1992. plant trip (LER 92-035) will assure that this event will not

  • recur.

NRC Form 388A U.S. NUCLEAR REGULATORY- COMMISSION li-83) APPR0\/£0 OMB NO. 3160.()10' EXPIRES: 8/311815 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION F.ACILITY NAME 111 DOCKET NUMBER 12! LER NUMBER 13!

SEQUENTIAL REVISION YEAR. NUMBER NUMBER Palisades Plant 0 5 0 . 0 *0 2 5 5 9 2 - 0 3 5 - 0 1 Q 2 OF 0 5 EVENT DESCRIPTION -

  • On July 1, 1992, at approximately 1232 hours0.0143 days <br />0.342 hours <br />0.00204 weeks <br />4.68776e-4 months <br />, th~ plant was stable, operating at 100% power with all systems in~ normal full power alignment, when the

. reactor automatically tripped o_n a loss of lead signal. A turbine generator trip was generated by the turbine generator digital electrohydraulic (DEH) control system [JJ;CPU], resulting in a loss of load sigrial to the reactor protection system (RPS). The auxiliary f eedwater system actuated on low levels in the steam generators as required. At 1233 hours0.0143 days <br />0.343 hours <br />0.00204 weeks <br />4.691565e-4 months <br />, operations entered Emergency Operating. Procedure (EOP) 1, "Post Trip Actions." At 1241 hours0.0144 days <br />0.345 hours <br />0.00205 weeks <br />4.722005e-4 months <br />,

_EOP 1 was completed and EOP 2; "Trip Recovery" was entered which included emergency boration~ The emergency boration was i~itiated due to the failure of Bus lA to transfer to start-up power and the resulting loss of 2 of the*4 operating primary coolant pumps. At 1243 hours0.0144 days <br />0.345 hours <br />0.00206 weeks <br />4.729615e-4 months <br />, it was noted that the nitrogen pressure in one of the safety injection tanks, T-82B, was at 178 lbs.

At 1320 hours0.0153 days <br />0.367 hours <br />0.00218 weeks <br />5.0226e-4 months <br />, the emergency boration was secured and EOP 2 was exited.

  • General Operating Procedure (GOP) 3, "Hot Shutdo~n to Critical and Hot Stan.dby ,_" was entered and the pl ant was stab le.

A post trip review was completed shortly after the event with the followin~

observations.

The DEH displa~ on the control room operators console did not .automatically display a turbine valve status screen' on the reactor trip as des_igned, but the . _,

alirm page display showed system actuations._

  • Th~ t~ntr~l roo~ annunciator chime appeared tb ma)function (not sound) for several minutes after the trip. It subsequently returned to servite on its own.

The critical* functions monitor and control room indication responded normally following the trip. Nucl~ar instrumentJtion power range ihannel NI-007, which provides data logger indication, slowly dropped.to zero following the re~ctor trip.

The 4160 VAC non-safety related Bus IA supplying two of the-four primary coolant pumps failed to transfer to start-up power from station powe~,

resulting in*. loss of two primary c.oolant pumps.

The safety injection tank (T-828) relief valve (RV-3128) lifted and reseated at the time of the trip which caused the low pressure indication.

The condensate storage tank (.T-2) ov*erfi 11 ed during the event and leaked out a soft patch on an existing crack in the top of the tank. '

This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv) as an event that resulted in ~utomatic actuation of the RPS system.

NRC Form 388A. U.S. NUCLEAR REGULATORY 'coMMISSION 19-831 APPROVED OMB NO. 3160-0104 EXPIRES: 8/31 /86

. LICENSEE EVENT REPORT (LERI TEXT CONTINUATION

~. FACILITY NAME 11 I DOCKET NUMBER 121 LER NUMBER 131 PAGE 141 SEQUENTIAL REVISION YEAR NUMBER NUMBER Palisades Plant 0 5 0 0 0 2. 5 5 9 2 - 0 3 4 - 0 1 0 3 OF Q 5 CAUSE OF.THE EVENT The unit tripped_as a result of a loss communication within.the turbine Digital Electrohydraulic (DEH) control system which resulted in a loss of load signal being sent to the reactbr protection syste~ (RPS). The apparent cause of the. DEH system failure was initially thought to be loose cable. connections to*the DEH data processing units. However, on July 24, 1992, a failure of the DEH system again resulted in a plant trip. Reviews of the July 24, 1992 plant trip determined that a momentary loss of power to the ,DEH system computers caused *th~ plant t~ip. It's believed that the voltage transient ~bserv~d on the 345 KV Argenta-Palisades switchyard tie line, which otcurred immediately prior to the J~ly 1, 1992 event, may have had the same *ffect on the DEH c~mputers as was determined for th¢ July 24~ 1992 plant trip and is the ~ost

  • probable cause of the plant trip.
  • ANALYSIS OF THE EVENT Th~ DEH system was inve~tigated by system engin~ering and ve~dor.personnel.

The initial most likely cause of the system failµre appeared to be *1 interconnecting caples not properly connected to the DEH system circuit

\boards .. The latch mechanisms on these cables were not properly secured, thus the connections were loose. Loss of contact via these connectors.may have occurred causing redund_ant Data Processing Units (DPU)

  • 2 and 52 to simultaneously drop out and cause a turbine trip. Loss of ~ommuriication from DPU 2 and 52 for four seconds will, per design, cause a turbine trip signal.

Immediately prior to the trip, (approximately 1 sec), the 345 KV Argenta-Palisades switchyard tie Tine tripped causing breaker cycling in the Palisades swit~hyard. A~ a res~lt of our reviews and apparent ~imilarities to the July 24, 1992 plant trip, we have determined that this switchyard tie line trip and breaker cycling resOlted in.* voltage transient that caused the DEH

  • system to trip the turbine which resulted in the reactor trip.
  • The control room annunciator chime alarm sounds coincident* with a light illuminating at the top of whichever control room panel has alarmed. The chime alarm and the panel light help* the operator to more easily identify from which control room panel the alarm has come in on. In addition to the initial troubleshooting that was completed, long term actions are planried to investigate the failure of the chime alarm.
  • The ~uclear instrumen~at~on channel Nl-007 plant data logger input failed to respond correctly.
  • Control room and critical functions ~onitor indications were normal. It was during the post trip review that it was discovered that this channel ~f nuclear instrumentation had exhibited a data logger indication problem~ The problem was corrected. The failure of this data logger indication had no impact on this event.

NRC Form 388A _U.S. NUCLEAR REGULATORY COMMISSION 1~831 APPROVED OMS NO. 316<Hl1~

EXPIRES: 8/31185 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION

.. -*. FACILITY NAME 111 DOCKET NUMBER 121 LER NUMBER 131 PAGE 141 SEQUENTIAL REVISION YEAR NUMBER

  • NUMBER Palisades Plant. 0 5 0 0 0. 2 5 5 9 2 - 0 3 .4 - 0 1 Q 4 OF 0 5 The 4160 VAC Bus IA supplies two of the four primary coolant pumps. During check-out of the breakers that failed to transfer, it was discovered that the cubicle interlock was not properly aligned, which caused the failure to

'*transfer. This .alignment is set at the factory as part of the breaker refurbishment, but actual *field conditions may exist that require this factory setting to be che~ked. In th~ future, the cubicle interlock will be adjusted as required to align with the speclfic cubicle tbe breaker is installed in.

An*. inspection of the other breakers refurbished by the same vendor and. **

installed during the last refueling outag~ was made. Four bre~kers were checked. Two of the breakers were re-adjusted to make the alignment better.*

One required no adjustment and the other one had operated satisfactory, was not safety related, and was not checked.

  • The safety injection tanks have a technical specification lower limit for

~ressure at 200 psig when the reactor is.critical. When relief valve RV-3128 for safety injection tank T-82B actuated it ~llowed the pressure in the.tank to drop to 178 lbs which is below the technical specificatiori limit, but did not reach the limit until after the reactor was in hot shutdown. Actuation of this tank could not be totally relied upon during a required post accident scenario, but three of the four safety injection tanks are all that is needed to fulfill the safety furiction.

As a result of.the plant trip, excess wate~ from the main condenser was

  • automatically rejected to the condensate storage tank. Because of the quantity

~f water .rejected and the fact that an area on the tdp of the tank was lightly patched, the tank overflowed from the tank overflow and broke through the

. ~atched area also.* The soft patch area was repafred and long term follow ~p actions will evalu~te if other scenario's exist which could duplicat~ this event.

  • The turbine valve status screrin did not display on the control room operators console as this screen display is generated from DPU 2 and 52, which were out of service. Corrective actions to keep the DPU.units in service will also assure that this display is generated.

CORRECTIVE ACTION A post trip revi~w wa~ completed in accordance with Pall sades. Administrative procedure 4.08,* "Post Trip Review Requirements."

Prior to start-up the following actions were taken.

Vendor and maintenance personnel secured and checked all connectors on the DEH system.and performed testin~ of the sjstem.

SIT relief RV-3128 was repaired, tested and returned to service, and SI Tank T82B was re ressurized.

N~C Form 388A U.S. NUCLEAR REGULATORY COMMISSION

'* {9-83) APPROVED OMB NO. 31~104 EXPIRES: 8/31186 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION FACILITY NAME Ill . DOCKET NUM_BER 121 . LOI NUMBER 131 PAGE 141 SEQUENTIAL REVISION YEAR NUMBER NUMBER Palisades Plant 0 5 0 0 0 2 5 5 9 2 - Q 3 4 - Q 1 Q 5 OF Q 5 The NI-007 indication was checked and its upper data-logger indication signal con9itioning card was found to be defective and replaced.

Bus IA start~up breaker was checked and found that its cubicle interlock needed adjustment. The interlock was adjusted and the breaker returned to service.

The* control room annunciator chime was tested,' several components of the chime replaced, ahd successfully tested.*

The condensate storage tank (T-2) soft patch wa*s repaired ..

  • Longer term correct~ve actions have b~en assigned as follows to continue to review some of the conditions discovered *dtiring this plant trip.

Further reviews will be completed to investigate the root cause of the DEH data processing failures. that appear to be re 1ated to the 1oose cable .

connections on th~ DPU units.

Other events where the possibility exists where.the main condenser may reject ~ater to the condensate storage tank will be investigated and recommendationsmade as appropriate to mitigate this scenario.

  • Further reviews wjll be-~ompleted to determine a rbot cause for the failure of the control room annunciator chime during ~his reactor trip and an apparent similar occurrence from December-9, 1991. -

Long term revi~ws and the July 24, 1992 plant trip have shown that switchyard events can adversely affect the DEH turbine control system.

Cofrective actions reported in LER 92-035, will assure that this condition will not recur.

  • ADDITIONAL INFORMATION On July 24, 1992 the plant tripped on a loss of load signal. Subsequently*

LER 92-035 described that a momentary loss of power resulted in a DEH systerri generated turbine and subsequent plant trip. A momentary loss of power to the DEH computers has also been determined to be the cause of the July I, 1992 plant trip.

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