ML17157C367

Requests Comments on ASP Analyses Characterization of Possible Plant Response Given Event Occurrence.Comments on Whether Individual Analyses Reasonably Represent Plant Safety Equipment Configurations,Also Requested

ML17157C367
Person / Time
Site:	Susquehanna
Issue date:	06/10/1993
From:	Clark R Office of Nuclear Reactor Regulation
To:	Byram R PENNSYLVANIA POWER & LIGHT CO.
References
NUDOCS 9306180241
Download: ML17157C367 (36)

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Docket No. 50-388 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 10, 1993 Hr. Robert G.

Byram Senior Vice President-Nuclear Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101

SUBJECT:

RE(VEST FOR COMMENTS ON PRECURSOR ANALYSES, SUS(UEHANNA STEAN ELECTRIC STATION, UNIT 2, LER 388/92-001 The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has prepared a draft 1992 Accident Sequence Precursor (ASP) report, "Precursors to Potential Severe Core Damage Accidents:

1992, A Status

Report, NUREG/CR-4674, Volume 17 and 18."

As described in LER 388/92-001, on March 18, 1992, only two Emergency Diesel Generators (EDGs) were available for a period of 11. 1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

This event is being considered for inclusion in the ASP report.

One analyses of the event is enclosed.

If the appropriate personnel are available, your comments would be appreciated on the ASP analyses characterization of possible plant response given the event occurrence.

We are also interested in comments concerning whether the individual analyses reasonably repr esent plant safety equipment configurations and capabilities which existed at the time of the event.

Lastly, comments on the analyst's assumptions regarding equipment recovery probabilities would be appreciated.

Your comments and suggestions on the enclosed report would be appreciated.

However, in order to meet the required September 1993 publication date for the ASP NUREG report, we are asking that any comments or suggestions you may wish to offer be provided to us by July 1, 1993.

This request is covered by Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994.

The estimated average number of burden hours is 80 person hours per owner response, including the time required to assess the new recommendations, search data sources, gather and analyze the data, and prepare the required letters.

Comments on the accuracy of this estimate and suggestions to reduce the burden may be directed to the Desk Officer, Office of Information and Regulatory Affairs (3150-0011),

NEOB-3019, Office of Management and Budget, Washington, D.C.

20503, and to the 9306180241 930610 PDR ADOCK 05000388 P

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Mr. Robert G.

Byram June 10, 1993 U.S. Nuclear Regulatory Commission, Information and Records Management Branch (HNBB 7714), Division of Information Support Services, Office of Information and Resources Management, Washington, DC 20555.

If you have any questions or wish to discuss the analyses, please contact me at (301) 504-1402.

Sincerely,

Enclosure:

Analyses of LER 388/92-001 cc w/enclosure:

See next page Original signed by'ichard J. Clark Richard J. Clark, Senior Project Manager Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation DISTRIBUTION Docket File NRC

& Local PDRs PDI-2 Reading SVarga JCalvo OFFICE CHiller RClark HO'Brien OGC ACRS 10 PDI-2 PM EWenzinger, RGN-I

JWhite, RGN-I PDI-2/D NAME I

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93

/5/93 OFFICIAL RECORD COPY DOCUMENT NAME:

SULER.GEN

Mr. Robert G.

Byram June 10, 1993 U.S. Nuclear Regulatory Commission, Information and Records Management Branch (MNBB 7714), Division of Information Support Services, Office of Information and Resources Management, Washington, DC 20555.

. If you have any questions or wish to discuss the analyses, please contact me at (301) 504-1402.

Sincerely,

Enclosure:

Analyses of LER 388/92-001 cc w/enclosure:

See next page Rsc ard J. Clark, Senior Project Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Mr. Robert G.

Byram Pennsylvania Power 8 Light Company Susquehanna Steam Electric Station, Units 1 5 2 CC:

Jay Silberg, Esq.

Shaw, Pittman, Potts

& Trowbridge 2300 N Street N.W.

Washington, D.C.

20037 Bryan A. Snapp, Esq.

Assistant Corporate Counsel Pennsylvania Power 5 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. J.

M. Kenny Licensing Group Supervisor Pennsylvania Power 5 Light Company 2 North Ninth Street Allentown, Pennsylvania l8101 Mr. Scott Barber Senior Resident Inspector U. S. Nuclear Regulatory Commission P.O.

Box 35 Berwick, Pennsylvania 18603-0035 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Resources Commonwealth of Pennsylvania P. 0.

Box 2063 Harrisburg, Pennsylvania 17120 Mr. Jesse C. Tilton, III Allegheny Elec. Cooperative, Inc.

212 Locust Street P.O.

Box 1266 Harrisburg, Pennsylvania 17108-1266 Regional Administr ator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Mr. Harold G. Stanley Superintendent of Plant Susquehanna Steam Electric Station Pennsylvania Power and Light Company Box 467 Berwick, Pennsylvania 18603 1

Mr. Herbert D. Woodeshick Special Office of the President Pennsylvania Power and Light Company Rural Route 1,

Box 1797 Berwick, Pennsylvania 18603 George T. Jones Manager-Engineering Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101

PRELIMINARY B.22 LER Number 388/92-001 Event

Description:

Three of Five EDGs Unavailable for Eleven Hours Date of Event:

March 18, 1992 Plant:

Susquehanna 2

B.22.1 Summary During surveillance testing, emergency diesel generator (EDG) "B"tripped on "Generator Loss ofField" at 0831 hours0.00962 days <br />0.231 hours <br />0.00137 weeks <br />3.161955e-4 months <br /> on March 18, 1992. While in the proce'ss ofsubstituting in the "E" EDG (a fifthand spare EDG), the operator reset a relay target on Engineered Safeguard System (ESS) 4.16-kV bus 2C.

When the relay was reset at 0949 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.610945e-4 months <br />, the ESS bus 2C locked out. This caused several engineered safeguard feature (ESF) actuations, including the autostart of the "C" EDG, which remained unloaded.

The "C" EDG could not be loaded because the ESS bus 2C was locked out.

Normal ofisite power was supplied to ESS bus 2C at 2053 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.811665e-4 months <br />.

The "B EDG was determined to be unavailable for 19 d, the "C" EDG was unavailable for 11 h, and the "E" EDG was unavailable for possibly as long as 72 h. The conditional core damage probability of this event is estimated to be between 3.8 x 10 and 1.4 x 10. The relative significance of this event compared to other postulated events at Susquehanna 1 is shown in Fig. B.44.

Range$ ar IZR38852401 360 h HPCf + ROC Fig. B.44.

Relative event significance of LER 388/92401 compared with other Susquehanna 2 potential events.

B-371 LER NO: 388/92401 PRELIIlQNARY

PRELIMINARY 8.22.2 Event Description On March 18, 1992, with Unit 2 operating at 100% power, the "B" EDG was being run for its monthly surveillance test. At 0831 hours0.00962 days <br />0.231 hours <br />0.00137 weeks <br />3.161955e-4 months <br />, the "B" EDG tripped on "Generator Loss of Field." The loss of field trip occurred when a large load, the reactor building chiller, was started.

The resulting large increase in KVARS may have precipitated a failure of the generator field rectifier diode, which caused the uip.

The "B" EDG was unavailable for 19 d, 7 h and 49 min, which included time the "B" EDG was kept out ofservice for Unit 1 ESS bus refueling outage modification activities.

While in the process of substituting in the "E" EDG for the "B".EDG, the operator reset a relay target on ESS 4.16-kV bus 2C. When the relay target was reset at 0949 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.610945e-4 months <br />, the bus locked out because of the misoperation of the primary bus differential relay. The loss of this bus resulted in the auto-start of EDG "C" as well as several other ESF actuations.

Because bus 2C was locked out, the "C" EDG could not be loaded.

Power was restored to bus 2C at 2053 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.811665e-4 months <br />.

Therefore, EDG "C" was unavailable for 11 h and 4 min. The "E" EDG was substituted for the "B" EDG within the required limitingcondition for operation (LCO) action time of 72 h.

Loss of bus 2C resulted in these additional ESF actuations:

reactor water cleanup (RWCU) and containment instrument gas (CIG) containment isolations and Unit2 reactor building heating, venitilation, and air conditioning (HVAC) isolations.

Also, the drywell cooling fans were lost.

Because of the CIG isolation to the inboard main steam isolation valves (MSIVs), operators reduced recirculation flow and scrammed the reactor. The reactor water level reached Level 3, resulting in associated Level 3 isolations.

The maximum reactor pressure reached 994 psig, and the average drywell temperature reached 165 'F.

B.22.3 Additional Event-Related Information Susquehanna's emergency power system consists of four EDGs (A, B, C, and D) and one spare EDG (E) that are shared by two plants.

EDG E is capable of being substituted for any of the other EDGs

. without violating the independence of the redundant safety-related load groups.

ESS bus 2C supplies the following loads:

1 of 4 core spray pumps, 1 of 4 core spray pump room

coolers, 1 of4 residual heat removal (RHR) pumps, 1 of4 RHR room coolers, 7 of 14 drywell coolers, I of 2 instrument air compressors, 1 of 2 reactor building chillers, 1 of 2 reactor core isolation cooling (RCIC) room coolers, both standby liquid control heaters, 1 of2 standby liquid control injection pumps, 1 of 3 battery chargers, 1 of 4 containment hydrogen recombiners, and the main condenser vacuum pump.

B.22.4 Modeling Assumptions This event was modeled as a loss of two of four EDGs for 11.1 h.

The default nonrecovery factor of 0.8 was used for one calculation.

The calculation was then rerun using a nonrecovery factor of 0.34.

This nonrecovery factor corresponds to a failure that appears recoverable in the required period at the failed equipment when the equipment is accessible; in this instance, recovery from the control room does not appear possible.

Since the operators knew that the "C" EDG was running unloaded because of the B-372 LER NO: 388/92-001 PRELIMXNARY

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ISESI UCENSEE EVENT REPORT ILERI TEXT CONTINUATION A99IIOVEOONS NO. Sl SDCID9 EXSINES: AISLES EST¹9ATEO SUNOEN Sei NE~ TO CONSI,V WTH TH>>

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~ACE ISI OF DESCRIPTION OF EVENT On March 18, 1992, with Unit 2 operating in Condition 1 at 100X power and Unit 1 in refueling

, Condition 5, at OX power, the 'B'mergency Diesel Generator (EDG; EIIS Code:

EK) was being run for its monthly surveillance test.

At 0831 hours0.00962 days <br />0.231 hours <br />0.00137 weeks <br />3.161955e-4 months <br /> on 3/18/92, the 'B'DG tripped on "Generator Loss Of Field".

While in the process of substituting in the 'E'DG (which is a fifth and spare EDG) for the 'B'DG, the Operator reset a relay target on Engineered Safeguard System (ESS) 4.16 KV Bus 2C (EIIS Code:

EB).

When the relay target was reset (at 0949 on 3/18/92),

the ESS Bus 2C locked out.

The loss of the ESS Bus 2C resulted in several Engineered Safety Feature (ESF) actuations including, auto start of t e

'C'DG (remained unloaded),

Reactor Water Cleanup (RWCU; EIIS Code:

CE) and Containment Instrument Gas (CIG) system containment isolations and Unit 2 Reactor Building Heating, Ventilating and Air Conditioning (HVAC) Zones II and III (EIIS Code:

VA) isolations.

Additional ESS Bus 2C loads were lost, including Drywell Cooling Fans (EIIS Code:

VB) and additional isolations occurred.

Because the CIG system became isolated to the inboard Hain Steam Isolation Valves (MSIV; EIIS Code:

SB), Operators reduced reactor recirculation flow to minimum and manually scrammed Unit 2 in anticipation of MSD'losure.

All control rods fully inserted.

Following the scram, reactor water level reached Level 3 (+13") resulting in associated Level 3 isolations.

Minimum reactor level reached was -17.6 inches.

Maximum reactor pressure reached was 994 psig.

Average Drywell temperatuze reached 165 degrees F.

Unit 2 was taken to Cold Shutdown to allow Drywell entry for inspection.

CAUSE OF EVENT An Event Revie~ Team was formed to perform investigations and zoot cause analysis of this event.

Investigations into the cause of the 'B'DG trip identified a failed diode in the generator field rectifier bridge as a

potential cause.

Also investigated was the effect on EDG stability when a large load, Reactor Building Chiller (EIIS Code: VA), was started during the

'B'DG surveillance test run.

Preliminary computer modeling has indicated that the start of a large load, such as this chiller, can result in a large increase of KVAR output from the EDG when in the test (DROOP) mode.

When the chiller was started at 0831 on 3/18/92, the 'B'DG load increased from 4000 KW to 5075 KW and KVARS increased fzom +161 to -6025 KVARS and the 'B'DG tripped on "Generator Loss Of Field".

The large increase in KVARS measured correlates with the computer model data and may have precipitated failure of the generator field rectifier diode, resulting in the loss of field trip.

The lockout of ESS Bus 2C was unrelated to the tri of the '

G. It was during the evolution of substituting in the E

EDG for the 'B'DG that the Bus lockout occurred.

Specifically,'in accordance with operating proceciures, the Operator (utility; non-licensed) was checking all Unit 2 ESS 4.16 KV buses for indicating targets and resetting the targets as necessary.

The Operator

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LlCENSEE EVENT REPORT ILER)

TEXT CONTINUATION ATTIIOVKOON0 NO. 31 004104 0 XTINN(EITCICT 00TNIATCO 0(NtOtN tSI Nt&OICR TO CONPLT WTII T"+

INEOINMTION COLLtCTION AtQVCSTI 000 IIALSONW~O TTIENE0N10 IITOANOINO01NIOCN 00TNSATt TO TII00CCOIIO0 A103 NnaaVe EIANAOCNaNT0NANCN (PAIN.ua NMCL0AN 00(WLATONrCOINENCION. WACNINOTON.OC Xag0. ANO TO

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TEAN o so oo 38892 LtlIMPS 101

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0 tA00 IQ 040/06 "spark" leads to the conclusion thit the seal-in circuit conducted for a sufficient period to initiate, lockout relay operation at the same moment the Operator was depressing the target reset pushbutton.

The absence of any other anomaly after thorough investigation of all related buses, relays, relay circuitry and station activities at the time of the event leads to the conclusion that the Operator action and the lockout relay operation were not coincidental.

Therefore, the misoperation of the 87Al-B

Relay, on this occasion, was directly related to the operation of the target reset pushbutton.

REPORTABILITY/ANALYSIS The events resulting from the lockout of the ESS Bus 2C and the subsequent manual scram of the Unit 2 reactor were determined reportable per

'10CFR50.73(a) (2) (iv) as unplanned Engineered Safety Feature (ESF) actuations and an ESF actuation in response to a plant transient (manual scram).

The following unplanned ESF actuations occurred upon the lockout of the ESS Bus 2C:

~ 1

'C'DG auto start (remained unloaded)

RWCU containment isolation CIG containment isolation Unit 2 Reactor Bldg.

HVAC Zones II and III isolations In anticipation of a MSIV closure, Unit 2 was manually scrammed resulting in an ESF actuation of the Reactor Protection System (RPS; EIIS Code: JC).

Following the scram, reactor water level reached Level 3 (+13").

The reactor water Level 3 isolations constituted unplanned ESF actuations.

All control rods fully inserted during the manual scram.

Maximum reactor pressure reached was 994 psig.

Minimum reactor water level reached was -17e6 inches.

All system initiations and isolations occurred per desiFn in response to both the lockout of the ESS Bus 2C and the manual scram of the Unit 2 reactor.

declared inoperable following its surveillance run trip at 0831 on ~92 and the 'C'OG could not energ ze ESS Sus 2C due to the hus heing locked out This.constitueed a condition reportag?d pdr'OCFS"SO 73( )(2)a(v.)

and 10CFR50.73(a)(2)(vi) in that a condition existed which alone could have prevented fulfillment of the safety function of structures or systems needed to shutdown the reactor and maintain it in a safe shutdown, remove residual heat, control rad release or mitigate consequences of an accident.

Specifically, the Susquehanna Safe~Analysis requires three OPERABLE EDGs to safely shut down (ChanIrel 'C') and the lnoperable

'B'DG (Channel

.'BQ represented the potential for two channels being unavailable in the event of an accident.

The

'C'DG, successfully. ptartcd and continued running in an unloa~econditfonp's'er

design, given the locked-out bus condition.

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~TT HANS I'l Unit 2 n

t m E1ec tric Stati on TtXTN~~N~~~NACAHH~&IITI OOCSST Saaaaa ITI ytAA o

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0 fAOS ISI 06o>06 inspected for proper relay alignment and operation of the seal-in relay target.

The similar relays on Unit 1 will be inspected prior to startup from its 1992 refueling/inspection outage.

An engineering evaluation was performed to determine the effects, if any, of reaching an average Drywell temperature of 165 degrees F during the event.

The evaluation concluded that the increase in Drywell temperature had insignificant effect on equipment qualified life and no effect on Drywell or piping structural integrity.

A Drywell walkdown confirmed that there were no visual indications of heat induced damage present.'he diode was replaced on the 'B'DG generator field rectifier and the EDG was successfully retested and restored to operable status.

Engineering is continuing to study the dynamics of EDG response to voltage transients.

ADDITIONAL INFORMATION Failed Component Identification:

The 87A1-B Relay is not considered to be a

failed component but rather a relay misoperation.

Field rectifier diode Manufacturer:

PORTEC, Inc. P292 Diesel Manufacturer:

Cooper-Bessemer C634 Previous Reported Similar Events:

None identified.

t NOTE:

THIS LETTER IS BEING~ISTRIBUTED WITH ENCLOSURE XEROXED BACK-TO-BACK~

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Docket No. 50-388 UNITEO STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. O.C. 20555-0001 June 10, 1993 Mr. Robert G.

Byram Senior Vice President-Nuclear Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101

SUBJECT:

RE(VEST FOR COMMENTS ON PRECURSOR

ANALYSES, SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2, LER 388/92-001 The NRC's Office for Analysis and Evaluation of Operational Data (AEOD) has prepared.

a draft 1992 Accident Sequence Precursor (ASP) report, 4Precur sors to Potential Severe Core Damage Accidents:

1992, A Status

Report, NUREG/CR-4674, Volume 17 and 18."

As described in LER 388/92-001, on March 18, 1992, only two Emergency Diesel Generators (EDGs) were available for a period of ll.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

This event is being considered for inclusion in the ASP report.

One analyses of the event is enclosed.

If the appropriate personnel are available, your comments would be appreciated on the ASP analyses characterization of possible plant response given the event occurrence.

We are also interested in comments concerning whether the individual analyses reasonably represent plant safety equipment configurations and capabilities which existed at the time of the event.

Lastly, comments on the analyst's assumptions regarding equipment recovery probabilities would be appreciated.

Your comments and suggestions on the enclosed report would be appreciated.

However, in order to meet the required September 1993 publication date for the ASP NUREG report, we are asking that any comments or suggestions you may wish to offer be provided to us by July 1, 1993.

This request is covered by Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994.

The estimated average number of burden hours is 80 person hours per owner response, including the time required to assess the new recommendations, search data sources, gather and analyze the data, and prepare the required letters.

Comments on the accuracy of this estimate and suggestions to reduce the burden may be directed to the Desk Officer, Office of Information and Regulatory Affairs (3150-0011),

NEOB-3019, Office of Management and Budget, Washington, D.C.

20503, and to the

~800"~

$ 3og IFQ9 t l op/.

'C er E

Mr. Robert G.

Byram June 10, 1993 U.S. Nuclear Regulatory Commission, Information and Records Hanagement Branch (MNBB "7714), Division of Information Support Services, Office of Information and Resources Management, Washington, DC 20555.

If you have any questions or wish to discuss the analyses, please contact me at (301) 504-1402.

Sincerely,

Enclosure:

Analyses of LER 388/92-001 cc w/enclosure:

See next page Original signed by'ichard J. Clark Richard J. Clark, Senior Project Manager Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation DISTRIBUTION

,-Docket File NRC 5 Local PDRs PDI-2 Reading SVarga JCalvo OFFICE CHi 1 1 er RClark MO'Brien OGC ACRS 10 PDI-2 PM EWenztnger, RGN-I

JWhite, RGN-I PD I-2/D NAME DATE S%n k/ C/93 RC k:rb cS/

/93 CMiller Q/93 OFFICIAL RECORD COPY DOCUMENT NAME:

SULER.GEN

Hr. Robert G.

Byram June 10, 1993 U.S. Nuclear Regulatory Commission, Information and Records Management Branch (HNBB 7714), Division of Information Support Services, Office of Information and Resources Management, Washington, DC 20555.

If you have any questions or wish to discuss the analyses, please contact me at (301) 504-1402.

Sincerely,

Enclosure:

Analyses of LER 388/92-001 cc w/enclosure:

See next page Rsc ard J. Clark, Senior Project Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Mr. Robert G.

Byram Pennsylvania Power

& Light Company Susquehanna Steam Electric Station, Units 1

& 2 CC:

Jay Silberg, Esq.

Shaw, Pittman,.Potts

& Trowbridge 2300 N Street N.W.

Washington, D.C.

20037 Bryan A. Snapp, Esq.

Assistant Corporate Counsel Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. J.

M. Kenny Licensing Group Supervisor Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. Scott Barber Senior Resident Inspector U. S. Nuclear Regulatory Commission P.O.

Box 35 Berwick,'Pennsylvania 18603-0035 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Resources Commonwealth of Pennsylvania P. 0.

Box 2063 Harrisburg, Pennsylvania 17120 Mr. Jesse C; Tilton, III Allegheny Elec. Cooperative, Inc.

212 Locust Street P.O.

Box 1266 Harrisburg, Pennsylvania 17108-1266 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Mr. Harold G. Stanley Superintendent of Plant Susquehanna Steam Electric Station Pennsylvania Power and Light Company Box 467 Berwick, Pennsylvania 18603 Mr. Herbert D. Woodeshick Special Office of the President Pennsylvania Power and Light Company Rural Route 1,

Box 1797 Berwick, Pennsylvania 18603 George T. Jones Manager-Engineering Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101

PRELIMINARY B.22 LER Number 3SS/92-001 Event

Description:

Three of Five EDGs Unavailable for Eleven Hours Date of Event:

March 18, 1992 Plant:

Susquehanna 2

B.22.1 Summary During surveillance testing, emergency diesel generator (EDG) "B" tripped on "Generator Loss ofField" at 0831 hours0.00962 days <br />0.231 hours <br />0.00137 weeks <br />3.161955e-4 months <br /> on March 18, 1992. While in the process ofsubstituting in the "E" EDG (a fifthand spare EDG), the operator reset a relay target on Engineered Safeguard System (ESS) 4.16-kV bus 2C.

When the relay was reset at 0949 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.610945e-4 months <br />, the ESS bus 2C locked out. This caused several engineered safeguard feature (ESF) actuations, including the autostart of the "C" EDG, which remained unloaded.

The "C" EDG could not be loaded because the ESS bus 2C was locked out.

Normal ofisite power was supplied to ESS bus 2C at 2053 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.811665e-4 months <br />.

The "B" EDG was determuied to be unavailable for 19 d, the "C" EDG was unavailable for ll h, aad the "E" EDG was unavailable for possibly as long as 72 h. The coaditional core damage probability of this event is estimated to be between 3.8 x 10 aad 1.4 x 10'. The relative significance of this eveat compared to other postulated events at Susquehanna 1 is shown in Fig. B.44.

360 h HPCI + RCXC Fig. B.44.

Relative event significance of LER 388192401 compared with other Susquehanna 2 potential events.

B-371 LER NO: 388/92401 PRELDQNARY

PRELIMINARY B.22.2 Event Description On March 18, 1992, with Unit 2 operating at 100% power, the "B" EDG was being run for its monthly surveillance test. At 0831 hours0.00962 days <br />0.231 hours <br />0.00137 weeks <br />3.161955e-4 months <br />, the "B" EDG tripped on "Generator Loss of Field.

The loss of field trip occurred when a large load, the reactor building chiller, was started.

The resulting large increase in KVARS may have precipitated a failure of the generator field rectifier diode, which caused the trip.

The "B" EDG was unavailable for 19 d, 7 h and 49 min, which included time the "B" EDG was kept out of service for Unit 1 ESS bus refueling outage modification activities.

While in the process of substituting in the "E" EDG for the "B".EDG, the operator reset a relay target on ESS 4.16-kV bus 2C.

When the relay target was reset at 0949 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.610945e-4 months <br />, the bus locked out because of the misoperation of the primary bus differential relay. The loss of this bus resulted in the au~tart of EDG "C" as well as several other ESF actuations.

Because bus 2C was locked out, the "C" EDG could not be loaded.

Power was restored to bus 2C at 2053 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.811665e-4 months <br />.

Therefore, EDG "C" was unavailable for 11 h and 4 min. The "E" EDG was substituted for the "B" EDG within the required limitingcondition for operation (LCO) action time of 72 h.

F Loss of bus 2C resulted in these additional ESF actuations:

reactor water cleanup (RWCU) and "coatainmeat instrument gas (CIG) containment isolations and Unit2 reactor building heating, venitilation, and air conditioning (HVAC) isolations.

Also, the drywell cooling fans were lost.

Because of the CIG isolation to the inboard main steam isolation valves (MSIVs), operators reduced recirculation flow and scrammed the reactor'. The reactor water level reached Level 3, resulting in associated Level 3 isolations.

The maximum reactor pressure reached 994 psig, and the average drywell temperature reached 165 'F.

B.22.3 Additional Event-Related Information Susquehanna's emergency power system consists of four EDGs (A, B, C, and D) and one spare EDG (E) that are shared by two plants.

EDG E is capable of being substituted for any of the other EDGs without violating the independence of the redundant safety-related load groups.

ESS bus 2C supplies the following loads:

1 of 4 core spray pumps, 1 of 4 core spray pump room

coolers, 1 of4 residual heat removal (RHR) pumps, 1 of 4 RHR room coolers, 7 of 14 drywell coolers, 1 of 2 iristrumeat air compressors, 1 of 2 reactor building chillers, 1 of 2 reactor core isolation cooling (RCIC) room coolers, both standby liquid control heaters, 1 of2 standby liquid control injection pumps, 1 of 3 battery chargers, 1 of 4 containment hydrogen recombiners, and the main condenser vacuum pump B.22.4 Modehng Assumptioas This event was modeled as a loss of two of four EDGs for 11.1 h.

The default nonrecovery factor of 0.8 was used for one calculatioa.

The calculation was then rerun using a aonrecovery factor of 0,34.

'Ihis aonrecovery factor corresponds to a failure that appears recoverable in the required period at the failed equipment when the equipment is accessible; in this instance, recovery from the control room does aot appear possible.

Since the operators knew that the "C" EDG was running unloaded because of the LER NO: 388/92401 PRXXB4INARY

PRELIMNARY lockout of the 2C bus, and since there was no failure of the "E" EDG (spare), it is reasonable to assume that the faHure was recoverable within the required time.

Finally, the calculation was run a third time using a nonrecovery factor of 0.04, which corresponds to a failure that appears recoverable in the required period from the control room and is considered routine or procedurally based.

Since it is reasonable to assume that substituting the spare "E" EDG for the "B" EDG is a routine activity which occurs in the control room, this third nonrecovery factor appeared to be the best choice for calculating the core damage frequency.

8.22.5 Analysis Results

'Ihe calculated core damage frequency using the default nonrecovery factor of0.8 is 3.8 x '. Changing the nonrecovery factor to 0.34 resulted in a core damage frequency of 1.6 x '. Using the nonrecovery factor of0.04, which appeared to be the best estimate for this event, resulted in a core damage frequency of 1.4 x 'nd is shown in Fig. B.45.

LER NO: 388/9?A01 B-373 PRELIMZNARY

PRELIMINARY 40

~ I IS 44

~4

~I 4S OI SI SS SS SS

~I a

SS

~4 SC SI

~I Sl IO IÃ CO CX CK CO C

IÃ CO IX CO OC CO IW CO CO OI OC CO OI CK CO OI OC CO OI CO OI CO IÃ CO EX OI CO CO OI CO IX CO CO OI CO OC CO CO CO 4IOS IR IÃ CD OI OC CO OC CO CO AIDS Hg. BAS.

Dominant core damage sequence for LER 388/92401.

B-374 LER NO: 388/92401 emu.mmARY

PRELIMINARY CONOlTlONAL CORE OANAGE PROBABlLlTT CALCULATlONS Event tdentf f 1 ere 388/92-001 (caee 1)

Event Oeecrlptlons Three of five EOOs uavalleble for 11 h Event Datec 03/18/92 P lant z Suequelwaa 2

MNAVAlLABll.lTY,NNATltNe 11.1 houre NON RECOVERABLE lNlTlATlNO EVENT, PROBABlLITlES CO" TRANS LOP LOCA Total ATMS TRANS LOOP LOCA

'otal",

Sequence'3'loop',

8Kk0"POKR'rx;ehutckan/ep, ep.'rec' 9? 'oop-K%RQ'POMER"'rx"shutdae'"

98

~

loop ENC:POKR'"., rx;ehutdoia....

95':; loop -K%RO'.PNSM~ rx."ehutdoun.,

N"" loop"K%RG.$NNKR'-'rx';"ahutdoun/ep, ep.reo

,97 '- loop" E%kC;POKE.-': rxiehutdceel> "-.

E "Event ldentfffart."388/92-.'001:::(ca'i>>" I)'

1.7E 03 9.6E 05 1.8E.OS O.OE+00 3.8E-06 O.OE+00 3.8E-06 O.OE+00 2)2E-16 O.OE+00 2.2E 16 End State CO.

AIMS.

AIMS.

AIMS CD" AIMS

'3.7E-06.

2QK.09

( 2.3E 09 )

(

23E-09.7E-06.

2'3E 09

'O'ec 4.2E-01 4.2E 01 1.1E 01 1 1E 01" 4.2E~01 4,2E 01 LER NO: 388/92401 B-375 PRELIIMINARY

PRELIMINARY SEOOENCE iKOELc cchasp'yadetslhircseet.~

BRANCH NCOELc c:Rasp'Vaodetslt fmrfck.st 1 PROBABlLLTT FiLEc c:KaspgiodeLAbrr call.pro BRANCN FRECRKIICIES/PROBABLL1TLES Branch Systc<<

Non. Recov Opr FaiL trans 1.6E.04 Loop 1.6E.05 loca 3.3E-06 rx.shutdown 3.0E-05 rx.slwtdcsal/ep 3.5E.04 pct/tr<<>>

1.7E.01 trv.chal(/trans.-term 1.0E+00 srv.chaL L/Loop.-term 1.0E+00 SrVoCLC>>e 4.6E.02 8%RG. PCNIER 1.4E-02 > 1.0E+00 Branch Nodetc 3.OFA Trafn 1

Cond Prob:

5.0E.02 > Failed Trafn 2 Cond Prob!

5.7E-02 i Failed Train 3 Cond'rob:

1 o9E-01

'Train 4 Cced Prob:

5.0E-01 IPo reo 4.9E.02 fic/pcs. trans 4.6Eo01 f<pcs.loca '" '-

1.0E+00

. hpcf.-:-,.

r2.9E-02 I'cic, -.;

'=="." -'.

'6.'OE-02 crd-'",;,,:, '...',, '1.0E-OZ' srv;ade", ",,.':;-,,

=...'3.7E-03",,.;Lpci(rlii)/Lpcs';:rc"'.";-.,-;

'1;OE-'03.',

rhr(sdc);-':::.~'"". -. ',.

..;Z.TEo02.

rhr(sdc)/.'Lpcf-':- >.,.:

"'- '.'-"- ~.:2;OE-02.-

,rhr(edc)/tpcf:', "., ".::-'-.q.:.

- "1.0E+00;";

-.

rhr(spcoot)/rhr(sdc)".o".-..-,;;,,'.-;,";.;Z.OE-03,,

"."-rfir(spcoot)/-'pcf;rhr(sdc) i '-2:OE-03

" "rhr'(spco'ot)/L'pct.'rhr(sdc) "'.,'."

.',, '9;3E-OZ ",

'hrsci.-'"~;,',.'~::-

":, '+;.-'.,; -,"

2.0E-OZ'

+ bran'ch-siodet ffLe)

. -'.'~,,

, Eovant,ifodsntf ffelitr":306/92'001 "(Case 1),'....

1.0E+00 5.3E.01 5 OE-01 1.0E+00 1.0E+00 1.0E4X) 1.0E+00 1.0E+00 1.0E+00 B.OE.01 1.0E+00 3.4E 01 3AE-01 T.OE-01 loOE 01 1.0E+00 T 1E-01;,",

"3'.'4E.01'o1E.OL',

r'AE01'*

3 4Eo01:

1:OE+M lo3'4E 01'.-

3'.4E-'01"

"~ 1"OE~

3AE 01 o

o 1.0E-02

,,., '1,.0E-OZ,

'loOE 03 1cOE.03Eo03 2 OE-03 o

,r B-376 LER NO: 388/92401 PRELGGNARY

PRELIM%NARY CONDITIONAL CORE DAHAGE PROBABILITY CALCULATIONS Event ldentiffere 3M/92-001 (case 2)

Event Descriptionc Three of five EDGs csIevat lable for 11 h Event'Detec 03/18/92 PLentc Susquchecv>>

2 LNAVAILASILITY, DLNATIONN 1'1.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> NCN RECOVERABLE INITIATINGEVENT PROBABILITIES TRANS L(NN>>

LDCA 1.TE 03 9.6E-05 1.8E.05 SEQUENCE CCNNITIONAL PROBABILITY SQCS End State/Initiator Probabf lity a,

/

CO..

TRANS O.DE+00 LOP 1.6E-06

,, LOCA O.DE+00

.." TotaL',

1.6E.06 ASS:

TRANS O.DE+00 LON'-';DE+00

".,LOCA, 0.0&00 Total ~ / 4 O.OE+00-.

, SEDUENCF~CQNITlONAL, PROBASIL'ITlES-'(PROBASIL'ITY CNDERI',. '

'gj,';;;"-">>,Sequence:.';.-,,><<

'.=,:

EndState prob

.N.-Recce

'83<<.-',.';! Loop" E%RC.'POKR". rxcahutdcieI/ip,'<ep.rec,>> <<j,',";,,-,';;

CO.

1.5K. 0$"..,8E 01

',97.,",,"Loop-- K%Nt'PORN" <<m.ehutcfaiet.'=,,'i>>",,;.,'; '~ -",-'",-:-", ~.,'"i'~'q ",,;.',, ',. ATIN,,",;

.,.>>9.'5E'T0'p'j. 1AE-.01

. 96(-'!Loop>>BKRQI OK%.~rx shutdoen:'.->>:":"

ATQS..9~-.;1O,"4'6K-Ol "cxa-recovery cred ft forcedf Ncf.,case'I,",;;;;,.;~

"Pp'.>> 'c' SEANCE>> C(NNIITIOIAL':PROBASIL'ITLES'(SEOKltCE M)ER) ~'

>>.;"':,:-::,;:~<<,>>4:;, ',",;,"~".,I!~e.,";.

'.'EndcState-

'Prab"-~4',

I<<%>>ReCIIe V

95"'."-Loop'.'BKk0;POKR'"'rx'shutdoant;""

-=

ATMS

{ 9 5K '10-'J

" 3.5E 01

--'83-.Loop'-BKRO'PQKk-rx.shutdown/ep

.ep.rec

= -

= 'CD 1.5&06~:~ *-TAf.01.

,.9F)$';",.loop'~c<<~RB,~",;ra'shutd~...

-,," =.

~-...

'ATIN'-

9>> 5E;.10,~, '

8E,01 eecncsc-r<<evilly',"credit!for-edfted,case Note: ):",for.':ce>>vetlibflftte>>a'cindftfonaL probebflfty;values'ere dffferactfal vale>>e abfcb reflect the

~dded'rfsfc<<due<to fefLures".essocfeted Nfth-en event.:" Perenthetfcal-"values: fndfcate areduction fn rfsk"coepered;to;;e afar far:perfodlut thout the:exfstfnB fafluree.-,,

V SEQUENCE 1NNEL'iA',,'$c:Wasp~La>>lburcaeeL. cap.-,

"Identfffar'i; 388/92-'001'-",(c'ase':.Yj."

B-377 LER NO: 388/92401 PRELIh @NARY

'RELHNINARY SRANCR NNELT c:~eep~tahti~rick.at1 PRLRNILL'LTY flLE:

c:LaspWdeta'lbw call.pro Ro RecoverY Lialt SRARCH FREQXWCIES/PROBASlLlTLES System 1.6E.04

'1.6E-05 3.3E-06 3.0E.05 3.5E.04 1.7E.01 1.0E+00 1.0E+00 4.6E-02 1 4E-02 i 1.0E+00 1.0E+00 5.3E 01 5.0E 01 1.0E+00 1.Of+00 1.'OENO 1.0E+00 1.0E+00 1.0E+00 d.OE-OT i 3.4E-01 fai led felled 1.'OE+00 3AE-01 3AE-01 T+OE Ol 7.0E-01 1.0E+00 7'1E.01

.'3AE-01

'"'-.'-'7'E '0l".

<< '.3;4E-'01" 3AE.01'

...,.1.0E+N 3

4E-'Ol'",,

'p~"

',; ~,",)3"4E)01"

,1.0E+00,,

, '3'i4E.OT

)

)

trans Loop Loca rx.ahutdosn rx.shutdown/ep".

pca/trans erv.chat t/tram.-ace arv.chat t/Loop.-acrm arv.close EARG. PQKR drench Nodet:

3.0f.4 Train 1

Cond Prob:

5.0E-02 i Train 2 Cond'Prob:

5.7E-02 i "Train 3'ond Prob:

'.9E-01

'Train 4 Cond Probe 5.0E Ol ep.rec" 4.9E 02 fe/pca.'trans 4.6E.01 1a/pci.toca,'-'," '." '

=,.

1.0E+00

'hpct';:.~', ".=,,

...2.9E-02roice"~j.'

6.OE-OR

. -crd -.-";..

~"

='*; ',"""

1.0E-O2

.=arv.ade';-.

3.7K-03,.-

>Lpci{rhr)/L'p)ca- ~,,

=.,

-. 1:OE-03.

~.rhr(adc);.';";,;;.:. "

- -:.2.-1E-02

rhr{edc)/-'tp)cf

,"-.~; -~"-';-

R.OE-02

. 'ver(adc)/tpct;.'.,j.':,;,~.,'...",',',,

",:.,T.OE+00

'. rhr(apcoot)'/rhr(adc)'."""

". )(:;:;;, '- 2;OE-'03,

" rhr{epcoot}f-;Lpct';rh'r(adc)'<,';::.;.',:,;.;,:,<'2;OE-'L6.

'hr{epcoot)/Lpcl.rhr{adc)"',-"".;"

'-: ""9)3E'-'ORI"

<<"'branch'aidet fito.,'"'~",:-'-';:,,~::"'('".,;, '..

Event Tdentfff'er',"38d/92-'001 <Cease 2):;:, '"

~;.'~).

, '(-j j.:,? (@.>x4eA'i;.,'<.', <.i') a.g g',;cy.'.).;)-~-+))~w,))')';(

apr fail "1.%'-02 1'Of 02

'1'E-'03' 1'%3'1 OE.:03.

'R %-03 B-378 LKRNO: 38S/92401 PRELIMINARY

PRELIMINARY CONDITIONAL CORE DANACE PROBABILITT CALCULATIONS Event Ident(ffer:

388/92-001 (case 3)

Event Deacrfptfon: Three of ffve EDGa unavailable for 11 h Event Date:

03/18/92 Plant:

Suequeharaa 2

WAVAILABILITY, DCNIATION+ 11.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Mesc RECOVERABLE IMITIATlltOEVENT PROBABILITIES TRAMS LAP'OCA v

SE(RKNCE C(aOITIINIAL PROBABILITY SQIS 1.7E-03 9.6E.05 1.8E.05 Probabf Ifty O.DE+00 1.4E-07 O.DE+00 1.4E.07 O.DE+00 5 AE-17 O,DE+00 5v6E '17.

State Prob-,

M'Roc~'D'AE"07v.""'".2v 1

&02'TMS'4E-11 2c !E-OR" ASS

( a;4E. 11' 5'E.O'I

-State.

Prob M kec ATMS:

(.a.4E.-T 1'.)

5.-1E-01

~'CO

- i '.4E-"OF";,

2'TE.O2.

vATQS;,.a.'4E-.T12 TE OZ erentf el'aluaa.-ccMch reflect.the Luce fndfcatc e reductlocl In Jv vv End State/Inftiator CD TRAMS LOP.

LOCA':

, Total;", ",,

ATMS TRAMS;.",

-* L XP". ".

vv v

SE(RKMCEIC(NOITI(NIAL'.PROBABIL'ITIES>(PRMlSILITY (NB)ER)

83

-',,:;:::.(O'Op',.vv EIKRe;PIER.<<rX";ahutdOiCn/ap.'eP;reC;

~

97>", (oop",'E%'RO.P(SKR;""'rx ahutdoccn;"v v

96~! loop-.'-'BKRG;P(NKR, rx ahutdoe:;

~covery:credft 'or edftacf caae;:.-',

. SE(RKMCE C(BEIITI(NNL',PR(NIAallITlES'(SEINKMCE ORDER>:

.~ '

96'-'-;(oop>-.EKING'.P(NKR",,::,,rx;ahutdoun'.": s.... ~

, v""83j~.::(oops'E%ROPSKR,,'xlahutdoccn/op i;.'ep. rec';.':;:>'.l"" -'

ee:non-recovery~'credf t'for'dftid ~"'.".

='ote

',".,"'-For'evif(ibflftfaa","condftfonavl.'probabflftyvaluea" are dfff

~ddad'rfak.chio:,to,faflurea'aaaocfated~ufth an avant; 'Parenthetfcal va

.;; rfalc coaparad'to:a'afÃl'ar,'"perfod'Nfthout-,the: exfetfngfailures.,

', SE(RANCE';'-IBM)ELt'9'i;::-;;:;"ciiasvpvv~laMrcse~I;cap..

Event Tdentfffet". '388/92.'001 (case v3)'; -,.:,

B-379 LER NO: 388/92401 PRELIhGNARY

PRELIMINARY Opr; Faf t h

SRANCN NNEL's, c:Kaap~ta'itfmrfck.a(1 PRtNASlLITT'FlLEc

', ctkaap~takbvr catl.pro

>>V*

No Raeovaty Lfaft

"'~'RANCN FREOKIClES/PROBASILlTlES Sraneh Syatee Non-Racov

~

true",'," '"

='.";",

1.6E-'04", "',.

1.0E+00

~

5'W-01 toca'.';."'..'>>.".-'.,,

'-, '3.3E-06.

5.0E.01 rx.ahutdae '

~

'.0E-05

1.0E+00

. ~,,rx'afwtdael/ap:o

...-.3.5E.04 1!OEKO

"~pca/trana'.t-",,

1.7E-01 1.D&40' arv.chat t'/trana.-aerie

-", ';OE+00' 1.0E+00

. arv;chat t/Loop;-acrm"

"~ 1."OE+00

'"1.>>OE+00

,arv.ct'oaa>>~; ',"';,

~ ','.6E.02 1 OE+00

<<EIKRC,'POKR".-'"'-",~ ~

',1 4E-02 > 1,0E+00

~,,

d.OE-01

> 4.0E-02

,,-;.,;'Nranch 'godet'3.0F.4; 5.0E 02 > Fat ted

"'-;".,Trafn'2""Cond'Prob>>,,

5.7E-02 > Faf tad

.Trairi.3'ond Prob:j: '-

";, ".1;:9E-'01; '. ",, '..';

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UCENSEE EVENT REPORT tLERI AAKIOVSOOM0 NO. 51104IIM 5NtHIt5 M50t05 SSTIMATSO SUNOSN tt1 Nt~ TO CCNMLT Wftl TWS INKNIMATIONCOLLSCTION 15OUSST 504 NNS. >ONWAAO COMMSNTS 15OANOINO 01NIOSN SSTIMATS fO THt 1lCONOS ANO 15KHIT5 MANAOlMSNT0NANCH IP0501. U.S NVCLtA1 AlOVLATOATCOMMOSION WASNINOTON. OC l0050 ANO TO THt PAttNWOAN 1lOVCTION tNOISCT 15150410il OttICS Ot MANACIMSNTANO SUOOST.WASrrINOTON, OC 50$Q FACILITYNAMtIll Sus uehanna Steam Electric Statioh - Unit 2 OEKSST~1 Ifl o

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ttttCTSO SU0MISSION OATt IISI MONT<<OAT TSAA AASTNACT ILrrrrrt M rt00 MttM r ~ MAr0trrrterrr trtNHt IISI On March 18, 1992, with Unit 2 operating in Condition 1 at 100Z power aad Unit 1 in refueling, Condition 5, at OZ power, the 'B'mergency Diesel Generator (EDG) vas being run for its monthly surveillance test.

ht 0831 hours0.00962 days <br />0.231 hours <br />0.00137 weeks <br />3.161955e-4 months <br /> on 3/18/92, the 'B'DG tripped on "Generator Loss Of Field".

While in the process of substituting in the 'E'DG for the

'B'DG, the Operator reset a relay target on Engineered Safeguazd System (ESS) 4.16 IG~

Bus 2C.

When the relay target vas reset at 0949 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.610945e-4 months <br /> on 3/18/92, the bus locked out.

The loss of the ESS Bus 2C resulted fn several ESF actuatfoas including, auto start of the

'C'DG, Reactor Water Cleanup and Contaimaent Iastrumeat Gas (CIG) containment isolstions and Unit 2 Reactor Buildfng HVhC Zones II and III isolations.

hddftional bus loads, including Dryvell cooling fans vere lost and additfonal isolations occurred.

Because the CIG system became isolated from the Main Steam Isolation Valves (MSIV) ~ operators manually scrammed Unit 2 in anticipation of MSIV closure.

Folloving the scram, reactor vatez level reached Level 3 (+13") zesulting in Level 3 isolatioas.

Unit 2 vas taken to Cold Shutdown.

The root cause of this event was attrfbuted to misoperation of a primary bus differential relay, vhich occurred vhen the target reset pushbutton vas depressed by the Operator.

Folloving electrical investigation/evaluation of the bus and fts protective cizcui.try, power was restored to the bus at 2053 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.811665e-4 months <br /> on 3/18/92.

The sub5ect relay was tagged to identify that in thc event a relay target is observed, Operations should contac:

Systems Engineering prior to resetting.

The relay vill be replaced at a later date and more thoroughly examined/tested to aid in understanding the misopcration resulting from resetting thc target.

All similar relays on Unit 2 were inspected and those on Unit I vill be inspected prior to startup from its 199" refueling outage.

Repairs vere completed on the 'B', EDG cad Q wag gggorcd gp ppeggbpc gpppps<f Qppe mph no safety consequences NAC Irma 550 I045r

IAWI IACILITTIIAANI'Il UCENSEE EVENT REPORT tLERl TEXT CONTlNUATION OOCACT~A QI AAAOOVCOOaee ae. IIII4IAA CIAIOCI:AIIIIIT CCTOAATCO %NO%I AT% Ilt&CWÃtTO CCWAI,Y~ TIIIS NIAONNTIOIICOLLICTIOOA~. IAA~. SOaWAAO COM~ OIOANTONKJNKNCITNATTTO TNI AICOIIOA AN) ARCS!HA~IT~ %4%I. IIA.NICLAAII IIIOWATCOYCCAOAMCWO.W~IOOTOII.OC~. ANITO TIIA AAAI~AIOIICTCNIMOICCT IJIIOAIAII.OAAICT OA INIIAOCINIITAIIOOIIOOCT.NAAIINOTOII.OC ISCL INIAIIT>Ah y.'.I unit 2 Susauehanna Steam Electric Station 11Xt&~~O~~AN:

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OF DESCRIPTION OF EVENT On March 18, 1992, vith Unit 2 operatiag in Conditioa 1 at 100X pover and Unit 1 ia refueling

, Condition 5, at OX poverI the 'B'mergency Diesel Generator (EDG; EIIS Code:

EK) vas being rua for its monthly surveillance test.

At 0831 hours0.00962 days <br />0.231 hours <br />0.00137 weeks <br />3.161955e-4 months <br /> on 3/18/92, the 'B'DG tripped oa "Generator Loss Of Field".

While in the process of substituting in the 'E'DG (vhich is a fifth and spare EDG) for thc 'B'DG, the Operator reset a relay target on Engineered Safeguard System (ESS) 4.16 KV Bus 2C (EIIS Code:

EB).

When thc relay target vas reset (at 0949 on 3/18/92),

the ESS Bus 2C locked out.

The loss o

the ESS Bus 2C resulted in several Engineered Safety Feature (ESF) actuations including, auto start of t e

'C'DG (remained unloaded),

Reactor Mater Cleanup (RWCU; EIIS Code:

CE) and Containment Instrument Gas (CIG) system containment isolations and Unit 2 Reactor Building Heating, Ventilating and Air Conditioning (HVAC) Zones II and III (EIIS Code:

VA) isolations.

Additional ESS Bus 2C loads vere lost, including Drywell Cooling Fans (EIIS Code:

VB) and additional isolations occurred.

Because thc CIG system became isolated to thc inboard Main Steam Isolation Valves (MSIV; EIIS Code:

SB), Operators reduced reactor recirculation flov to minimum and manually scrammed Unit 2 in anticipation of MSIV closure.

All control rods fully inserted.

Following the scram, reactor vater level reached Level 3 (+13") resulting in associated Level 3 isolations.

Minimum reactor level reached was -17.6 inches.

Maximum reactor pressure reached was 994 psig.

Average Dryvell temperature reached 165 degrees F.

Unit 2 vas taken to Cold Shutdown to allov Drywcll entry for inspection.

CAUSE QF EVENT An Event Reviev Team vas formed to perform investigations and root cause analysis of this event.

Investigations iato the cause of the 'B'DG trip identified a failed diode in the generator field rectifier bridge as a

potential causa.

Also investigated was the effect on EDG stability vhen a large load, Reactor Buildiag Chiller (EIIS Code: Vh), vas started during the

'B'DG surveillance test rua.

Preliminary computer modeling has indicated that the start of a large load, such as this chiller, caa result ia a large increase of KVAR output from the EDG when ia the test (DROOP) mode.

When the chiller vas started at 0831 on 3/18/92, the 'B'DG load iacreased from 4000 KM to 5075 KW and KVARS increased from +161 to -6025 KVARS and the 'B'DG tripped on "Generator Loss Of Held".

The large increase in KVARS measured correlates vith thc computer model data and may have precipitated failure of the generator field rectifier diode, resulting in the loss of field trip.

The lockout of ESS Bus 2C vas unrelated to the tri of t e '

G. It vas during thc evolution of substituting in t c EDG for the 'B'DG that the Bus lockout occurred.

Specifically,'in accordaace vith operating procedures, the Operator (utilitv; non-licensed) vas checking all Unit 2 ESS 4.16 KV buses for inc'cating targets and resetting the targets as necessary.

The Operator

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N401 PACILITY~ Ill LlCENSEE EVENT REPORT LER)

TEXT CONTlNUATlON OOCNCT~A OI v

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found the relay target actuated on Primary Bus Differential Relay 87hl-B on ESS Bus 2C.

Mhen the 0 erator resc~emalay

target, he noticed a "spark" in the area of the relay seal-in unit (internal to the relay) and bus lockout relays

actuated, tripping and locking out all ESS Sua 2C circuit kreakara.

The investigation of the ESS Bus 2C lockout relay operation vas divided into Caro areas:

Physical and electrical checks of the bus and its appurtenances to determine i.f physical damage occurred vhich could have caused the 87Al-B relay to operate in a normal fashion foz bus protection.

Physical and electrical checks of the 87AI-B Primary Bus Differential Relay and its associated circuits to determine if relay misopcration vas the cause of the ESS Bus 2C lockout.

The ESS Bus 2C vas found intact and not degraded.

This was determined.by me@ger testing of the bus and associated potential transformer circuitry.

No faults vere detected.

Additionally, a faulted bus condition is likely to trip at least tvo Primary Bus Diffcrcntial Relays, vhich did not occur.

The 87Al-B Primary Bus Differential Relay and its associated circuits vere found to function properly.

No physical or electrical defects or anomalies vere observed.

The relay vas checked for functional calibration and alignment/distorti.on both in place and removed and manually manipulated several times to verify that no mechanical binding or erratic motion vas present.

a It is PP&L's engineering

)udgement that the zoot cause of the ESS Bus 2C lockout was a misoperation of the 87Al-B Primary Bus Differential Relay vhich occurred vhen the target reset pushbutton vas depressed by the Operator.

Several factors support this conclusion:

The operation of the 87hl-B Pzimary Bus Differential Relay is designed to cause thc lockout af ESS Bus 2C in the exact manner observed on 3/18/92.

We observations of the Operator, from the moment he depressed the target reset pushbutton.

are consistent vith the intended design function of this electrical protection scheme for the bue alignment which existed prior to the eveTIt.

Thc mechanism foz the pastulated mi.soperation of the 87hl-B Relay could not be replicated during subsequent investigation.

However, the seal-in contact of the 87Al-B'Relay is part of the seal-in target assemblv.

The action of depressing the target reset applies a force in the direction of seal-in contact closure.

The seal-in contact is the primary circuit path to trip the lockout relays.

As such, the Operator's observation of a

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~~v v UCENSEE EVENT REPORT LER)

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"spark" leads to the conclusion that the seal-in circuit conducted for a sufficient period to initiate lockout relay operation at the same moment the Operator was depressing the target reset pushbutton.

The absence of any other anomaly after thorough investigation of all related buses, relays, relay circuitry and station activities at the time of the event leads to the conclusion that the Operator action and the lockout relay operation were not coincidental.

Therefore, the misoperation of the 87A1-B

Relay, on this occasion, was directly related to the operation of the target reset pushbutton.

REPORTABILITY/ANALYSIS The events resulting from the lockout of the ESS Bus 2C and the subsequent manual scram of the Unit 2 reactor were determined reportable per 10CFR50.73(a)(2)(iv) as unplanned Engineered Safety Feature (ESF) actuations and an ESF actuation in response to a plant transient (manual scram).

The following unplanned ESF actuations occurred upon the lockout of the ESS Bus ZC:

'O'DG auto start (remained unloaded)

RWCU containment isolation CIG containment isolation Unit 2 Reactor Bldgs HVAC Zones II and III isolations In anticipation of a MSIV closure, Unit 2 was manually scrammed resulting in an ESF actuation of the Reactor Protection System (RPS; EIIS Code: JC).

Following the scram, reactor water level reached Level 3 (+13").

The reactor water Level 3 isolations constituted unplanned ESF actuations.

All control rods fully inserted during the manual scram.

Maximum reactor pressure reached was 994 psig.

Minimum reactor water level reached was -17.6 inches.

All system initiations and isolations occurred per design in response to both the lockout of the ESS Bus 2C and the manual scram of the Unit 2 reactor.

Th declared inseparable follouint its surveillance run trip at 0831 vali locked out This.constituted a condition reporta5?a pdr 3008830'13(a) (2.) (v) and 10CFR50.73(a)(2)(vi) in that a condition existed which alone could have prevented fulfillment of the safety function of structures or systems needed to shutdown the reactor and maintain it in a safe shutdown, remove residual heat, control rad release or mitigate consequences of an accident.

Specifically, the Susquehanna

~S ggg~alvsis reouires three OPERABLE EDGs to safely shut down (Chappel 'C') and the inoperable 'B'DG (Channel

.'BQ. represented the potential for two channels being unavailable in the event of an accident.

The C

EDG successfu3.ly. ytartcd and continued running in an"unloafee conhition,'s per design, given the locked-out bus condition.

10001 LICENSEE EVENT REPORT ILER)

TEXT CONTINUATION t

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The plant was safely shut dovn and there vere no radioactive releases recorded by effluent instrumentation.

The emergency operating procedures vere properly implemented by Operations personnel.

s An engineering evaluation concluded that the maximum D

teeesracuc~~

degrees F reached duringcha,eva,nChsd,insignifiesnc efface on equipment qualified life and no effect on Dryvell or pipiag structural integrity.

A Drywell walkdown confirmed that there vere no visual indications of heat induced damage present.

There were no safety consequences or compromise to public health or safety during this event.

Investigations into the cause of the 'B'DG trip at 0831 on 3/18/92 identified a failed diode in thc gcncrator ffeld rectifier bridge. It fs believed that the start of a large load (Reactor Bldg. Chiller) vhile the 'B'DG vas in the Test (DROOP) mode resulted in a large increase in the KVAR output from the EDG and mav have precipit'ated failure of the diode, resulting in the trip on "Generator required Technical Specification 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCK.Actfgn,~O~~~.ZO~gq.

unavailable for 19 days 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and 49 minutes.

However, the total time out of service included tfmc fn which the 'B'DG was kept out o serv c

re t of Unit 1

Bus ref outage modification activities.

The 3/18/92

'B'DG trip fs considered a valid test and valid failure.

The~BI EDG Start Log indicates there is one (1) 'B'DG failure in the last 20 valid tests.

The

'B'DG test interval is one start at least once per 31 days per Technical Specification Table 4.8.1.1.2-1.

This Licensee Event Report also satisffes reportability pursuant to Technical Specification section 4.8.1.1.4.

In accordance with the guidelines provided fa NUREG 1022 Supp.

1 Item 14.1 and 10CFR50.4(d),

the required submission date for this report vas determined to be April 20, 1992.

CORRECTIVE ACTIONS Folloving investigations and evaluations of the ESS Bus 2C aad fts protective circuitry, pover vas restored to the ESS Bus 2C via its normal offsite supply at 2053 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.811665e-4 months <br /> oa 3/18/92.

The subject 87AI-B Relay vas tagged to identify that ia the event a relay target is observed, Operations should contact Systems Engineering prior to resetting the target.

The relay will be replaced at a later data and thoroughly examined/tested to aid in understanding the misoperation resulting from resetting the target.

The incident vas revieved with all Operations personnel including a discussion of proper relay target reset practices.

All similar 'Primary Bus Difierent'al Relays on Unit 2 vere

~ 1

~tt IIAIfIIII ILLIIUCLAAAAECULATOAT LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION COACT INSH4% Ql AffWOIIIOOW IIO.tIIOOICA IIfIAOf4IIWOT ttfOaATCO CUWOIW fIAA~ TO ~T WTII TIIII IWfOWWATIOWCOLLICTIOW >%Cater. MO IIWI. fOWWAAO

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tnr If~~e~ ~~WSC fWW~tfIm inspected for proper relav alignment and operation of the seal-in relay target.

The similar relays on Unit 1 will be inspected prior to startup from its 1992 refueling/inspection outage.

An engineering evaluation was performed to 'determine the effects, if any, of reaching an average Drywall temperature of 165 degrees F during the event.

The evaluation concluded that the increase in Drywall temperature had insignificant effect on equipment qualified life and no effect on Drywell or piping structural integrity.

A Drywell walkdown confirmed that there were no visual indications of heat induced damage present.

The diode was replaced on the 'B'DG generator field rectifier and the EDG was successfullv retested and restored to operable status.

Engineering is continuing to study the dynamics of EDG response to voltage transients.

ADDITIONAL INFORMATION Failed Component I'dentification:

The 87Al-B Relay is not considered to be a

failed component but rather a relay misoperation.

Field rectifier diode Manufacturer:

PORTEC, Inc. P292 Diesel Manufacturer:

Cooper-Bessemer C634 Previous Reported Similar Events:

None identified.

4.

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