Severe Accident Analysis in Nuclear Power Plants - downloads

149 Pages·2012·40.49 MB·English

Science and Technology of Nuclear Installations
Severe Accident Analysis in
Nuclear Power Plants
Guest Editors: Gilberto Espinosa-Paredes, Lluís Batet,
Alejandro Nuñez-Carrera, and Jun Sugimoto
Severe Accident Analysis in
Nuclear Power Plants
Science and Technology of Nuclear Installations
Severe Accident Analysis in
Nuclear Power Plants
Guest Editors: Gilberto Espinosa-Paredes, Llu´ıs Batet,
Alejandro Nun˜ez-Carrera, and Jun Sugimoto
Copyright©2012HindawiPublishingCorporation.Allrightsreserved.
Thisisaspecialissuepublishedin“ScienceandTechnologyofNuclearInstallations.”Allarticlesareopenaccessarticlesdistributed
undertheCreativeCommonsAttributionLicense,whichpermitsunrestricteduse,distribution,andreproductioninanymedium,pro-
videdtheoriginalworkisproperlycited.
Editorial Board
NusretAksan,Switzerland MichelGiot,Belgium ManmohanPandey,India
A.C.MarquesAlvim,Brazil ValerioGiusti,Italy YuriyParfenov,Russia
WonPilBaek,RepublicofKorea HorstGlaeser,Germany YvesPontillon,France
StephenM.Bajorek,USA SatishKumarGupta,India NikPopov,Canada
GeorgeBakos,Greece AliHainoun,Syria PieroRavetto,Italy
JozsefBanati,Sweden KeithE.Holbert,USA FrancescReventos,Spain
RicardoBarros,Brazil KostadinIvanov,USA EnricoSartori,France
AnisBousbiaSalah,Belgium YacineKadi,RepublicofKorea CarloSborchia,France
GiovanniB.Bruna,France AhmedKhedr,Egypt MassimoSepielli,Italy
NikolaCˇavlina,Croatia TomaszKozlowski,USA ArkadySerikov,Germany
XuCheng,China TomoakiKunugi,Japan JamesF.Stubbins,USA
LeonCizelj,Slovenia MikeKuznetsov,Germany IztokTiselj,Slovenia
AlejandroClausse,Argentina H.-YeonLee,RepublicofKorea RizwanUddin,USA
FrancescoD’Auria,Italy BunditLimmeechokchai,Thailand EugenijusUsˇpuras,Lithuania
MarcosP.deAbreu,Brazil JiriMacek,CzechRepublic RichardWright,Norway
GiovanniDell’Orco,France AnnalisaManera,USA ChaoXu,China
JuanCarlosFerreri,Argentina BorutMavko,Slovenia YankoYanev,Bulgaria
NikolayFil,Russia OlegMelikhov,Russia ZhiweiZhou,China
CesareFrepoli,USA RafaelMiro´,Spain EnricoZio,Italy
GiorgioGalassi,Italy JosefMisak,CzechRepublic MassimoZucchetti,Italy
ReginaGaletti,Brazil RahimNabbi,Germany
Contents
SevereAccidentAnalysisinNuclearPowerPlants,GilbertoEspinosa-Paredes,Llu´ısBatet,
AlejandroNun˜ez-Carrera,andJunSugimoto
Volume2012,ArticleID430471,2pages
PreliminaryAssessmentofthePossibleBWRCore/VesselDamageStatesforFukushimaDaiichiStation
BlackoutScenariosUsingRELAP/SCDAPSIM,C.M.Allison,J.K.Hohorst,B.S.Allison,D.Konjarek,
T.Bajs,R.Pericas,F.Reventos,andR.Lopez
Volume2012,ArticleID646327,25pages
SimulationoftheLowerHeadBoilingWaterReactorVesselinaSevereAccident,
AlejandroNun˜ez-Carrera,Rau´lCamargo-Camargo,GilbertoEspinosa-Paredes,andAdria´nLo´pez-Garc´ıa
Volume2012,ArticleID305405,8pages
FailureAssessmentMethodologiesforPressure-RetainingComponentsunderSevereAccidentLoading,
J.Arndt,H.Grebner,andJ.Sievers
Volume2012,ArticleID487371,10pages
ResponseAnalysisonElectricalPulsesunderSevereNuclearAccidentTemperatureConditionsUsingan
AbnormalSignalSimulationAnalysisModule,Kil-MoKoo,Jin-HoSong,Sang-BaikKim,Kwang-IlAhn,
Won-PilBaek,Kil-NamOh,andGyu-TaeKim
Volume2012,ArticleID656590,15pages
SevereAccidentSimulationoftheLagunaVerdeNuclearPowerPlant,GilbertoEspinosa-Paredes,
Rau´lCamargo-Camargo,andAlejandroNun˜ez-Carrera
Volume2012,ArticleID209420,11pages
StationBlack-OutAnalysiswithMELCOR1.8.6CodeforAtucha2NuclearPowerPlant,AnaliaBonelli,
OscarMazzantini,MartinSonnenkalb,MarceloCaputo,JuanMatiasGarc´ıa,PabloZanocco,
andMarceloGimenez
Volume2012,ArticleID620298,17pages
AnEvaluationMethodologyDevelopmentandApplicationProcessforSevereAccidentSafetyIssue
Resolution,RobertP.Martin
Volume2012,ArticleID735719,13pages
HeatandMassTransferduringHydrogenGenerationinanArrayofFuelBarsofaBWRUsingaPeriodic
UnitCell,H.Romero-Paredes,F.J.Valde´s-Parada,andG.Espinosa-Paredes
Volume2012,ArticleID878174,10pages
Large-ScaleContainmentCoolerPerformanceExperimentsunderAccidentConditions,RalfKapulla,
GuillaumeMignot,andDomenicoPaladino
Volume2012,ArticleID943197,20pages
TheEuropeanResearchonSevereAccidentsinGeneration-IIand-IIINuclearPowerPlants,
Jean-PierreVanDorsselaere,AriAuvinen,DavidBeraha,PatrickChatelard,ChristopheJourneau,
IvoKljenak,AlexeiMiassoedov,SandroPaci,Th.WalterTromm,andRolandZeyen
Volume2012,ArticleID686945,12pages
HindawiPublishingCorporation
ScienceandTechnologyofNuclearInstallations
Volume2012,ArticleID430471,2pages
doi:10.1155/2012/430471
Editorial
Severe Accident Analysis in Nuclear Power Plants
GilbertoEspinosa-Paredes,1Lluı´sBatet,2
AlejandroNun˜ez-Carrera,3andJunSugimoto4
1DepartamentodeIngenier´ıadeProcesoseHidra´ulica,UniversidadAuto´nomaMetropolitana-Iztapalapa,
AvenidaSanRafaelAtlixco186Col.Vicentina,09340M´exico,DF,Mexico
2DepartmentofPhysicsandNuclearEngineering,UniversitatPolit`ecnicadeCatalunya(BarcelonaTECH),Av.Diagonal647,
08028Barcelona,Spain
3Comisio´nNacionaldeSeguridadNuclearySalvaguardias,DoctorBarraga´n779,Col.Narvarte,M´exicoCity,DF,Mexico
4DepartmentofNuclearEngineering,GraduateSchoolofEngineering,KyotoUniversity,Yoshida,Sakyo,Kyoto606-8501,Japan
CorrespondenceshouldbeaddressedtoGilbertoEspinosa-Paredes,[email protected]
Received9September2012;Accepted9September2012
Copyright©2012Gilberto Espinosa-Paredes et al. This is an open access article distributed under the Creative Commons
AttributionLicense,whichpermitsunrestricteduse,distribution,andreproductioninanymedium,providedtheoriginalworkis
properlycited.
Safety of nuclear power plants is essential and safety K.-M. Koo et al. presented a response analysis on elec-
standards are continuously reviewed and upgraded as new trical pulses under severe accident temperature conditions
developments and research are performed. Continuous usinganabnormalsignalsimulationanalysis.Theseauthors
researchregardingthissubjectisfundamentalforthenuclear obtained a special function for abnormal pulse signal pat-
industry. Although severe accident analysis and research ternsthroughacharacteristicresponseundersevereaccident
have been performed throughout the evolution of nuclear temperature conditions, which in turn makes it possible
industry, it has not yet considered all plausible scenarios. to analyze the abnormal output pulse signals through a
Adequate analyses are needed for all phases of severe characteristic response of a 4∼20mA circuit model and a
accidents in order to maintain or improve safety margins. range of the elements changing with temperature under an
Hence, it is essential to encourage researchers to keep accidentcondition.
performinganddevelopingresearch,codes,andsimulations G.Espinosa-Paredes etal.presentedthesimulation and
of these potentially hazardous events. The original works analysisoftheloss-of-coolantaccident(LOCA)intheboiling
publishedinthisspecialissuecanhelptoimprovesafetyand water reactor (BWR) of Laguna Verde Nuclear Power Plant
understandthephenomenainvolvedinsevereaccidentsand (LVNPP)at105%ofratedpower.Thesimulationconsiders
their consequences in existing generation II nuclear power a LOCA in the recirculation loop of the reactor with and
plants(NPP)aswellasingenerationIIINPPbeingbuiltand without cooling water injection. The LVNPP model was
ingenerationIII+andIVNPPbeingdeveloped. developed using the RELAP/SCDAPSIM code. The lack of
Inthisspecialissuetenresearcharticleswerepublished.J. coolingwateraftertheLOCAleadstheLVNPPtocoremelt-
Arndtetal.describeamethodtoperformsimplifiedanalyses ingthatexceedsthedesignbasisofthenuclearpowerplant
concerning integrity of the components of the primary (NPP) sufficiently to cause failure of structures, materials,
cooling circuit during a severe accident. A second method, andsystemsthatareneededtoensurepropercoolingofthe
using complex calculation models, was used to analyze a reactorcorebynormalmeans.Facedwithasevereaccident,
postulated high-pressure core melt accident scenario in a the first response is to maintain the reactor core cooling
PWR caused by a station blackout. Authors found that by any means available, but in order to carry out such an
temperature values of more than 800◦C can be reached attempt it is necessary to fully understand the progression
in the reactor coolant line and the surge line before the of core damage, since such action has effects that may be
bottomofthereactorpressurevesselexperienceasignificant decisive in accident progression. During the progression
temperatureincreaseduetocoremelting. of core damage, these authors analyzed the cooling water
2 ScienceandTechnologyofNuclearInstallations
injection at different times and the results show that there (M-configuration) and the top (T-configuration) of the
are significant differences in the level of core damage and containment vessel. The experiments are characterized
hydrogen production, among other variables analyzed such by a 3-phase injection scenario. In Phase I, pure steam
asmaximumsurfacetemperature,fissionproductsreleased, is injected, while in Phase II, a helium-steam mixture is
anddebrisbedheight. injected. Finally, in Phase III, pure steam is injected again.
AdescriptionoftheresultsforaStationBlackoutanalysis FortheM-configuration,astrongdegradationofthecooler
forAtucha2NuclearPowerPlantispresentedbyA.Bonelli performance was observed for these authors during the
et al. Calculations were performed with MELCOR 1.8.6 injectionofthehelium/steammixture(PhaseII).FortheT-
YV3165Code.Atucha2isapressurizedheavywaterreactor, configuration,weobserveamainlydownwardsactingcooler
cooled and moderated with heavy water, by two separate resultinginacombinationofforcedandnaturalconvection
systems,presentlyunderfinalconstructioninArgentina.The flow patterns. The cooler performance degradation was
initiatingeventislossofpower,accompaniedbythefailure much weaker compared with the M-configuration and a
of four out of four diesel generators. All remaining plant goodmixingwasensuredbytheoperationofthecooler.
safetysystemsaresupposedtobeavailable.Itisassumedthat Forty-threeorganizationsfrom22countriesnetworking
during the station blackout sequence the first pressurizer their capacities of research in SARNET (Severe Accident
safety valve fails stuck to open after 3 cycles of water Research NETwork of excellence) to resolve the most
release.Duringthetransient,thewaterinthefuelchannels important remaining uncertainties and safety issues on
evaporatesfirstwhilethemoderatortankisstillpartiallyfull. severeaccidentsinexistingandfuturewater-coolednuclear
Themoderatortankinventoryactsasatemporaryheatsink power plants (NPP) are discussed by J. P. Van Dorsselaere
for the decay heat, which is evacuated through conduction etal.Accordingtotheseauthors,thefirstprojectinthe6th
andradiationheattransfer,delayingcoredegradation.These FrameworkProgramme(FP6)oftheEuropeanCommission,
authors found that this feature, together with the large theSARNET2project,coordinatedbyIRSN,startedinApril
volume of the steel filler pieces in the lower plenum and a 2009for4yearsintheFP7frame.After2.5years,somemain
highprimarysystemvolumetothermalpowerratio,derives outcomes of joint research (modeling and experiments) by
inaveryslowtransientinwhichRPVfailuretimeisfourto thenetworkmembersonthehighestpriorityissuesarepre-
fivetimeslargerthanthoseofGermanPWRs. sented: in-vessel degraded core coolability, molten-corium-
R.P.Martinpresentedageneralevaluationmethodology concreteinteraction,containmentphenomena(waterspray,
development and application process (EMDAP) paradigm hydrogen combustion), and source-term issues (mainly
for the resolution of severe accident safety issues. For the iodinebehavior).TheASTECintegralcomputercode,jointly
broader objective of complete and comprehensive design developedbyIRSNandGRStopredicttheNPPSAbehavior,
validation, severe accident safety issues are resolved by capitalizes in terms of models the knowledge produced in
demonstrating comprehensive severe accident-related engi- thenetwork:afewvalidationresultsarepresented.
neeringthroughapplicabletestingprograms,processstudies C. M. Allison et al. focused on the Fukushima Daiichi
demonstrating certain deterministic elements, probabilistic accident; they present an assessment which includes a brief
riskassessment,andsevereaccidentmanagementguidelines. review of relevant severe accident experiments and a series
The basic framework described in this paper extends the ofdetailedcalculationsusingtheRELAP/SCDAPSIMmodel
top-down, bottom-up strategy described in the US Nuclear whichwereprovidedbytheComisionNacionaldeSeguridad
Regulatory Commission Regulatory Guide 1.203 to severe Nuclear y Salvaguardias, the Mexican nuclear regulatory
accident evaluations addressing US NRC expectation for authority. The authors concluded that detailed analysis for
plantdesigncertificationapplications. realistic bounding scenarios can provide general guidance
Thenumericalanalysisofheatandmasstransferduring oftimingofimportanteventsandthattheresponsestothe
hydrogen generation in an array of fuel cylinder rods, each accidentoncetheaccidentisunderwaycanmakeasignificant
coatedwithacladdingandasteamcurrentflowingoutside differenceintheconsequencesoftheaccident.
the cylinders, is presented by H. Romero-Paredes et al. Finally,A.Nu´n˜ez-Carreraetal.presentedtheanalysisof
The analysis considers the fuel element without mitigation theBoilingWaterReactor(BWR)lowerheadduringasevere
effects. The system consists of a representative periodic accidentusingSCDAPSIM/RELAP53.2.Thecomputercode
unit cell where the initial and boundary value problems wasusedinthisworktomodeltheheatupofthereactorcore
for heat and mass transfer were solved. In this unit cell, it materialthatslumpsinthelowerheadofthereactorpressure
is considered that a fuel element is coated by a cladding vesselduetoaloss-of-coolantaccident(LOCA)withsimul-
with steam surrounding it as a coolant. The numerical taneouslossofoff-sitepowerandwithoutinjectionofcool-
simulations allow describing the evolution of temperature ing water. The authors conclude that SCDAPSIM/RELAP5
and concentration profiles inside the nuclear reactor and hasthecapabilitytopredictthemeltingofthecore,control
couldbeusedasabasisforhybridupscalingsimulations. rod, and some structures, with an estimation of the main
R. Kapulla et al. presented validation experiments, parameterofthemoltenpooluntilthefailureofthecrust.
conductedintheframeoftheOECD/SETH-2Project.These
experiments address the combined effects of mass sources GilbertoEspinosa-Paredes
Llu´ısBatet
andheatsinksrelatedtogasmixingandhydrogentransport
AlejandroNun˜ez-Carrera
within containment compartments. A wall jet interacts
JunSugimoto
withanoperatingcontainmentcoolerlocatedinthemiddle
HindawiPublishingCorporation
ScienceandTechnologyofNuclearInstallations
Volume2012,ArticleID646327,25pages
doi:10.1155/2012/646327
Research Article
Preliminary Assessment of the Possible
BWR Core/Vessel Damage States for Fukushima Daiichi Station
Blackout Scenarios Using RELAP/SCDAPSIM
C.M.Allison,1J.K.Hohorst,1B.S.Allison,1D.Konjarek,2T.Bajs,2R.Pericas,3
F.Reventos,3andR.Lopez4
1InnovativeSystemsSoftware,1242SouthWoodruffAvenue,IdahoFalls,ID83404,USA
2ENCONET,Miramarska20,10000Zagreb,Croatia
3ETSEIB,UniversitatPolit`ecnicadeCatalunya,AvenidaDiagonal647,08028Barcelona,Spain
4ComisionNacionaldeSeguridadNuclearySalvaguardias,Dr.Barragan779,Cuartopiso,ColoniaVertizNarvarte,
DelegacionBenitoJuarez,03020CiudaddeM´exico,DF,Mexico
CorrespondenceshouldbeaddressedtoC.M.Allison,[email protected]
Received11January2012;Accepted6March2012
AcademicEditor:AlejandroNun˜ez-Carrera
Copyright©2012C.M.Allisonetal.ThisisanopenaccessarticledistributedundertheCreativeCommonsAttributionLicense,
whichpermitsunrestricteduse,distribution,andreproductioninanymedium,providedtheoriginalworkisproperlycited.
ImmediatelyaftertheaccidentatFukushimaDaiichi,InnovativeSystemsSoftwareandothermembersoftheinternationalSCDAP
Development and Training Program started an assessment of the possible core/vessel damage states of the Fukushima Daiichi
Units1–3.Theassessmentincludedabriefreviewofrelevantsevereaccidentexperimentsandaseriesofdetailedcalculations
usingRELAP/SCDAPSIM.ThecalculationsusedadetailedRELAP/SCDAPSIMmodeloftheLagunaVerdeBWRvesselandrelated
reactorcoolingsystems.TheLagunaVerdemodelswereprovidedbytheComisionNacionaldeSeguridadNuclearySalvaguardias,
the Mexican nuclear regulatory authority. The initial assessment was originally presented to the International Atomic Energy
AgencyonMarch21tosupporttheiremergencyresponseteamandlatertoourJapanesememberstosupporttheirFukushima
Daiichispecificanalysisandmodeldevelopment.
1.Introduction The KIT experiments were limited to peak temperatures
less than 2600K and thus covered the initial stages of
ImmediatelyaftertheaccidentatFukushimaDaiichi,Inno- core heat up and melting including the liquefaction and
vative Systems Software (ISS) and other members of the relocation of BWR control blades, structural material, and
international SCDAP Development and Training Program fuel rod cladding. The in-pile experiments reached higher
(SDTP) [1, 2] started an assessment of the possible peaktemperaturesandincludedtheliquefactionofthefuel
core/vesseldamagestatesoftheFukushimaDaiichiUnits1– andotheroxidizedcladdingmaterialsandtheformationof
3.Theassessmentincludedabriefreviewofrelevantsevere ceramicmeltsandblockages.
accident experiments and a series of detailed calculations As described in Section3, a combination of RELAP/
using RELAP/SCDAPSIM [3, 4] for a representative BWR SCDAPSIM/MOD3.4andRELAP/SCDAPSIM/MOD3.5was
vesselandrelatedcoolingsystems. usedtoperformthedetailedcalculations.Bothversionsuse
AsdescribedbrieflyinSection2,theexperimentalreview publicallyavailableRELAP5/MOD3.2andMOD3.3thermal
presented to the IAEA emergency response team included hydraulic models and correlations in combination with the
representative highlights and phenomena identified from detailedfuelbehaviourandsevereaccident(SCDAP)models
separateeffectsexperimentsBWRspecificandotherbundle and correlations [9, 10]. The RELAP/SCDAPSIM code is
experiments [5, 6], performed by the Karlsruhe Institute of designed to predict the behavior of reactor systems during
Technology (KIT), and selected in-pile experiments [7, 8]. normal and accident conditions including severe accidents
2 ScienceandTechnologyofNuclearInstallations
CORA-18: BWR Temperature (◦C) Elevation (mm)
30◦ 120◦ 210◦ 300◦ (◦C)
mm
(mm)
1100 1905 1745
1158 702 334
1000
1530 1895 1585
1016 560 269
500
1825 1895 1525
874 554 254
1840 1835 930
859 413 112
0
Figure1:Posttestverticalviewandhorizontalmetallographiccross-sectionsforCORA-18.
uptothepointofreactorvesselfailure.MOD3.4,thecurrent As discussed in Section7, these calculations which
productionversion,hasbeenusedbymemberorganizations showed the core uncovery, fuel melting, and relocation of
andlicenseduserstosupportavarietyofapplicationsinclud- the fuel and other molten materials into the lower plenum
ing the design and analysis of severe accident experiments. canoccurratherquicklyonceemergencycoolingisnolonger
MOD3.5,anexperimentalversionofthecode,hasbeenused available.Thetimingofsucheventsdependsonthedelaysin
tosupportthedesignandanalysisoftheFrenchPHEBUS-FP the loss of emergency cooling as well as the specific details
[11,12],GermanQUENCH[13],andRussianPARAMETER of an accident scenario such as the opening of safety relief
[14]experimentsandincorporatesthelatestSCDAPmodel valvestodepressurizethevesselortheadditionofwaterafter
improvements. It is the main version now being used by partialorcompletecoreuncovery.
SDTP members and selected licensed users for Fukushima The conclusions from the results and discussions pre-
analysisandassessment. sentedinSections2,3,4,5,6and7arepresentedinSection
As described in Section4, the initial Laguna Verde 8. This section first presents initial conclusions regarding
RELAP/SCDAPSIM input models were provided by the possible core/vessel damage in Fukushima Daiichi Units 1–
Comision Nacional de Seguridad Nuclear y Salvaguardias 3. These initial conclusions are based on the information
(CNSNS), the Mexican nuclear regulatory authority. The publishedandcalculationsperformedinlateMarchof2011
LagunaVerdeBWRsareBWR5designswiththermalpower immediately after the accident. Next, general conclusions
of∼2370MWand444fuelassemblieswithanaveragecore foratypicalBWR,LagunaVerde,subjectedto“Fukushima-
burnupof3.57×105MWs/Kg.Forcomparison,Fukushima Daiichi-like” scenarios (short to extended periods of emer-
Daiichi Unit 1 has a thermal power of 1380MW from 400 gency core cooling after reactor scram, variation in vessel
assemblies.Units2and3havethermalpowersof2381MW pressure,andwaterinjectionaftercoreuncoveryandstartof
from548assemblies. coreheatup)arepresented.Finallytheconclusionsforsevere
The initial RELAP/SCDAPSIM calculations performed accidentmanagementstrategiesarepresented.
priortoMarch21stincludedaseriesofstationblackouttran-
sientswithavarietyofemergencycorecoolinganddepres-
surization strategies. Representative results are provided in 2.HighlightsofRelevant
Section5.AdditionalcalculationsperformedbetweenMarch ExperimentsandPhenomena
21st and 25th to support the IAEA emergency response
teamincludedavarietyofscenarioswithlossofemergency Theexistingdatabaseandsevereaccidentcodeandmodels
core cooling ranging from 0 to 70 hours after scram. developedoverthepast40yearssincetheaccidentatTMI-2
Representativeresultsfromtheseadditionalcalculationsare are considered to be adequate to predict the likely states
presentedinSection6. of a BWR core and vessel during “Fukushima-Daiichi-like”