Auswahl der wissenschaftlichen Literatur zum Thema „In-vessel melt retention“

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.

Considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Worldwide several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model has been de-veloped simulating the thermal processes and the viscoplastic behaviour of the ves-sel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal and the mechanical calculations are sequentially and recursively coupled. The model is capable of evalu-ating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post test calculations for the FOREVER test se-ries representing the lower head RPV of a PWR in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stock-holm. The results of the calculations can be summarised as follows: # The creeping process is caused by the simultaneous presence of high tem-perature (>600 °C) and pressure (>1 MPa) # The hot focus region is the most endangered zone exhibiting the highest creep strain rates. # The exact level of temperature and pressure has an influence on the vessel failure time but not on the failure position # The failure time can be predicted with an uncertainty of 20 to 25%. This uncer-tainty is caused by the large scatter and the high temperature sensitivity of the viscoplastic properties of the RPV steel. # Contrary to the hot focus region, the lower centre of the vessel head exhibits a higher strength because of the lower temperatures in this zone. The lower part moves down without significant deformation. Therefore it can be assumed, that the vessel failure can be retarded or prevented by supporting this range. # The development of a gap between melt crust and vessel wall could not be proofed. First calculations for a PWR geometry were performed to work out differences and commonalities between prototypic scenarios and scaled experiments. The results of the FOREVER-experiments cannot be transferred directly to PWR geometry. The geometrical, mechanical and thermal relations cannot be scaled in the same way. Because of the significantly higher temperature level, a partial ablation of the vessel wall has to be to expected in the PWR scenario, which is not the case in the FOREVER tests. But nevertheless the FOREVER tests are the only integral in-vessel retention experiments up to now and they led to a number of important insights about the behaviour of a vessel under the loading of a melt pool and pressure.

Considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Worldwide several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model has been de-veloped simulating the thermal processes and the viscoplastic behaviour of the ves-sel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal and the mechanical calculations are sequentially and recursively coupled. The model is capable of evalu-ating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post test calculations for the FOREVER test se-ries representing the lower head RPV of a PWR in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stock-holm. The results of the calculations can be summarised as follows: # The creeping process is caused by the simultaneous presence of high tem-perature (>600 °C) and pressure (>1 MPa) # The hot focus region is the most endangered zone exhibiting the highest creep strain rates. # The exact level of temperature and pressure has an influence on the vessel failure time but not on the failure position # The failure time can be predicted with an uncertainty of 20 to 25%. This uncer-tainty is caused by the large scatter and the high temperature sensitivity of the viscoplastic properties of the RPV steel. # Contrary to the hot focus region, the lower centre of the vessel head exhibits a higher strength because of the lower temperatures in this zone. The lower part moves down without significant deformation. Therefore it can be assumed, that the vessel failure can be retarded or prevented by supporting this range. # The development of a gap between melt crust and vessel wall could not be proofed. First calculations for a PWR geometry were performed to work out differences and commonalities between prototypic scenarios and scaled experiments. The results of the FOREVER-experiments cannot be transferred directly to PWR geometry. The geometrical, mechanical and thermal relations cannot be scaled in the same way. Because of the significantly higher temperature level, a partial ablation of the vessel wall has to be to expected in the PWR scenario, which is not the case in the FOREVER tests. But nevertheless the FOREVER tests are the only integral in-vessel retention experiments up to now and they led to a number of important insights about the behaviour of a vessel under the loading of a melt pool and pressure.

This thesis aims to provide the thermal condition of melt pool convection by CFD simulation, which is important to the assessment of the invessel melt retention (IVR) strategy widely adopted in Generation III pressurized water reactors (PWRs). As a severe accident mitigation measure, the IVR strategy is realized through external cooling of the lower head of a reactor pressure vessel (RPV). To achieve the coolability and retention of the corium pool in the RPV lower head, the heat flux at the outer surface of the vessel should be less than the critical heat flux (CHF) of boiling around the lower head. Under such condition, the integrity of the RPV is guaranteed by the adequate thickness of the unmelted vessel wall. The thesis work starts from the selection and validation of a turbulence model in the CFD computational tool chosen (Fluent). Afterwards a numerical model is set up for estimation of melt pool heat transfer of a reference PWR with the power capacity of 1000 MWe, including a mesh sensitivity study. Based on the numerical model of a twolayer melt pool, four tasks are carried out to investigate the effects of Zr oxidation ratio, Fe content, and radiation emissivity on heat flux profiles, as well as the focus effect under extreme conditions. Selection and validation of the turbulence model are conducted by comparing the simulation results of different turbulence models with the DNS data on the convection of volumetrically heated fluid layer bounded by rigid isothermal horizontal walls at equal temperature. The internal Rayleigh numbers of the flow reach up to 10e6. The comparison shows a good agreement of the SST k-ω turbulence model results with the DNS data. The simulations with the Zr oxidation ratio of 0, 0.2 and 0.5, correspondingly, the oxide layer of 1.389m, 1.467m and 1.580m, and the metal layer of 0.705m, 0.646m and 0.561m in height, show that, the temperature of the oxide layer will increase with Zr oxidation ratio, while the temperature of the metal layer will decrease resulting in more heat transfer through the oxide layer sidewall and less top radiation. Nevertheless, the effect of the Zr oxidation ratio is not pronounced in the range of 00.5. The simulations with the Fe mass of 22t, 33t and 45t, and respective height of the metal layer of 0.462m, 0.568m and 0.646m, show that, the inner metal layer will significantly increase the temperatures of both the metal layer and the oxide layer. The percentage of heat transfer at the oxide layer sidewall will increase to supplement the reduction of that at the metal layer. The simulations with the radiation emissivity of 0.2, 0.35, 0.45 and 0.7 show that, the emissivity below 0.45 has an impact on heat transfer, and the temperatures and sidewall heat flux of both the oxide layer and the metal layer will increase with decreasing emissivity. The impact is negligible when the emissivity is above 0.45. The simulations under the hypothetically extreme conditions with either an adiabatic top boundary or a very thin metal layer show the focusing effect may occur, i.e., the heat flux through the metal sidewall is larger than that in the oxide layer. But the local high heat flux is flattened by the vessel wall with good heat conductivity. In summary, the simulations demonstrate that, except for the cases under extreme conditions, the heat fluxes of the melt pools in all other cases are significantly lower than the CHF of external cooling of the lower head. Therefore, the safety margin of the IVR strategy of the PWR chosen is seems sufficient. However, due to some limitations (e.g., simplification and assumptions) in the simulation cases and coupling of different influential factors, as indicated by the present study, the precise predictions of heat flux under all scenarios are still difficult. Therefore, the conclusions could not be generalized to the other conditions or other configurations of the molten pools. By discussing the model and simplifications/assumptions adopted in this work, the improvement directions of the numerical model and other perspectives are proposed at the end of the thesis.
Denna avhandling syftar till att tillhandahålla det termiska tillståndet för smältbassängskonvektion genom CFD-simulering, vilket är viktigt för bedömningen av IVR-strategin som allmänt antagits i tryckvattenreaktorer (PWR) i Generation III. Som en åtgärd för att mildra allvarliga olyckor realiseras IVR-strategin genom extern kylning av det nedre huvudet av ett reaktortryckkärl (RPV). För att uppnå kylbarhet och kvarhållning av koriumbassängen i det nedre RPV-huvudet bör värmeflöde vid den yttre ytan av kärlet vara mindre än det kritiska värmeflödet (CHF) som kokar runt det nedre huvudet. Under sådant tillstånd garanteras RPV: s integritet av den osmälta kärlväggens tillräckliga tjocklek. Examensarbetet startar från valet och valideringen av en turbulensmodell i det valda CFD-beräkningsverktyget (Fluent). Därefter sätts en numerisk modell upp för uppskattning av smältbassängens värmeöverföring av en referens PWR med en effektkapacitet på 1000 MWe, inklusive en nätkänslighetsstudie. Baserat på den numeriska modellen för en tvålagers smältbassäng utförs fyra uppgifter för att undersöka effekterna av Zr-oxidationsförhållande, Fe-innehåll och strålningsemissivitet på värmeflödesprofiler, liksom fokuseffekten under extrema förhållanden. Val och validering av turbulensmodellen utförs genom att jämföra simuleringsresultaten för olika turbulensmodeller med DNS-data för konvektionen av volymetriskt uppvärmt fluidskikt avgränsat av styva isoterma horisontella väggar vid lika temperatur. De interna Rayleigh-siffrorna i flödet når upp till 10e6. Jämförelsen visar att SST k-ω turbulensmodellresultaten överensstämmer med DNS-data. Simuleringarna med Zr-oxidationsförhållandet 0, 0,2 och 0,5, motsvarande oxidskiktet på 1,389 m, 1,467 m och 1,580 m, och metallskiktet på 0,705 m, 0,664 m och 0,561 m i höjd, visar att temperaturen av oxidskiktet kommer att öka med Zr-oxidationsförhållandet, medan metallskiktets temperatur kommer att minska vilket resulterar i mer värmeöverföring genom oxidskiktets sidovägg och mindre toppstrålning. Ändå är effekten av Zr-oxidationsförhållandet inte uttalad i intervallet 00,5. Simuleringarna med Fe-massan på 22t, 33t och 45t och respektive höjd av metallskiktet på 0,462m, 0,568m och 0,664m visar att det inre metallskiktet avsevärt kommer att öka temperaturerna för både metallskiktet och oxiden lager. Andelen värmeöverföring vid oxidskiktets sidovägg ökar för att komplettera minskningen av den vid metallskiktet. Simuleringarna med strålningsemissiviteten 0,2, 0,35, 0,45 och 0,7 visar att emissiviteten under 0,45 påverkar värmeöverföringen, och temperaturerna och sidoväggens värmeflöde för både oxidskiktet och metallskiktet kommer att öka med minskande emissivitet. Effekten är försumbar när strålningen är över 0,45. Simuleringarna under de hypotetiskt extrema förhållandena med antingen en adiabatisk övre gräns eller ett mycket tunt metallskikt visar att fokuseringseffekten kan uppstå, dvs. värmeflödet genom metallsidan är större än det i oxidskiktet. Men det lokala höga värmeflödet plattas ut av kärlväggen med god värmeledningsförmåga. Sammanfattningsvis visar simuleringarna att, förutom fall under extrema förhållanden, är värmeflödet från smältpoolerna i alla andra fall betydligt lägre än CHF för extern kylning av nedre huvudet. Därför verkar säkerhetsmarginalen för IVR-strategin för den valda PWR tillräcklig. På grund av vissa begränsningar (t.ex. förenkling och antaganden) i simuleringsfall och koppling av olika inflytelserika faktorer, vilket indikeras av den aktuella studien, är de exakta förutsägelserna av värmeflöde under alla scenarier fortfarande svåra. Därför kunde slutsatserna inte generaliseras till de andra förhållandena eller andra konfigurationer av de smälta poolerna. Genom att diskutera modellen och förenklingar / antaganden som antagits i detta arbete föreslås förbättringsriktningarna för den numeriska modellen och andra perspektiv i slutet av avhandlingen.

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Vom Institut für Sicherheitsforschung des FZD wurden Finite-Elemente-Modelle erstellt, die sowohl die Temperaturfeldberechnung für die Wand als auch die elastoplastische Mechanik der Behälterwand beschreibt. Die thermischen und mechanischen Berechnungen sind gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Es existieren Modelle für die Druckwasserreaktortypen KONVOI und WWER-1000. Es wurden prototypische Szenarien mit und ohne externe Flutung des RDB untersucht, wobei die homogen und die segregierte Schmelzepoolkonfiguration betrachtet wurden. Zusätzlich wurde eine bruchmechanische Bewertung des Thermoschocks, der durch die externe Flutung entsteht, vorgenommen. Auf Grundlage der Experimente im Rahmen des ISTC-Projekts METCOR wurde außerdem die Auswirkung der thermochemischen Wechselwirkung zwischen Corium-Schmelze und RDB-Wand auf das Versagensverhalten des RDB untersucht. Das wichtigste Ergebnis ist, dass eine erfolgreiche Schmelzerückhaltung im RDB auch bei größeren Reaktoren möglich erscheint, wenn eine rechtzeitige Flutung der Reaktorgrube gelingt. Mittels einer statistischen Analyse wurden die Empfindlichkeiten von Ergebnissen gegenüber den Eingangsparametern und die Unsicherheiten der Ergebnisse quantifiziert. Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZD finite element models have been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. The thermal hydraulic and the mechanical calculations are coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. Models exist for the pressurised water reactor types KONVOI and VVER-1000. Prototypic scenarios with and without external flooding were investigated with consideration of homogeneous and segregated melt pool configurations. Additionally a fracture mechanic evaluation of the thermal shock, originating from the external flooding, was performed. Based on the experimental results of the ISTC project METCOR, the effects of the thermal chemical interaction between corium melt and vessel steel were investigated in the IVR scenarios. An important result of the project is that a successful in-vessel melt retention seems to be possible even for large reactors if the reactor pit can be filled with water before the corium melt is relocated to the lower plenum. By means of statistical analysis the sensitivity of results against input parameter variations was studied. The uncertainty of results was quantified.

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Vom Institut für Sicherheitsforschung des FZD wurden Finite-Elemente-Modelle erstellt, die sowohl die Temperaturfeldberechnung für die Wand als auch die elastoplastische Mechanik der Behälterwand beschreibt. Die thermischen und mechanischen Berechnungen sind gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Es existieren Modelle für die Druckwasserreaktortypen KONVOI und WWER-1000. Es wurden prototypische Szenarien mit und ohne externe Flutung des RDB untersucht, wobei die homogen und die segregierte Schmelzepoolkonfiguration betrachtet wurden. Zusätzlich wurde eine bruchmechanische Bewertung des Thermoschocks, der durch die externe Flutung entsteht, vorgenommen. Auf Grundlage der Experimente im Rahmen des ISTC-Projekts METCOR wurde außerdem die Auswirkung der thermochemischen Wechselwirkung zwischen Corium-Schmelze und RDB-Wand auf das Versagensverhalten des RDB untersucht. Das wichtigste Ergebnis ist, dass eine erfolgreiche Schmelzerückhaltung im RDB auch bei größeren Reaktoren möglich erscheint, wenn eine rechtzeitige Flutung der Reaktorgrube gelingt. Mittels einer statistischen Analyse wurden die Empfindlichkeiten von Ergebnissen gegenüber den Eingangsparametern und die Unsicherheiten der Ergebnisse quantifiziert. Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZD finite element models have been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. The thermal hydraulic and the mechanical calculations are coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. Models exist for the pressurised water reactor types KONVOI and VVER-1000. Prototypic scenarios with and without external flooding were investigated with consideration of homogeneous and segregated melt pool configurations. Additionally a fracture mechanic evaluation of the thermal shock, originating from the external flooding, was performed. Based on the experimental results of the ISTC project METCOR, the effects of the thermal chemical interaction between corium melt and vessel steel were investigated in the IVR scenarios. An important result of the project is that a successful in-vessel melt retention seems to be possible even for large reactors if the reactor pit can be filled with water before the corium melt is relocated to the lower plenum. By means of statistical analysis the sensitivity of results against input parameter variations was studied. The uncertainty of results was quantified.

Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluated between the FOREVER-experiments and a prototypical scenario. The thermodynamic and the mechanical model can be coupled recursively to take into account the mutual influence. This approach not only allows to consider the tem-perature dependence of the material parameters and the thermally induced stress in the mechanical model, it also takes into account the response of the temperature field itself upon the changing vessel geometry. New approaches are applied in this work for the simulation of creep and damage. Using a creep data base, the application of single creep laws could be avoided which is especially advantageous if large temperature, stress, and strain ranges have to be covered. Based on experimental investigations, the creep data base has been de-veloped for an RPV-steel and has been validated against creep tests with different scalings and geometries. It can be stated, that the coupled model is able to exactly describe and predict the vessel deformation in the scaled integral FOREVER-tests. There are uncertainties concerning the time to failure which are related to inexactly known material parame-ters and boundary conditions. The main results of this work can be summarised as follows: Due to the thermody-namic behaviour of the large melt pool with internal heat sources, the upper third of the lower RPV head is exposed to the highest thermo-mechanical loads. This region is called hot focus. Contrary to that, the pole part of the lower head has a higher strength and therefore relocates almost vertically downwards under the combined thermal, weight and internal pressure load of the RPV. On the one hand, it will be possible by external flooding to retain the corium within the RPV even at increased pressures and even in reactors with high power (as e.g. KONVOI). On the other hand, there is no chance for melt retention in the considered scenario if neither internal nor external flooding of the RPV can be achieved. Two patents have been derived from the gained insights. Both are related to pas-sively working devices for accident mitigation: The first one is a support of the RPV lower head pole part. It reduces the maximum mechanical load in the highly stressed area of the hot focus. In this way, it can prevent failure or at least extend the time to failure of the vessel. The second device implements a passive accident mitigation measure by making use of the downward movement of the lower head. Through this, a valve or a flap can be opened to flood the reactor pit with water from a storage res-ervoir located at a higher position in the reactor building. With regard to future plant designs it can be stated - differing from former presump-tions - that an In-Vessel-Retention (IVR) of a molten core is possible within the reac-tor pressure vessel even for reactors with higher power.

Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluated between the FOREVER-experiments and a prototypical scenario. The thermodynamic and the mechanical model can be coupled recursively to take into account the mutual influence. This approach not only allows to consider the tem-perature dependence of the material parameters and the thermally induced stress in the mechanical model, it also takes into account the response of the temperature field itself upon the changing vessel geometry. New approaches are applied in this work for the simulation of creep and damage. Using a creep data base, the application of single creep laws could be avoided which is especially advantageous if large temperature, stress, and strain ranges have to be covered. Based on experimental investigations, the creep data base has been de-veloped for an RPV-steel and has been validated against creep tests with different scalings and geometries. It can be stated, that the coupled model is able to exactly describe and predict the vessel deformation in the scaled integral FOREVER-tests. There are uncertainties concerning the time to failure which are related to inexactly known material parame-ters and boundary conditions. The main results of this work can be summarised as follows: Due to the thermody-namic behaviour of the large melt pool with internal heat sources, the upper third of the lower RPV head is exposed to the highest thermo-mechanical loads. This region is called hot focus. Contrary to that, the pole part of the lower head has a higher strength and therefore relocates almost vertically downwards under the combined thermal, weight and internal pressure load of the RPV. On the one hand, it will be possible by external flooding to retain the corium within the RPV even at increased pressures and even in reactors with high power (as e.g. KONVOI). On the other hand, there is no chance for melt retention in the considered scenario if neither internal nor external flooding of the RPV can be achieved. Two patents have been derived from the gained insights. Both are related to pas-sively working devices for accident mitigation: The first one is a support of the RPV lower head pole part. It reduces the maximum mechanical load in the highly stressed area of the hot focus. In this way, it can prevent failure or at least extend the time to failure of the vessel. The second device implements a passive accident mitigation measure by making use of the downward movement of the lower head. Through this, a valve or a flap can be opened to flood the reactor pit with water from a storage res-ervoir located at a higher position in the reactor building. With regard to future plant designs it can be stated - differing from former presump-tions - that an In-Vessel-Retention (IVR) of a molten core is possible within the reac-tor pressure vessel even for reactors with higher power.

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.

Abstract In-Vessel Core Melt Retention (IVMR) strategy via external vessel cooling is widely applied for reactors of relatively low power like VVER-440. In this study, IVMR strategy was applied for Rivne-1, 2 units to prove the pressure vessel integrity. Based on initial data like heat flux for internal wall and external wall temperature, a series of calculations for different scenarios were performed. These calculations include non-elastic material properties: creep and plasticity. As the result, the wall ablation, radial displacements, stress and strains were obtained. To prove pressure vessel integrity four criterions have been checked. The first one is obvious — remaining wall thickness, to prove that that RPV won’t be melted right through. The second one is visco-plastic collapse — lack of monotonous increase in deformations, in case of FEM solution result convergence can be interpreted as resist against such failure. The third — sustainable external cooling, thus the gap between RPV (due to radial elongation) and thermal protection shield must be 10 mm at least. The last one is brittle strength, this calculation was performed on a separate model.

A helical ex-vessel cooling device is being designed and evaluated to insure retention of the molten corium that may discharge from the failed lower head of a reactor pressure vessel (RPV) during postulated severe accidents of a light water reactor. This conceptual design consists of four functional parts located in the reactor cavity, including an intermediate chamber for temporary retention of corium, an helical channel for corium spreading, cooling pipes embedded in the channel wall and venting system. The heat flux profile and steam generation along the cross section of the spreading channel T are analyzed under several conservative assumptions. The results show the melt coolability is achievable even under Benrenson film boiling regime.

In-vessel retention (IVR) of core melt through external reactor vessel cooling (ERVC) is a key severe accident management strategy to ensure that the vessel head remains intact and eliminate consequent major threats to containment integrity. To maintain the margin against the failure of the reactor vessel and make sure of the feasibility of IVR for its role to confine the molten corium with 1400MW power, CAP1400, Chinese version of large passive PWR, systematic investigations including experimental and analytical researches of IVR are very important to the development of CAP1400. This paper briefly reviews the progress and tasks of a four-year project which was planned to analyze, evaluate, improve and validate the effectiveness of IVR employed in the CAP1400.

To precisely understand the accident process of reactor core melt in In-vessel retention (IVR) condition, 3-dimensional transient thermal conduction analysis with moving boundary is performed on quarter reactor core model. The decline of decay power and water level in reactor pressure vessel (RPV), and the radial distribution of assemblies of different material is considered. Convective heat transfer on rod surface and coolant interface is computed with empirical correlation of natural convection of saturated steam vapor / water. Radiation heat transfer with 16 neighboring rod is considered. Also, a dynamic ablation model is developed to simulate the ablation of lower support plate (LSP) caused by continuously accumulation of molten corium. The impingement heat transfer of the falling corium and the molten pool formed in LSP ablation cavity is taken into account. The simulation gives the ablation process on the surface of LSP as well as temperature history and molten proportion of the reactor core, which shows agreement with reference. Simulation shows: the melt process of reactor core accelerated in the accident process of 2600s, when coolant in RPV dry up. 65% of the core mass has molten at 8000 second. LSP is completely penetrated in 6000s, the ablation of LSP is mainly focused on an annular region of radius 700mm.

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and retain the core melt in the reactor pressure vessel. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactors below 500MWe, like VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000. One of the main issues for the demonstration of the success of the IVR strategy lies in the evaluation of the transient heat fluxes applied by the corium pool along the vessel wall. Indeed, these transient heat fluxes, during the corium pool stratification evolution, are expected to be higher than the steady-state ones, in particular due to the concentration of the heat flux in the top metal layer when it is thin (so called focusing effect). Another issue appears when a heavy metal is initially formed and rises later to the top (inversion of stratification): in such a situation, the metal goes through the oxide phase and accumulates a significant superheat which is likely to produce a high transient heat flux. Thus, it is of primary importance to be able to evaluate the duration of these transient peaks in order to evaluate the minimal residual vessel thickness after such fast transient ablation and draw conclusions about the vessel integrity. This paper first presents the phenomenology associated to the transient molten pool stratification and the model implemented in the severe accident integral code ASTEC (Accident Source Term Evaluation Code) to evaluate this kinetics. Then, evaluations are presented, based on a typical PWR reactor configuration. A sensitivity study is proposed to consider the impact of the main uncertainties on parameters which govern this kinetics. A particular focus is made on the physical phenomena driving the transient stratification of material layers in the corium pool and on the identification of critical situations with possible consequences in terms of vessel failure. The characteristic times of each individual process (chemistry, stratification, natural convection) are compared. In particular, the limiting cases of very fast chemistry or very slow chemistry are evaluated. This work is performed in the frame of the European H2020 project IVMR (In-Vessel Melt Retention) coordinated by IRSN. This project has been launched in 2015 and gathers 27 organizations with, as main objective, the evaluation of feasibility of IVR strategy for LWR (PWR, VVER, BWR) of total power 1,000MWe or higher.

One of the main goals of severe accident management strategies is to mitigate radiological releases to people and environment. To choose the most appropriate strategy, one needs to know the probability of its success taking into account the associated uncertainties. In the field of corium and debris behavior and coolability, research programs are still on going and the possibilities to efficiently cool and retain corium and debris inside the Reactor Pressure Vessel (RPV) then inside the containment are difficult to evaluate. This leads to uncertainties in safety assessments particularly when margins to RPV or containment failure are too weak. In Vessel Melt Retention (IVMR) strategies for Light Water Reactors (PWR, BWR, VVER) intend to stabilize and retain the core melt in the RPV (as it happened during the TMI-2 accident). This would reduce significantly the threats to the last barrier (the containment) and therefore reduce the risk of release of radioactive elements to the environment. This type of Severe Accident Management (SAM) strategy has already been incorporated recently in the SAM guidance (SAMG) of several operating medium size Light Water Reactors (reactor below 500MWe (like VVER440)) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000. A European project coordinated by IRSN and gathering 23 organizations (Utilities, Technical Support Organizations, Nuclear Power Plant vendors, Research Institutes…) has been launched in 2015 with as main objective the evaluation of feasibility of IVMR strategies for Light Water Reactors (PWR, VVER, BWR) of total power around 1000MWe (which represent a significant part of the European Nuclear Power Plants fleet). This paper intends to show how it is possible to introduce transient evolutions of the stratified corium pool in the evaluation of the heat flux profile along the vessel wall. Indeed, due to chemical reactions in the U–Zr–O–Fe molten pool, separation between non-miscible metallic and oxide phases may occur, modifying the thermal load applied to the RPV. If stabilized stratified corium configurations are well defined and modeled, transient evolutions of material layers in the corium pool are still difficult to predict. The evaluations presented are based on calculations performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. The modeling of transient evolution of corium layers leads to configurations with a thin light metal layer on top of the oxidic one, increasing the so called “focusing effect” (intense heat fluxes on the RPV walls adjacent to the top metal layer). A sensitivity study on some uncertain parameters is proposed to evaluate the impact on the kinetics of layers inversion. Depending on the duration of these transient heat fluxes, the mechanical strength of the RPV could be challenged.

An evaluation of the ex-vessel core catcher system of a sample advanced light water reactor was presented. The core catcher was designed to cool down the molten corium through a combined injection of water and gas from the bottom of the molten corium, which could be effective in the reduction of rapid steam generation and in the suppression of a steam explosion. By using the MELCOR code, a scenario analysis was performed for a representative severe accident scenario of the ALWR, that is, the 6-inches large break loss of coolant accident without safe injection. The corium spreading regime was estimated by an asymptotic calculation. The composition of the molten corium, the decay power level, and the sacrificial concrete ablation depth with time were obtained by a sacrificial concrete ablation analysis. The corium cooling history in the core catcher during the coolant injection was evaluated to calculate the temporal steam generation rate by considering an energy conservation equation. These were used as the major inputs for the temporal calculations of containment pressure which was performed by using the GASFLOW code. Several cases with change of water and gas injection rates were calculated. It was confirmed that the bottom water injection system was an effective corium cooling method in the ex-vessel core catcher to preclude a possible steam explosion and to suppress the quick release of steam.

After Fukushima nuclear power plant accident, severe accident is getting more and more concerns all over the world. In order to mitigate severe accident and improve the safety of nuclear power plant, two different strategies are applied in different plants. One is in-vessel melt retention strategy, and the other is ex-vessel melt retention strategy. Tianwan nuclear power plant is an improved Gen II nuclear power plant and in-vessel melt retention strategy is adopted in the plant. In order to achieve this strategy, cavity injection system is designed for the plant. Probabilistic Safety Analysis is the most commonly used quantitative risk assessment tool for decision-making in selecting the optimal design among alternative options. For this plant, in order to optimize the design of cavity injection system, improve the safety level of nuclear power plant, and meanwhile, improve the engineering implementation and economization, Level 2 PSA was used for this decision-making process. In this paper, the Level 2 PSA for this plant and the application for the design of cavity injection system are introduced.

In-vessel melt retention has drawn renewed concern as an important severe accident management measure in existing and advanced light water reactors. Despite numerous studies the central question whether the maximum heat flux in a melt pool could exceed the critical heat flux (CHF) is not fully answered. The uncertainty comes from the variety of accident scenarios and the corresponding melt pool configurations, as well as from the applicability of the experimental results to the reactor case. It is therefore necessary to examine the melt heat transfer under different pool configurations and cooling conditions, as well as to compare the experimental results coming from different test vessel geometries and cooling regimes. This study investigates the heat transfer characteristics of an oxidic pool in the PWR lower plenum in the case when the vessel wall is externally cooled by water, and the melt upper surface is free in a closed insulated environment. Thus the melt pool cooling conditions are quasi-isothermal at the inclined sidewall and at the upper surface free surface with thermal radiation. This pool configuration can occur before the melt layer stratification begins or the melt pool is composed only of oxide melt under certain melt relocation sequences. A non-eutectic nitrate mixture with the composition of 20% NaNO3−80% KNO3 in mole relation is used as the simulant melt. Besides the determination of melt temperature and heat flux in their global average values, emphasis are given on the characterization of the axial distribution of melt temperature and heat flux at different power densities and pool heights. Results obtained in hemispherical geometry are analyzed and compared with other studies conducted under similar boundary conditions. The characterization of the heat flux distribution provide important data for the prediction of the maximum heat flux in the reactor case with similar boundary conditions and the evaluation of the concept of in-vessel melt retention by external cooling.