Search

• The KAIST model for fragment-size distribution is developed using upper-limit log-normal distribution. • The suggested model is validated against FARO and TROI experiment. • The suggested model will reduce computational effort for severe accident analysis. As molten corium is ejected from a reactor pressure vessel (RPV), the melt undergoes fuel-coolant interaction (FCI). Through the FCI molten corium is fragmented into small particles, and the fragments will form a debris bed stacking on the bottom of the cavity. Heat removal performance from both fragments and the debris bed is highly related to the size of fragments. Also, as FCI is a thermal-hydraulic interaction on the interface between the corium and the coolant, the estimation of the Sauter mean diameter is required. Thus, in this study, we develop a simplified strategy to predict the Sauter mean diameter with particle size distribution. The model is proposed to calculate the particle size distribution based on upper-limit log-normal distribution and the critical Weber number criterion. The calculation results are compared with 9 cases of FARO experiments and 7 cases of TROI experiments. The particle size distribution calculated with the suggested method well follows the trend of measured distribution from experiments. The estimated Sauter mean diameter varies from 1.58 mm to 3.13 mm, while the results for the ratio of the mass mean diameter to the Sauter mean diameter is almost constant to 1.55. It turns out that the suggested model gives good predictions of the particle size distribution and the Sauter mean diameter without large computational effort. [ABSTRACT FROM AUTHOR]

Within the framework of the severe accident mitigation strategy for GEN-IV reactors, the behaviour of molten corium in the SFR core catcher is a subject of significant interest. This article aims to define the main considerations within the core catcher during a prospective severe accident scenario, in particular with regard to the rate of heat transfer to the core catcher structures. On the basis of the core catcher design adopted at the end of the ASTRID project, a simplified model is proposed to conservatively evaluate the quasi-steady state behaviour in a stratified oxide-metal molten corium pool with mechanically stable crusts at the interfaces between the stratified layers and at the core catcher walls. The corium oxide layer is subject to significant internal heat generation due to the radioactive decay heat. The proposed model implies that the core catcher should withstand reasonable worst case estimates for the thermal load from the corium pool, avoiding sodium boiling and thermal ablation of the core catcher structures. Thermal ablation is only predicted in the cases of an order of magnitude under-prediction of the heat transfer coefficients at the lateral boundary of the corium oxide layer, or for local heat flux values in excess of 3.7 times the surface average. Further investigations will be dedicated to the revision of the more conservative model assumptions, in particular regarding the magnitude of the heat flux from high aspect ratio encloses with large internal Rayleigh numbers. A refined transient version of the model is intended for implementation within the PROCOR platform developed by the CEA. • A liquid corium pool on the core catcher of a SFR is investigated. • A simplified model is proposed to evaluate the system at quasi-steady state. • The core catcher remains ablation resistant under predicted therrmal loads. • A refined transient model is being implemented in the CEA PROCOR platform. [ABSTRACT FROM AUTHOR]

In the event of a severe nuclear accident such as that at the Fukushima Daiichi Nuclear Power Plant (NPP) in 2011, it is essential to minimize the release of radioactive material into the environment. Studies on Canada Deuterium Uranium (CANDU) reactors showed that thermal stresses are a potential threat to vessel wall integrity. There is thus a need to analyse the heat transfer and fluid flow within the corium and determine the spatial distribution of heat flux at the calandria vessel wall to ensure the ex-vessel cooling is sufficient. Canadian Nuclear Laboratories (CNL) designed and conducted an experiment to analyse the heat transfer within the corium using a 1/5 scale CANDU calandria vessel with molten salts as the corium simulant. The current study used computational fluid dynamics (CFD) to simulate the heat transfer and crust formation in the experiment to provide insight into the flow pattern and heat transfer and to assess the adequacy of the CFD modelling. Unsteady Reynolds-Averaged Navier–Stokes (URANS) and enthalpy-based solidification models were used to simulate the experiment. The numerical results show that crust forms along the vessel wall up to an approximate polar angle of 42 °. There are natural convection circulations close to the top surface, and boundary layers form along the vessel wall. The results predicted by the numerical model match closely with the experimental results. • The Canadian Nuclear Laboratories convection experiment was modelled numerically. • Results show that crust forms at the boundaries. • Additionally, results showed three are three regions in the liquid domain. • Comparison between the experimental and numerical results shows a good agreement. [ABSTRACT FROM AUTHOR]

Highlights • Detailed study of suitable numerical methods to simulate spreading Corium flows. • Comparison of advection schemes for the discretisation of the energy equation. • Novel viscosity evaluation method that mitigates unphysically large viscosity values. • Substantial improvements for the prediction of strongly temperature-dependent fluids. Abstract This paper investigates the numerical simulation of Corium spreading, focussing on modelling the two fluid Corium/air system. The complex Corium spreading flow is simplified by assuming the Corium is a homogeneous non-reactive fluid with a strongly temperature-dependent viscosity. A detailed investigation of the numerical methods most suitable to simulate Corium spreading flows is conducted, focusing on numerical errors, such as numerical diffusion, associated with the spatial discretisation of the governing equations. Numerical errors in the prediction of temperature may significantly affect the flow prediction because of the temperature-dependency of the viscosity. To this end, advection schemes to discretise the energy equation widely applied in the literature are found to be too diffusive to model the flow of interest, and the application of a discretisation scheme originally developed for the discretisation of sharp interfaces is shown to greatly reduce the numerical diffusion of temperature at the interface between the two immiscible fluids. While numerical diffusion can be limited using a suitable discretisation scheme, additional numerical errors in the calculation of the temperature-dependent viscosity arise as a result of the implicit temperature averaging inherent to the VOF method. Various viscosity averaging and smoothing strategies are tested, with the aim to mitigate errors associated with the viscosity calculation at the interface, and an algorithm to improve the calculation of the temperature-dependent viscosity of Corium in interfacial cells is proposed. The different discretisation schemes, viscosity calculation methods and smoothing strategies are tested on a spreading test case where a hot dense fluid with temperature-dependent viscosity, similar to Corium, spreads under the influence of gravity in a domain filled with a light cold fluid, highlighting the significant influence of the applied numerical methods for the simulation of Corium spreading. [ABSTRACT FROM AUTHOR]

Highlights • A novel algorithm calculating viscosity after pressure is proposed for MPS. • Numerical creeping of highly viscous fluid is eliminated by the new algorithm. • Viscosity escalation can reasonably represent crust formation with the new algorithm. • VULCANO VE-U7 experiment demonstrates improvement of the new algorithm. Abstract Corium (lava-like mixture of fissile material) spreading prediction is of great significance in the severe accidents of nuclear power plants. Crust formation due to solidification distinguishes corium spreading from common isothermal spreading. The Lagrangian moving particle semi-implicit (MPS) method is potential for such spreading flow with both free surface and crust-melt interface. Crust formation is usually represented by viscosity escalation, but crust creeping is an associated problem. In the original MPS algorithm, creeping velocity cannot be reduced steadily by the continuous increase of viscosity, owing to the numerical creeping. A new solution algorithm is proposed for particle methods to eliminate such numerical creeping, so that creeping velocity decreases proportionally with viscosity rise. In this situation, high enough viscosity can effectively represent crust behaviors. Three numerical examples, leakage flow with high viscosity, dam break flow with low viscosity and the VULCANO VE-U7 corium spreading experiment with both high and low viscosities simultaneously, are investigated to contrast the performance difference between the original and new algorithms. It is demonstrated that the current algorithm is suitable for crust formation in corium spreading. [ABSTRACT FROM AUTHOR]

Highlights • The present model can predict the morphology of the crust using two simple thermal stress criteria. • Low dependency of heat flux on pressure can be predicted considering thermal limit of quench front propagation. • The vapor flowing out from the crust significantly contributes to the removal of thermal energy of corium. • Effect of water ingression on removal of thermal energy of corium is prominent for corium with low concrete content. Abstract During an LWR severe accident, top flooding strategy is usually used in order to achieve the coolability of corium. Water ingression phenomenon, where water percolates through cracks in the crust, is identified as one of the phenomena that enhance the heat removal during cooling of the corium via top flooding. In this study, a model for the water ingression phenomenon is suggested. The model assumes that the morphology of fractured crust is determined during the period before onset of water ingression. Especially, the present model predicts the morphology of the crust using two thermal-stress based criteria: a tensile-strength criterion and a toughness criterion for the fracture. In addition, unlike previous models that assume heat removal is limited hydraulically to the dryout heat flux, heat flux during water ingression is estimated considering the phenomenon where quenching is limited thermally by large thermal energy of crust. When the current model was used to estimate the heat removal rate during the quenching of the fractured crust, it showed 26.5% of root-mean-square error (RMSE) for data (Lomperski and Farmer, 2007), which was improved from 39.3% RMSE of the existing hydraulic-limit based models (Lomperski and Farmer, 2007). In addition, comparing removed corium energies calculated from the present model with one calculated without considering water ingression, we found that water ingression can remove three-times larger energy when the concrete content in the corium is lower than 5 wt%. [ABSTRACT FROM AUTHOR]

• Parameters related to debris bed shape on pre-flooded cavity were summarized. • Empirical model was developed to predict ex-vessel debris bed shape. • Debris bed shape model with poly-disperse particles was also confirmed. • Model predicted bed shape from experiments within 20% error. To secure the containment integrity of a nuclear power plant (NPP) in severe accident situation of ex-vessel corium release into the reactor cavity, a pre-flooding strategy of a reactor cavity has been adopted in various NPPs, including Korean PWRs. Molten corium released into the pre-flooded water pool is fragmentized and forms a particulate debris bed at the bottom of the cavity. To prevent the containment failure, it is important to properly remove the heat from the debris bed, which can be accomplished by water ingression into the bed. The water ingression characteristics are highly dependent on the debris bed shape. In this study, several main parameters affecting the debris bed shape were determined with respect to phenomena of free-fall & sedimentation, particle dispersion by steam spike and convection flow, and self-leveling. Then empirical model was developed and validated using the results of DEFCON (Debris bed Formation and COolability experimeNt) experiments. The model's error was within ±10% when only free-fall & sedimentation were present. When steam-spike or convection was added to free-fall & sedimentation condition, the error increased to within ±20%. When all three were simultaneously present (free-fall & sedimentation, steam-spike, and convection), the model predicted most of the experimental results within ±20% error, except for several cases where convection had a significant impact. The effect of self-leveling itself was within ±10%. Considering all phenomena related to particulate debris bed formation, most cases had an error of ±20%. In addition, the debris bed shape formed by poly-disperse particles was predicted appropriately by the mode using the mass mean diameter of the particle size distribution, with an error range of ±20%. Consequently, it is expected that the proposed model of the debris bed shape can be utilized in an integrated severe accident analysis code for the evaluation on the ex-vessel debris bed coolability in the pre-flooded reactor cavity. [ABSTRACT FROM AUTHOR]

• Modified OpenFOAM accurately predicts large-scale 2-phase flow, conjugate heat transfer & boiling in CE-PECS using temp-dependent properties. The PECS (Passive Ex-vessel corium retaining and Cooling System) is installed inside the EU (European Union)-APR (Advanced Power Reactor) 1400 to prevent CCI (corium-concrete interaction) and to cool down the corium under an ex-vessel severe accident. The main phenomenon in the PECS is the natural convection of the cooling water generated by the heat from the corium. A verification experiment on the cooling capability of the PECS has been conducted in the CE (Cooling Experiment)-PECS. A benchmark CFD analysis of the CE-PECS (Cooling Experiment-PECS) was performed using the modified OpenFOAM solver. In this study, the physical properties such as specific heat, dynamic viscosity, and density of water were modified to provide inputs as a function of temperature. Since the modified solver was developed to analyze PECS, the mesh was made as simple as and the number of cells is also optimized to reduce the computational load. As a result of a simulation using the improved solver, the prediction of both the mass flow rate and the temperature change of the cooling water showed good accuracy compared to the case of inputting the existing constant properties. [ABSTRACT FROM AUTHOR]

• Backfitting of VVER-440/V213 reactors for design extension conditions. • Blackout scenario with depressurization of primary system. • ASTEC code for analysis of severe accidents. • Source term estimation. The contribution deals with the backfitting of Dukovany NPP for design extension conditions with emphasis on severe accidents. This NPP, located in Czech Republic, is equipped with 4 units of VVER-440/V213 reactors. During the last decades significant attention was paid to backfitting of these units. New safety systems aimed at mitigating the consequences of severe accidents have been designed and have been (are being) implemented. At first, special features of VVER-440/V213 reactor and containment equipped with bubbler condenser pressure suppression system are briefly described. The systematic process of plant backfitting for design extension condition is summarized further. Attention is paid to description of crucial severe accident management strategies that have been adopted and design and implementation of new safety systems that are necessary to fulfil these strategies. Than the requirements of Czech regulatory body (SUJB) set for practical elimination of early or large radioactive releases are discussed. Finally, the performances of severe accident management strategies are illustrated using the specific case of blackout scenario. For this purpose, an analysis was performed using ASTEC code developed by the French IRSN. The results of analysis are briefly described and discussed further. From the obtained results it follows that the high-pressure melting scenario is reliably avoided, corium is localised and stabilised inside the reactor vessel and the hydrogen in containment atmosphere is safely controlled by passive autocatalytic recombiners. Thus, stable plant state with preserved containment integrity is reached. Source term that has been estimated under these conditions indicates that the predicted releases are well below the criteria set by SUJB for early or large radioactive releases, and therefore long-term contamination of the environment is avoided. [ABSTRACT FROM AUTHOR]

• The Transient Interface Model is used to investigate MCCI ablation instabilities. • The ablation instability leads to concrete ablation in preferential directions. • The range of instability conditions is represented on a pseudo binary phase diagram. • An ablation instability may occur in the initial stage of the MCCI transient. The Transient Interface Model (TIM) (Seiler and Combeau, 2014; Seiler and Jamet, 2019) dedicated to Molten Core-Concrete Interaction simulation is used to investigate ablation instabilities. The ablation instability leads to concrete ablation in preferential directions, potentially jeopardizing the mitigation criterion based on residual concrete thickness. The ablation instability occurs when all the corium/concrete interfaces do not undergo the same ablation regime. These regimes are specific of the interface temperature; in the homogeneous Low Interface Temperature (LIT) regime, the interface temperature is close to the liquidus temperature of concrete matrix, whereas in the homogeneous High Interface Temperature (HIT) regime, the interface as well as the pool temperatures are close to the pool liquidus temperature. When these different ablation regimes are present simultaneously (non-homogeneous situation), only the surfaces under LIT regime are ablated. It is shown that ablation instabilities are favored at low power dissipation and at small test scale. Calculations suggest that the limit between HIT and LIT dominant regimes can be represented in the pseudo-binary phase diagram. The surfaces initially under HIT regime (not ablated) may return to the LIT regime (with ablation) after a time delay which depends on the extension of the surface initially under HIT regime, on the value of the mass transfer coefficient and, to some extent, on the melt to interface heat transfer. Finally, the calculation at reactor scale with siliceous concrete indicates that an ablation instability may occur in the initial stage of the MCCI transient. Such ablation instability does not induce significant preferential ablation in the reactor configuration if it is limited to the horizontal surface of the reactor pit. If it affects the lateral walls, and due to the small thickness of the corium layer, a significant ablation depth could be reached (∼1 m in a reactor pit of 6 m inner diameter). [ABSTRACT FROM AUTHOR]

• The Transient Interface Model simulates Refractory Material Dissolution by Corium. • Core-catcher thickness jeopardizes by transient corium/ceramic interaction. • Final residual thickness of ceramic liner predicted by TIM model. Core-catchers made of ceramic materials have been considered to stabilize corium pool in case of severe accidents due to the high melting temperature of ceramics. The design of such mitigating dispositions requires being able to describe the transient corium/ceramic interaction. The Transient Interface Model (TIM) applies to several severe accident situations: in vessel retention as well as corium concrete interaction under transient and steady state regimes. The paper shows that it is also able to handle situations relative to transient refractory material dissolution by a corium melt submitted to residual power dissipation. As long as the pool temperature is below its liquidus temperature, the refractory material is not ablated but could be deposited on the interface provided that the heat flux, which is extracted by conduction through the ceramic layer, is sufficient. When the pool temperature increases above its liquidus temperature, the ceramic wall can be dissolved. The transient evolution of the interface temperature depends on the pool composition and on the ablation rate. The final steady state is reached when the residual thickness of ceramic is compatible with the heat extraction by conduction through the residual wall thickness. The model can be implemented for geometries presenting different heat exchange surfaces. [ABSTRACT FROM AUTHOR]

The paper is devoted to development and testing of a technology to manufacture the ingot of the prototype corium (resulted from out-of-pile conditions) of fast reactor followed by an experimental determination of the thermophysical properties (TP) (thermal diffusivity a , specific heat capacity C p , and thermal conductivity λ ) of such corium at the room temperature (298 K). The data on the thermo-physical properties of corium (melt of structural and fuel materials of the reactor core) will be used to calculate the temperature fields in the modeling the processes of keeping corium inside the power reactor vessel under the conditions of a severe accident. [ABSTRACT FROM AUTHOR]

As a part of the integrated severe accident code development project in Korea, KAERI (Korea Atomic Energy Research Institute) has been developing a stand-alone severe accident analysis code, COMPASS (COre Meltdown Progression Accident Simulation Software), which simulates the in-vessel severe accident phenomena including the core heat up, material melting and relocation, corium behavior in the lower plenum, and vessel failure. For the purpose of COMPASS code validation, a numerical simulation was conducted for the PHEBUS FPT3 experiment, which is an integrated severe accident experiment, initiated in 1988 by IRSN. In the COMPASS code, the core and surrounding structure in a test section consist of a two-dimensional node system, and the mass and energy equations are established for the main component in the core and surrounding structures. The temperature evolution of the core and surrounding structure at various locations of the test section are pursued and compared with the experimental data as well as the numerical results of the MELCOR code. The hydrogen generation rate and final mass distribution are compared among the experimental data and the numerical results of COMPASS and MELCOR. Generally, the COMPASS and MELCOR codes predicted the temperature evolution and hydrogen generation rate during the experiment well. In addition, COMPASS predicted a slightly smaller mass relocation compared with the experimental data, whereas MELCOR has a larger mass relocation. [ABSTRACT FROM AUTHOR]

The model for U-Zr-O molten pool oxidation under non-equilibrium conditions is further extended to consideration of the vessel steel corrosion by the oxidized corium melt. The heat and mass exchange processes between various interaction layers, including U-Zr-O melt, (Zr,U)O 2− x crust, FeO corrosion layer and vessel steel wall, are considered taking into account a steep temperature gradient in the layered system. The new model is numerically realized and applied to consideration of typical temperature scenarios of METCOR tests on corrosion of vessel steel during its interaction with oxidized molten corium. [ABSTRACT FROM AUTHOR]

In case of severe accident in a light water reactor, large debris beds may form inside the core. If, in some part of the debris bed, the temperature increases up to the fuel melting point, it will lead to the formation and expansion of a molten pool which becomes more and more difficult to cool down as it grows. Therefore, it is of primary interest to determine the maximum size of a molten pool surrounded by debris which may be stabilized under water either after reflooding the vessel (as in TMI-2) or in case of a debris bed formed in a flooded reactor pit. One of the key parameters for that issue is the maximum heat flux (CHF) that may be extracted from the pool by water flowing within the debris bed. To determine that CHF, a series of tests have been conducted at IRSN. A heated copper cylinder generates a heat flux at one of its ends which simulates the boundary of the pool and is placed in contact with a debris bed made of monodisperse steel balls. Temperature measurements along the cylinder allow to calculate the heat flux transmitted to the debris bed through the cylinder tip. Several experimental parameters have been investigated: - The tilting angle of the heated surface, to take into account the shape of the molten pool. - The diameter of the balls. - The upwards steam flowrate simulating the steam coming from the part of the core located below the molten pool. - The upwards liquid flowrate simulating the natural circulation of water along the debris bed. As it is usual for CHF experiments, a significant dispersion of the measurements was observed, due to unavoidable evolutions of the contact zone between the copper surface and the balls. However, the impact of that random dispersion could be reduced by performing a large number of tests. This allows to identify clear tendencies on the CHF for the 3 studied parameters, with a rather good degree of confidence. From those tendencies, it was possible to derive a general CHF correlation with the tilting angle and the steam flowrate. When the liquid velocity increases, the CHF increases in most cases, except for some cases with low velocity. However it is more difficult to find a clear correlation with the liquid velocity or flow rate. [ABSTRACT FROM AUTHOR]

• Two PWR1000-Like input deck were developed for MELCOR 2.2 and ASTEC V2.2 codes. • The severe accident analysis is focused on containment response ex vessel phenomena and FP release. • A SBLOCA and CSBO scenarios were simulated for 10 days. • The crosswalk results are compared, discussed and the difference highlighted. The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are both used by Tractebel to perform Severe Accident (SA) analyses for the Belgian Nuclear Power Plants (NPPs). This paper contains the second part of the comparative study between these two integral software focused on the SA progression in the containment after the Vessel Failure (VF). The main phenomena characterize this phase of the accident usually called ex-vessel phase are the Molten Corium Concrete Interaction (MCCI) and ex-vessel corium/debris coolability, which determine the containment Thermal-Hydraulics (TH) response and can cause its late failure. Despite their importance from a safety point of view, chemical and physical processes behind these phenomena are not yet well understood, partly because of their extreme complexity and partly because of the limitations of current experimental capabilities in reproducing them. As result, models developed and implemented in the codes to describe the MCCI and debris coolability are extremely simplified because based on limited data availability. The benchmark study consists of a comparison of the results calculated by the ASTEC and MELCOR codes during the ex-vessel phase following two different scenarios a Complete Station Blackout (CSBO) and a Small Break Loss Of Coolant Accident (SBLOCA). For each scenario, two different Severe Accident Management Strategies (SAMS) are adopted, the first one is to ensure that the corium, once fully relocated in the cavity, has top-flooding conditions throughout the transient, while the second one is to keep the reactor cavity completely dry. Furthermore, for every case reproduced the reference and best-practice models are investigated. The results show that in case of dry cavity conditions, despite the different VF timing calculated by the two codes, the final results provide very similar indications even though some dissimilarity is observed regarding the amount of fission product (FP) released through the Containment Filtering Venting System (CFVS). In contrast, in the case of wet cavity conditions, the accident progression predicted is significantly different regarding number of venting operations, TH containment conditions, and amount of concrete ablated. The discrepancies observed, should be sought in the corium and debris coolability models, which given their parametric nature and oversimplification can lead to these results. [ABSTRACT FROM AUTHOR]

• The quenching phenomena of multi-dimensional porous debris bed were studied using experiments. • Visualization of quenching of porous debris bed is performed. • Quench front movement inside the debris bed were mapped. • Different zones of heat transfer during quenching of porous debris bed are identified. In case of severe accidents in light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue. In a flooded cavity, a porous corium debris bed is expected to develop on the bottom of the cavity pool due to the melt jet breakup and fragmentation. The debris bed needs to be quenched by water either flooding from the top or flooding from the bottom until continuous cooling is established. If it is not quenched, this mass will begin to re-melt due to decay heat and leads to Molten Corium Concreate Interaction (MCCI). Most of the studies reported have conservative assumptions of one-dimensional homogeneous particulate beds with top or bottom flooding condition of coolant. Understanding of debris beds quenching with multi-dimensional characteristics are limited. To investigate the quenching behavior of the conical debris bed by means of experiments, a test facility MCCI-mitigatiON through passive cooling Effect Test (MONET) apparatus is established. Specifically, the objective is to understand the mechanism of heat transfer and quench front movement inside the conical heated debris bed during quenching by cooling water supplied from the bottom of the bed. Two different particle bed bottom wall boundary conditions are studied; closed bottom wall and open bottom wall to understand the effect of water ingression from the bottom wall. The particulate debris bed is formed with two different types of particles; Ø 3 mm SS 304 and Ø 3 mm alumina. In addition, the effect on the quench front movement inside the particle bed has been reported by using temperature measurements together with direct visualizations. [ABSTRACT FROM AUTHOR]

This paper is a study of debris beds that can form in the core of a nuclear power plant under severe accident conditions. Such beds are formed of fragments of pellets and cladding remnants, as observed in the TMI-2 core. Many important issues are related with the morphology of those debris beds: are they coolable in case of water injection and how does molten corium progress through them if they are not coolable? The answers to those questions depend on the structure of the debris bed: porosity, number and arrangement of particles. In order to obtain relevant information, a numerical simulation of the formation of the debris bed is proposed. It relies on a granular approach of the type called “Contact Dynamics” to simulate the collapse of debris and their accumulation. Two different schemes of fuel pellet fragmentation are considered and simulations for different degrees of fragmentation of the pellets are performed. The results show that the number of axial cracks on fuel pellets strongly influences the final porosity of the debris bed. Porosities vary between 31% (less coolable cases) and 45% (similar to TMI-2 observations), with a most probable configuration around 41%. The specific surface of the bed is also evaluated. In the last part, a simple model is used to estimate the impact of the variation in geometry of the numeric debris beds on their flow properties. We show that the permeability and passability can vary respectively with a range of 30% and 15% depending on the number of fragment per pellet. The other benefits of the approach are finally discussed. Among them, the possibility to print 3D samples from the calculated images of debris beds appears as a promising perspective to perform experiments with realistic debris beds. [ABSTRACT FROM AUTHOR]

A viscoplastic model for lava is presented, and its use for corium spreading is argued with dimensionless numbers. It describes the tridimensional flow as non-Newtonian and non-isothermal. Using an adaptive finite element method based on C++ library Rheolef, simulations are ran and compared to the VEU7 corium spreading test. A sensitivity analysis on the viscoplastic parameters of corium that are not well-known is conducted. Tridimensional solution of the problem is computed from the bidimensional height-averaged solution. • Molten nuclear fuel is simulated using Herschel–Bulkley viscoplastic model. • Encouraging results are found in comparison to prototypic experiment. • Sensitivity analysis on viscoplastic parameters is performed. • 2D simulation solution yields 3D results through reconstitution. [ABSTRACT FROM AUTHOR]

In the in-vessel corium retention studies conducted on the Rasplav-3 test facility within the ISTC METCOR-P project and OECD MASCA program, experiments were made to investigate transient processes taking place during the oxidation of prototypic molten corium. Qualitative and quantitative data have been produced on the sensitivity of melt oxidation rate to the type of oxidant, melt composition, molten pool surface characteristics. The oxidation rate is a governing factor for additional heat generation and hydrogen release; also for the time of secondary inversion of oxidic and metallic layers of corium molten pool. [ABSTRACT FROM AUTHOR]

• Highly melt spreading is analyzed with the moving particle hydrodynamics (MPH). • No-slip/slip boundary conditions are developed for reproducing the spreading behavior under active ceramics and concrete substrates. • VULCANO VE-U9 spreading experiment is utilized for the validation of MPH. • Crawling and gliding spreading motions are observed under ceramics and concrete, respectively. Understanding the spreading of molten core materials (corium) is vital for achieving post-accident heat removal and maintaining the integrity of the reactor containment during a severe nuclear accident. The spreading of highly viscous corium over sacrificial concrete and ablation-resistant ceramic substrates, consistent with recent VE-U9 experiments at the VULCANO facility, is investigated using a Lagrangian moving particle hydrodynamics (MPH) method. The spreading dynamics are coupled with a heat transfer model to account for the increase in melt viscosity due to solidification. Three alternative boundary conditions are considered at the melt-substrate interface to mimic the influence of low viscosity concrete ablation products on the spreading dynamics. Simulation of spreading under no-slip conditions closely resembled the spreading behavior observed over the inert ceramic substrate, advancing with a crawling motion with a steep profile in the proximity of the melt front. Spreading terminated due to a combination of thermal radiation from the free surface and the heat transfer to the substrate, leading to the formation of a continuous crust. Introduction of slip boundary models, to mimic lubrication of the interface by a film of low-viscosity concrete decomposition products, enabled the prediction of the gliding flow observed over the sacrificial concrete substrate, characterized by a shallow melt topology near the spreading front. The simulation results demonstrate excellent potential for Lagrangian models such as MPH to predict complex spreading dynamics in the presence of both melt solidification and lubrication at the substrate. [ABSTRACT FROM AUTHOR]

During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. This model was then used to analyze the formation of debris beds at the reactor scale. A sensitivity analysis was carried out based on key accident parameters, including the quantity of corium melt, cavity flooding level, volumetric decay heat rate, and the size of the melt jet. [ABSTRACT FROM AUTHOR]

• Heat absorption from carbonate mineral decomposition can be used to quickly solidify molten corium and mitigate an accident. • This injectable granular material concept can be integrated into an existing nuclear power plant accident management system. • MELCOR was used to demonstrate the effectiveness of the novel injectable mitigation system in a simulated reactor accident. • The MELCOR analysis demonstrates that using this carbonate injection system can delay the destructive hydrogen explosion. To mitigate adverse effects from molten corium following a reactor pressure vessel failure (RPVF), some new reactor designs employ a core catcher and a sacrificial material (SM), such as ceramic or concrete, to stabilize the molten corium and avoid containment breach. Existing reactors cannot easily be modified to include these SMs but could be modified to allow injectable cooling materials. Current reactor designs are limited to using water to stabilize the corium, but this can create other issues such as reaction of water with the concrete forming hydrogen gas. The novel SM proposed here is a granular carbonate mineral that can be used in existing light water reactor plants. The granular carbonate will decompose when exposed to heat, inducing an endothermic reaction to quickly solidify the corium in place and producing a mineral oxide and carbon dioxide. Corium spreading is a complex process strongly influenced by coupled chemical reactions, including decay heat from the corium, phase change, and reactions between the concrete containment and available water. A recently completed Sandia National Laboratories laboratory directed research and development (LDRD) project focused on two research areas: experiments to demonstrate the feasibility of the novel SM concept, and modeling activities to determine the potential applications of the concept to actual nuclear plants. Small-scale experiments using lead oxide (PbO) as a surrogate for molten corium demonstrate that the reaction of the SM with molten PbO results in a fast solidification of the melt due to the endothermic carbonate decomposition reaction and the formation of open pore structures in the solidified PbO from CO 2 released during the decomposition. A simplified carbonate decomposition model was developed to predict thermal decomposition of carbonate mineral in contact with corium. This model was incorporated into MELCOR, a severe accident nuclear reactor code. A full-plant MELCOR simulation suggests that by the introduction of SM to the reactor cavity prior to RPVF ex-vessel accident progression, e.g., core-concrete interaction and core spreading on the containment floor, could be delayed by at least 15 h; this may be enough for additional accident management to be implemented to alleviate the situation. [ABSTRACT FROM AUTHOR]

• The prototypic materials have been conducted to investigate the stratification behavior of the steady molten pool. • The synergistic effect of M Fe /M ∑ and Cn on redistribution of the corium components. • The reason for layer inversion of the molten pool occurred is studied. The prototypic materials have been used to investigate the stratification behavior of the steady molten pool at 3100 K in the context of the IVR condition. Induction heating in a cold crucible was used for melting of the 2 kg-scale UO 2 -ZrO 2 -Zr-Fe. A post-test examination of SEM, EDS, and XRD has been performed to investigate the metallic and oxide layers for all ingots. The effects of M Fe /M ∑ (initial iron to corium ratio) and Cn (initial degree of zirconium oxidation) on the configuration of the molten pool were studied. The results indicate that structure of the molten pools consists of the top light metallic layer and oxide layer. Nevertheless, a layer inversion may occur when the density of the metallic layer increased. The synergistic effect of M Fe /M ∑ and Cn on redistribution of the corium components leads to a significant increase of the uranium concentration for the metallic phase, which can explain why the layer inversion of the molten pool occurred. [ABSTRACT FROM AUTHOR]

In the context of light water reactor severe accidents analysis, this paper is focused on one key parameter of in-vessel corium phenomenology: the immiscible phases stratification and its impact on the heat flux distribution at the corium pool lateral boundary with the so-called focusing effect related to a “thin” top metal phase and the potential vessel failure at that point. More particularly, based on the limited knowledge of the stratification transient phenomenon derived from the MASCA-RCW experiment, a basic model is proposed that can be used for corium in lower head sensitivity analyses. It has been implemented in the PROCOR platform developed at CEA Cadarache. A short parametric study on a simple hypothetical transient is presented in order to highlight the different focusing effect “modes” that can be encountered based on this in-vessel corium pool model. An early mode may occur during the formation of the top metal layer while two other modes may appear later during the thinning of this top metal layer because of thermochemically induced mass transfers. Some associated relevant parameters (model or scenario-dependent) and modelling issues are mentioned and illustrated with some results of a Monte-Carlo based sensitivity calculation on the transient behaviour of the corium in the lower head of a 1650 MWe GenIII PWR. Within the limiting modelling hypotheses, the thermal modelling of the steel layer for small (centimetre) heights and the mass diffusivity (limited in this case to the uranium diffusivity in the oxidic layer) are main sensitive parameters. [ABSTRACT FROM AUTHOR]

The present work addresses the question of corium–ceramic interaction in a core catcher during a core-melt accident in a nuclear power plant. It provides an original insight into transient aspects concerning dissolution of refractory material by a volumetrically heated pool. An experiment with simulant material (LIVECERAM) is presented. Test results clearly show that dissolution of solid refractory material can occur in a non-eutectic melt at a temperature which is lower than the melting temperature of the refractory material. During the dissolution transient, the interface temperature rises above the liquidus temperature, corresponding to the instantaneous average composition of the melt pool. With constant power dissipation in the melt and external cooling of the core-catcher, a final steady-state situation is reached. Dissolution stops when the heat flux (delivered by the melt to the refractory) can be removed by conduction through the residual thickness of the ceramic, with T interface = T liquidus (calculated for the average composition of the final liquid pool). The final steady state corresponds to a uniform pool composition and uniform interface temperature distribution. Convection in the pool is governed by natural thermal convection and the heat flux distribution is therefore similar to what would be obtained for a single component pool. An interpretation of the experiment with two model-based approaches (0D and 1D) is presented. The mass transfer kinetics between the interface and the bulk is controlled by a diffusion sublayer within the boundary layer. During the dissolution transient, the liquid composition at the interface is concentrated in the refractory species. During the transient, the interface temperature is equal to the liquidus temperature corresponding to the local and instantaneous composition of the liquid at the interface. Regarding the design of a protective layer made of refractory materials, we can answer the question of how much ceramic can be dissolved and its impact on melt temperature evolution during the dissolution process. It also impacts on subsequent corium solidification since the additional mass of dissolved ceramic leads to increased volume of the molten material, significantly increasing the time required for complete solidification. For the long term, ceramic material does not offer better confinement than a crust made of solidified corium. This work served as support to a generalisation of the model of transient evolution of interface temperature in various severe accident situations (Seiler and Combeau, 2014). [ABSTRACT FROM AUTHOR]

During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO 2+ x –ZrO 2 –FeO y corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 °C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR). [ABSTRACT FROM AUTHOR]

• LES were performed for thermal behavior in sloshing corium pools with high Re number. • The transient evolutions of sloshing corium pool temperature and heat flux were presented. • The correlation between heat transfer and Re number was exhibited in a sloshing corium pool. The convection flow and heat transfer actuated by internal decay heat and the sloshing of vessels are the primary problems in the study of thermal behaviors in volumetrically heated reactor corium pools under ocean conditions. The wall-modeled large eddy simulation (WMLES) method, volume of fluid model, and phase-change model combined with dynamic mesh technology, were used to develop a three-dimensional flow and heat transfer process simulation model of the sloshing corium pool. In addition, the heat conduction behavior in the crust and vessel wall, as well as the pressure-vessel-wall integrity evaluation was considered. The predicted internal heat transfer characteristics (the temperature distribution, the wall heat flux, and the shell thickness) in the corium pool and the evolution feature of the free surface height near the wall were in good agreement with the LIVE-L4 test results and sloshing experimental data, respectively. Furthermore, the coupled heat transfer phenomenon of the melt and vessel was simulated. The simulation results indicated that the time required to reach stable heat transfer was shortened, the structure of thermal stratification was transformed, and the heat flux along the inner surface of the vessel wall of the shell was improved at the resonance state. The comparisons of local Nusselt (Nu) number at the near-wall between static and sloshing corium pool conditions showed that heat transfer characteristics were significantly influenced by sloshing (T = 2.5 s, A = 0.07 rad) motion, and the Nu number of forced convection fraction had a strong correlation with the local Reynolds number in sloshing corium pools. Conclusively, the transient heat flux distribution of the sloshing melting pool differed from that of the static corium pool, increasing the molten range, and molten peak depth of the pressure-vessel-wall. [ABSTRACT FROM AUTHOR]

• Chemical interactions between (U-Zr-O) molten corium and molten steel are performed to study the impact of temperature difference and presence of crust between stratified phases. • The existence of a crust between stratified oxide and metal phases does not prevent mass transfers and a possible inversion of stratification due to changes in metal composition. • Under oxidizing atmosphere, two main processes may limit the rate of oxidation: the presence of a top crust or the oxidant starvation. • Chemical interactions between (U-Zr-O) porous debris bed and molten steel are performed to study the impact of temperature and oxidizing atmosphere on the progression of molten metal. • Under non-oxidizing atmosphere, an eutectic metallic mixture is formed around 900 °C and progresses easily through the oxide particle bed. Under oxidizing atmosphere, metal wettability is reduced and progression is stopped. One of the strategies to mitigate the consequences of a severe accident is to stop the progression of molten materials (corium) within the vessel by external cooling of the vessel. This strategy is called in-vessel retention (IVR). The safety demonstration of IVR requires the accurate evaluation of the thermal load imposed on the internal surface of the vessel by the corium debris. During the progression of the accident, solid debris may be generated initially in the lower plenum because of fragmentation in water and, after melting of those debris, a (possibly stratified) molten pool would be formed. The consequences of the thermal and chemical interactions of the corium with the vessel are difficult to predict because there are still uncertainties on the application of existing results to models for reactor cases. These include scaling effects of course, but also the influence of transient processes occurring simultaneously: oxidation, dissolution, melt progression and/or stratification. In order to be able to predict the heat fluxes through the RPV wall, it is necessary to establish an overall view of the possible transient evolutions of corium and determine which are the most probable 'steady-state' corium configurations after debris melting. The CORDEB program studies corium phenomena taking place at different stages of molten pool formation in the reactor vessel. It may be considered as an investigation of corium configurations that were not considered in previous programs such as the OECD RASPLAV and MASCA programs. All experiments within the program are carried out in the Rasplav tests facility of NITI, using the technology of induction melting in a cold crucible. This program is co-funded by IRSN, EDF, AREVA and CEA in the frame of the CPSIN agreement. This paper presents the main conclusions drawn from CORDEB tests, both for phenomenological knowledge and for transposition to reactor situations. Quantitative elements are provided for a clearer picture of kinetic effects. CORDEB tests have shown the importance of chemical interactions between molten steel and solid sub-oxidized (U-Zr- O) corium which lead to penetration of steel inside the solid oxide (crust or debris) and to the partial dissolution of the solid. As a consequence, the crust surrounding a molten pool cannot be considered as a barrier to mass transfer. CORDEB tests also made it possible to assess the oxidation kinetics of a molten pool: in the presence of surface crust, this is a rather slow process and it does not have a significant effect on the deterioration of accident scenario. The gaseous atmosphere above the pool plays an important role in several processes, in particular in the evaporation of molten steel. [ABSTRACT FROM AUTHOR]

• MSPREAD was validated based on the three dry spreading experiments. • Influences of generated gases on the friction force and heat transfer were modeled. • The cell size convergence was confirmed for the highly viscous melt spreading. • The gas enhanced heat removal at the upper surface decelerated spreading. • The weir anchoring model was developed to interpret the anisotropic spreading. MSPREAD has been developed to evaluate the multi-dimensional melt spread on both dry and wet floors while considering a number of influential mechanisms. Part-1 of this two-part series described models, numerical methods and verification results. Part-2 focused on the validation of the models related to dry spreading behaviors. The special models necessary for in-depth studies of the experimental results were introduced, then the target phenomena included in the dry spreading were related to the three dry spreading experiments employing molten oxide mixtures (KATS-12 and ECOKATS-1) and prototypic UO 2 -ZrO 2 corium (VULCANO-VEU7). To interpret data with significant gas generation, the bubble suspension viscosity model learned from the rheology field was implemented. This enabled to represent the wide-ranging viscosity change by the Ramacciotti model based on the solid fraction coefficient = 6.3 common to all three experiments. In KATS-12, the predicted spread distance was not sensitive to the cell size in the range of 10 to 50 mm. In VULCANO-VEU7, the highly viscous corium spread a short distance on the ceramic and concrete sectors and showed abrupt stop particularly on the concrete sector. These observations were explained by a gas sparging enhancement of the melt upper surface heat transfer that is in consistent with the heat flux reconstructed based on the measured temperature histories. In ECOKATS-1, the weir anchoring model was proposed to simulate the anisotropic spreading on the 2D plane. Although the present prototype model requires a user to specify locations and timings of the weir formation and break, the successful interpretation of complex spreading behaviors encouraged the authors to improve its predictability by mechanically modeling accumulation of floating crusts at the melt fronts, the transformation into the weir that can obstruct the melt flow, and the break due to ablation caused by the melt convection heat transfer. [ABSTRACT FROM AUTHOR]

• Two PWR1000-Like input deck were developed for MELCOR 2.2 and ASTEC V2.2 codes. • The severe accident analysis is focused on the RCS thermal–hydraulic and in-vessel phenomena. • SBLOCA and CSBO scenarios up to the lower head failure were simulated. • The crosswalk results are compared, discussed and differences highlighted. The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are used in Tractebel to perform source term assessments and plant Thermal Hydraulic (TH) analyses in case of Severe Accident (SA) for the Belgian Nuclear Power Plants (NPPs). These system codes were developed through intense and extensive validation activities which were mainly based on the verification of the capability of implemented physical models to reproduce the experimental results of Separate Effect Tests (SETs) and Coupled Effect Tests (CETs). However, the SETs, and in less extension the CETs, do not take into account interactions and possible synergistic or antagonistic effects which can arise among the different physical and chemical processes occurring during a SA. Thus, if a code is able to reproduce correctly most of SETs or CETs, this does not mean that it can ensure the same quality of results for the more realistic and complex Integral Effect Tests (IETs). Nevertheless, some contraindications exist also for the IETs, because of the scaling factor effect which, being difficult to quantify, could lead to development of models replicating correctly the experimental data but that are not totally reliable for reactor scale applications. In addition, waiting for the dismantling of Fukushima Dai-ichi units 1–2 and 3, which can provide new information in the field of the core degradation phenomena, in the present only data on in-vessel core melt progression for the Three Mile Island, Unit 2 (TMI-2) accident are available. Therefore, apart from simulating the TMI-2 accident scenario, there is no other way to assess models of in-vessel phenomena (H 2 production, corium pool/ debris formation and relocation, etc) implemented in the codes on reactor scale. As a consequence, there are still uncertainties in the capability of these integral tools to predict the progression of postulated severe accident transients in a NPP, notably core degradation and ex-vessel phenomena. In order to investigate the differences among the physical models adopted in MELCOR 2.2 and ASTEC V2.2 codes, and evaluate the impact that they could have on the assessment of Severe Accident Management Strategies, a crosswalk study was carried-out. This paper describes the first phase of this work, which consists of a detailed comparison between these two SA tools on a well-defined plant (PWR1000-Like) with prescribed boundary and initial conditions. Furthermore, to avoid additional and unwanted sources of discrepancies between code calculations, the two reactor models were developed in parallel following the recommendations suggested by developers for core and Reactor Coolant System (RCS) nodalization. The comparative study is focused on the TH of the RCS and in-vessel phenomena, for two different accident scenarios (SBLOCA and CSBO) up to the moment of lower-head failure. Ex-vessel behaviour will be examined in the second phase of this study. The crosswalk results obtained have shown some differences and similarities on reproducing TH in the RCS and In-vessel phenomena. Especially for these latter, the discrepancies predominate mainly due to how the codes treat corium behaviour, while thanks to the harmonisation of the initial steady-state and boundary conditions, the discrepancies on the prediction of the RCS behaviour have been minimized, at least before significant core degradation takes place. [ABSTRACT FROM AUTHOR]

• Description of PLINIUS prototypic corium experimental platform. • Prototypic corium melting/heating technologies: thermite – induction. • New water flooding facility: MERELAVA. • Small scale corium experimental setups in VITI. • Future STROMBOLI facility for in-vessel corium stratification R&D. In the recent years, CEA Cadarache PLINIUS prototypic corium platform has been upgraded: more powerful induction generator for e.g. MCCI experiments, development of thermitic melting, e.g. for new corium spreading tests in VULCANO, new MERELAVA facility for corium flooding, new VITI setups for thermophysical property measurements, KROTOS FCI test with full depth X-ray visualization of water pool. Main features of these facilities are described. Designs for further improvements such as the new FUJISAN and VITAE setups for characterizations of the aerosol released during prototypic fuel debris cutting, as well as the design of STROMBOLI test section for studying Lower Head pool stratification inversion evolution are also presented. [ABSTRACT FROM AUTHOR]

• Five scoping tests were carried out on the DEFOR-M facility using a molten Tin. • The visualization of jet fragmentation and debris formation was obtained. • The characteristics of debris bed were analyzed based on laser scanning system. • The morphology of metallic particles was also very different from that of the oxidic particles in the former tests. • Different scales (dm, cm and mm) of debris particles were found. During a hypothetical severe accident in a light water reactor, the core and internal structures melt down and the molten materials (corium) may fall into a water pool, forming a debris bed from so-called fuel coolant interactions (FCI), whose characteristics are important to corium coolability. The present study is concerned with the characterization of an ex-vessel debris bed forming form the FCI of a metal rich corium jet falling into a water pool in the reactor cavity. Five scoping tests were carried out on the DEFOR-M test facility using metallic Tin melt at different superheats. The molten tin was employed as the simulant of metal (Zr/Fe) rich corium melt. The processes of melt jet fragmentation and debris formation was recorded by high-speed cameras. The contour and volume of the resulting debris bed were measured by a three-dimensional laser scanning system, and the debris particles were sieved for their size distribution. The experimental results revealed the effects of melt superheat and water subcooling on the characteristics of a debris bed, including the bed's configuration and porosity, the particles' morphology and size distribution. A preliminary comparison of debris bed characteristics between metallic melt and oxidic melt was also provided. [ABSTRACT FROM AUTHOR]

In the frame of light water reactors' severe accidents (SA), the success of a mitigation based on the core melt retention in the vessel lower head largely depends on the corium pool phases stratification. Such a melt is composed of molten partially oxidised materials relocated from the core and molten steel mainly coming from the ablation of lower head internal structures and the vessel wall during the transient. Because of the miscibility gap in the U-O-Zr-steel thermodynamic system at liquid state, such a pool can go through various stratified states during a transient with a heavy metal phase at the bottom or a light metal one at the top. Actually, there is a significant knowledge gap related to steel migration to the pool that influences the stratification transient and can lead in some cases to a large uncertainty on the heat flux that the pool imposes on the vessel lower head. In particular, when the ablated vessel wall is considered, the molten steel relocation, either as a continuous lighter phase above the pool or as dispersed phase in the oxide pool, depends on the behaviour of the refractory crust at the pool/vessel interface (that can undergo dissolution or mechanical damage). In the case of a dispersed metal phase in the oxide pool, determining the subsequent state of the lower head pool requires the evaluation of the thermochemical interaction between both materials and their possible hydrodynamic separation. To the best of our knowledge, no model described these coupled phenomena in integral SA codes. In this paper, an integral model is proposed to better assess the phenomena at stake in such a case and clarify the need (or not) to enhance stratified pool models in SA codes. To do so, this integral model was constructed on "first-order" hypotheses; in particular, thermal equilibrium is assumed and a simplified representation of the U-O-Zr-steel thermodynamic system is considered. The model has been implemented in the PROCOR software platform and used for a parametric analysis that provides insightful order of magnitudes and trends. The main parameters are the droplet size, the corium chemical composition and the model closures associated with interfacial mass transfer. The results reported here show that mass transfer, "competing" with droplet hydrodynamics, plays an important role for droplets of millimetric size that can, depending on the pool composition, relocate under the oxide under the form of a continous heavy metal layer. For larger droplets, it is a light metal layer above the pool that can be expected. The prolonged existence of a dispersed metal phase that would remain mixed in the oxide phase seems to be only possible for a very narrow range of pool composition near the stratification inversion threshold. [ABSTRACT FROM AUTHOR]

• Water Ingression data of the SSWICS experiments are re-interpreted. • The thermal-hydraulic is analyzed with both an analytical model and a multi-phase CFD model. • The 2D analysis suggests important border effects. • A new semi-empirical model is proposed for the estimation of the crust's permeability. In the course of severe accident in a nuclear power plant, the corium mixture may flow down in the reactor pit and start a thermal attack of the basemat concrete. A simple way to terminate the melt progression may be to add water on top of it. Among the various physical processes that may participate in the quenching, water may penetrate into the solidified corium through cracks generated by the thermal stress during the solidification. The present paper aims at providing a clarification of the process though an analysis of the few dedicated experimental data available using proto-typical corium, namely the SSWICS 1-7 experiments. At first, a general reminder and analysis of the data is given. Then, the thermal-hydraulic aspects are investigated through the construction of a dedicated heat flux correlation and its use via the experimentally measured post-test permeabilities. The analysis is supported by 1D and 2D evaluations with the multiphase flow code MC3D-PREMIX, slightly modified for the purpose. It is concluded that the heat flux drastically decreases with the amount of added concrete material. Furthermore, 2D border effects are investigated and their importance highlighted in view of the experimental results analysed. These effects should explain the absence of observation of water ingression heat flux in the cases with large concrete amounts. Following, the paper proposes a model for the created permeability. Due to the complexity of the process and to the large uncertainties of the needed material properties in the considered situation with very high temperatures, a semi-empirical model is derived. Lastly, the model is adapted to the situation with internal decay heat, although no open experimental data is available to precisely support any model. A complete modeling is out of the scope of the paper, hence the focus is to provide hints for a first analysis of the impact of decay heat on its coolability. [ABSTRACT FROM AUTHOR]

• The maximum bubble pressure technique has been used to measure the density and the surface tension of molten materials up 2550 °C. • Original results have been obtained considering the corium C32 at the melting point 2470 °C. • These results have been compared to former RASPLAV test results showing good agreement. The paper proposes a description of the maximum bubble pressure (MBP) technique that has been used to measure the density and the surface tension of high temperature molten materials (up to 2550 °C). The technique has been implemented in the VITI facility in order to characterize corium thermophysical properties. The facility is presented together with its instrumentation and uncertainty assessment. In a first step, the VITI-MBP device has been qualified by measuring reference material surface tension at room temperature (ultrapure water) and at high temperatures (Pure Alumina, 99.99%). Then, original results of corium density and surface tension have been obtained considering the corium C32 (32% oxidized corium) at the melting point 2470 °C. These results have been compared to former RASPLAV test results showing good agreement. [ABSTRACT FROM AUTHOR]

• A novel timestep control has been developed for the MPS method based on particle number density. • This new technique can efficiently prevent numerical instability and reduce the calculation cost. • Crust formation and breach condition in ECOKATS-V1 have been investigated in detail. • Solidification Viscosity Threshold (SVT) and Crust Fracture Stress Threshold (CFST) have an important influence on crust formation and breach condition. Understanding the ex-vessel spreading behavior of molten core materials (corium) on the reactor containment floor is of great importance for designing, managing, and evaluating severe accidents of light water reactors. Among some simulated ex-vessel corium spreading experiments, temporal termination and restart of the flow (stop-and-go) were highlighted in the ECOKATS-V1 experiment, in which the simulant oxide melt (consisted of 41% of A l 2 O 3 , 24% of S i O 2 , 19% of C a O , and 16% of F e O) was released into a 1-D channel, and the spreading behavior seemed to have temporarily stopped and restarted following crust formation and breach at the spreading leading edge. The preceding study developed crust formation and breach models for the Lagrangian-based Moving Particle Semi-Implicit (MPS) method to analyze such spreading. However, the applicability of the models was limited due to numerical instability. In this study, the timestep control of the MPS method is revised by newly introducing particle number density control, which can efficiently reduce timestep to prevent instability when the crust is formed or breached in low viscous flow. The new method is successfully applied to ECOKATS-V1 spreading with different crust formations and breach conditions. Namely, with different Solidification Viscosity Threshold (SVT) and the Crust Fracture Stress Threshold (CFST) of the melt. The former determines the threshold viscosity above which the fluid is regarded as solid (crust). The latter determines the mechanical strength of the crust at the leading edge of the spreading front. There remain issues for future studies to further develop the method for more quantitative discussions. [ABSTRACT FROM AUTHOR]

• A pseudo-binary CFD model is proposed to simulate a corium pool. • Metal phase droplets during the stratification transitory are captured by the model. • Mobility, viscosity and surface tension are shown to influence the metal phase flow. During a severe accident in a nuclear reactor, the melting of the core may lead to the formation of a multiphase liquid pool (corium) in the vessel lower head. The heat transfer at the boundary with the vessel is affected by diffusive and convective mass fluxes. In particular, the development of Rayleigh–Taylor instabilities influence the thickness of the top metallic layer and therefore the "focusing effect" of the heat flux, which is the main risk for the vessel integrity. We use a Cahn–Hilliard pseudo-binary model to describe the uranium/oxygen/zirconium/iron mixture. The diffusion and the convection are governed by the Cahn–Hilliard equation and the Navier–Stokes equations under the Boussinesq approximation. In this work, the model is isothermal and the buoyancy force is only due to the gradient of chemical composition. The model is solved in three dimensions with a pseudo-spectral code. The initial configuration consists of a light layer of iron-rich fluid above a heavy layer of uranium/oxygen/zirconium mixture. A thin layer of heavier metallic phase lays at the interface and eventually triggers a Rayleigh–Taylor instability. The metallic phase forms a plume which falls downward and then breaks up into droplets due to the Rayleigh-Plateau instability. The phenomenon is alimented by diffusion which generates the heavy metallic phase at the interface. The droplet formation observed in an experiment of corium stratification transient from the literature is qualitatively captured. The mobility, the viscosity and the surface tension are shown to have an influence on the mass transfer. [ABSTRACT FROM AUTHOR]

• Determination of a new experimental correlation for assessing the ablation rate in the jet axis in the first ablation regime. • The succession of two ablation regimes is reaffirmed (the initial constant ablation regime and the pooleffect). • Global phenomenological analysis of cavity shape evolution with time. • Measurements of cavity depth to jet's diameter ratio at onset of pool-effect (such measurements were never done before to the best of our knowledge). In the scope of safety studies for future sodium cooled fast reactors (SFR), the use of discharge tubes and in-vessel core-catcher are foreseen for severe accident mitigation. During relocation of corium onto the core-catcher, complex thermal hydraulics phenomena can occur. This work focuses on the ablation of the core-catcher by the jet of corium. For that purpose, as experiments using corium and Na are difficult to achieve, we performed experiments using simulant materials: water jet on transparent ice. Transparent ice allows for ablation cavity real-time recording. Two main regimes were identified: a film flowing regime and a pool regime. We also studied the influence of temperature (30, 50, 70 °C) and velocity (1 , 2.5 , 5 , 7.5 , 10 m / s) of the jet. Comparison with existing data have been done and finally, a scaling law in the first regime, for which the ablation rate is constant, has been found. [ABSTRACT FROM AUTHOR]

• A model is proposed to describe the dissolution of a solid oxide crust in contact with molten iron. • Results show a good quantitative comparison with experimental data (CORDEB tests). • It is found that liquid channels are formed within the crust as a result of dissolution. • Under a pressure gradient, the metal percolates through the crust, after an incubation time. This paper deals with the study of dissolution phenomena in corium crust, in a severe accident. In case of an In-Vessel Retention (IVR) strategy, it is expected that the corium pool will be surrounded by an oxide crust, which will be in contact with molten steel from the top, and along the sides. CORDEB experiments have shown that this crust becomes permeable, having an impact on the thickness of molten steel layer above it. A thin molten steel layer atop crust may lead to an excessive heat flux to the Reactor Pressure Vessel (RPV), resulting in a rupture or melt-through. Corium being a mixture of (U , Zr , O) atoms, undergoes a quaternary interaction with Fe atoms. Such an interaction plays an important role in the stability (or not) of the crust. A thermochemical study of this quaternary system shows that it can be reduced to a pseudo-binary in each phase: (U , Zr) + O in the oxide phase and (U , Zr) + Fe in the metal phase. Therefore, in this paper, dissolution of a binary two-phase porous region by liquid, is studied by means of an up-scaled mathematical model. The model is derived by volume averaging transport equations – Mass, Momentum and Specie – over a Representative Elementary Volume (REV). Final system of Partial Differential Equations (PDEs) is closed by deriving empirical relations for the effective diffusivities and for the mass transfer coefficients. Model is solved numerically to study the progress of dissolution, facilitating, molten steel filtration through the two-phase region. A qualitative comparison has also been made with CORDEB experiment. Results without convection show that the model is able to demonstrate the creation of liquid metal channels similar to the ones observed in the CORDEB findings. Further, the result with pressure gradient driven convection shows the possibility of complete crust dissolution, which is due to the convection enhanced dissolution in the corium crust. [ABSTRACT FROM AUTHOR]

• Severe accident mitigation technology was developed for a small integral reactor of SMART. • Severe accident mitigation strategy was evaluated using MELCOR computer code. • The dominant sequence of SBO was selected from PSA results for SMART. • The containment pressure was analyzed for containment integrity maintenance. • The hydrogen behavior was analyzed to check the hydrogen safety criteria. Severe accident mitigation technology was developed and evaluated for a small integral reactor of SMART. The containment pressure and hydrogen behavior were analyzed using MELCOR computer code (Sandia National Laboratory, 2018) in the SBO (Station Black Out) sequence, which was selected from PSA (Probabilistic Safety Assessment) results for the SMART. The severe accident mitigation technology to improve the SMART safety include reactor vessel depressurization using the ADS (Automatic Depressurization System) to prevent DCH (Direct Containment Heating) in case of a reactor vessel failure, a reactor cavity flooding using the CFS (Cavity Flooding System) with the IRWST (In-containment Refueling Water Storage Tank) for the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) to prevent the reactor vessel failure, and a hydrogen control system of PARs (Passive Autocatalytic Recombiners) to remove hazards from hydrogen combustion considering the amount of hydrogen to be generated by 100% fuel cladding oxidation. The MELCOR results showed that the containment pressures during the SBO sequence was below the design pressure, which meant that the containment integrity is maintained during a severe accident in the SMART. The hydrogen mole fraction of the containment was much lower than the hydrogen safety criteria of 10 vol%, which meant that the possibility of the hydrogen burn in the containment was negligible. For this reason, it could be concluded that the hydrogen was controlled by the hydrogen moving path and the hydrogen control system of PARs in the SBO sequence of the SMART. [ABSTRACT FROM AUTHOR]

• Pioneer work with large-scale naturally separated simulant material at high temperature. • The turbulent flow pattern in the upper layer is influenced by upper boundary cooling. • The interlayer crust behavior reacts strongly on upper boundary condition. • Heat flux focusing in upper layer only exists under week upper cooling boundary. • The influence of Pr number in the upper layer heat transfer is discussed. • The influence of the 2D geometry on the heat flux focusing effect is discussed. In-Vessel Melt Retention (IVR) is credited as a promising Severe Accident Management strategy to stabilize the molten corium during severe accident scenarios. This strategy is not only adopted in VVER 440 or AP600 designs, but is also included in higher power reactors around 1000 MWe, such as AP1000 and Chinese CPR 1000. One of the current research priorities bound with large uncertainty is the heat flux at the cooled wall in the metallic layer atop of a ceramic pool, as heat flux focusing effect could appear at this location. The generation of experimental data with naturally separated melt layers is essential to improve the understanding of the heat transfer between the melt layers and on their boundaries. Under the framework of EU H2020 IVMR project, two large-scale test series with simulant materials were conducted in LIVE2D facility at KIT. Rigid top cooling and hot atmosphere on the upper layer were realized in a two-layer melt pool with different upper layer thicknesses. Important global heat transfer phenomena have been obtained, including the effect of the top cooling boundary, the behaviour of the interlayer crust and the impact of upper layer thickness. Finally, the influence of Pr number and geometry are discussed. [ABSTRACT FROM AUTHOR]

• FLEX strategies assessed for a Large Break LOCA in a three loop PWR-W with MELCOR. • Analyses on recirculation failure time, onset and mass flow from portable pumps. • No significant differences found on mass flow rate injection. • Methodology proposal for SAMG support. After the Fukushima-Daiichi accident, different methods, strategies and guidelines were developed to improve the Severe Accident Management Guidelines (SAMG). One of these additions, named the FLEX strategies, relies on portable equipment to obtain power or water to restore key safety functions in the reactor. In the present study, the FLEX strategies are studied under an in-vessel molten core retention scenario; these actions will try to avoid the reactor pressure vessel failure when the core is in a degraded or molten state. In order to analyze this sequence, a large break LOCA is simulated in a three loops PWR-W using the MELCOR 2.2 code. During the simulated sequences different configurations are assessed; the water injection by the ECCS during the recirculation phase is supposed to fail at different times, and the water injection with portable equipment subsequently occurs also at different times after the ECCS failure. Additionally, different water injection rates are also analyzed to compare different portable pumps capabilities. The obtained results agree with previous literature, i.e. the reactor vessel failure could be precluded if the water injection with portable equipment starts before there is a large amount of corium in the lower head. No significant differences are found between different water injection rates (20-60 kg/s) to prevent vessel failure, but it does have an impact on the containment pressure and hydrogen generated. [ABSTRACT FROM AUTHOR]

• CFD was used for modelling of corium spreading. • End fitting geometry of PHWR was simulated. • Melting and solidification effects were considered. • Empirical and mechanistic viscosity formulations were compared. • Corium rheology affected solidification process. This study investigates the numerical simulation of corium relocation during a hypothetical accident, focussing on the modelling of core melt, its solidification and redistribution. This paper presents results of a high-fidelity computational fluid dynamics (CFD) simulation of an ex-vessel corium flow through a horizontal end fitting geometry, a pressurized heavy water reactor (PHWR) relevant configuration. The objective of this work is to simulate the corium progression within the end fitting of a PHWR to assess the suitability of two viscosity models using a CFD code; Siemens STAR-CCM+. This first-of-a-kind, exploratory study demonstrates the potential of the existing CFD models for modeling corium spreading within the end fitting geometry. Mixed convection turbulent corium flow in the end-fitting has been modeled in the three-dimensional CFD simulations. Transient simulations using the volume-of-fluid approach and melting-solidification models were performed to predict the corium melt spread within the end fitting. Two suitable temperature-dependant models for corium viscosity were tested, one being empirical viscosity model by Ramacciotti and the other mechanistic viscosity model available in STAR-CCM+. The unique flow features such as the Rayleigh-Bénard instability could be observed using the Ramacciotti empirical viscosity modelling approach. A qualitative assessment of the results was undertaken, showing that the CFD model has predicted the corium spreading behaviour that is consistent with the expectations based on engineering judgement and experience. This kind of simulations are expected to serve as a guidance in planning lab-scale experiments and aid in CFD-informed modeling improvements for integral (or system) codes going forward. [ABSTRACT FROM AUTHOR]

• Thermal & structural analyses are conducted for a postulated severe accident in PHWR. • High temperature tensile and creep properties of Calandria material are generated and used. • In-vessel-retention duration is evaluated by assessing Calandria structural failure. • Plastic instability, excessive inelastic strains & stress rupture criteria are taken. • Calandria withstood the accident loads for about 70 h. The safety demonstration of nuclear power plant, even for very low probability severe accident scenarios, has been an important requirement. A postulated severe accident is one of the beyond design basis plant states which can have significant consequences resulting from the nuclear fuel degradation. Such accident scenarios call for detailed safety assessment and provision of design features and guidelines for the operating personnel to manage the accident. In the present work, structural integrity assessment of Calandria assembly of a typical 540 MWe Indian Pressurized Heavy Water Reactor has been carried out to assess the duration of in-Calandria retention of core debris/corium for a postulated severe core damage accident scenario. The premise of the study, methodology and analysis tools used to assess structural integrity and failure timeline are discussed. Sequentially coupled thermo-mechanical analysis has been carried out for the thermal and mechanical loads arising out of the postulated accident and structural response of Calandria assembly is evaluated. The Calandria undergoes significant creep/plasticity deformation after loss of cooling through Calandria vault water. Plastic instability, excessive inelastic strains and creep-stress rupture criteria are considered to assess the structural failure of Calandria assembly. It is observed that the Calandria is able to maintain the structural integrity for reasonable duration of more than two days without any management action. Thus, the robustness of the PHWR design is demonstrated. Realistic assessment of timeline of in-Calandria corium/debris retention under such severe core damage accident loading are vital for assessing the severe accident management guidelines. [ABSTRACT FROM AUTHOR]

• A severe accident analysis tool with simple 1-D semi empirical models, is developed. • The tool, which is named as COCOA, is able to predict thermal-hydraulic conditions. • COCOA gives comparable calculation result with COMETA. As molten corium is released from the reactor pressure vessel (RPV) under severe accidents, the molten corium is going to experience various complicated phenomena. If there is no steam explosion, the corium is going to be ejected through the reactor vessel, fragmented, and cooled through the interaction between the corium and water existing in the cavity during its falling stage. Our research group has recently developed a COrium COolability Analysis tool named COCOA. COCOA is the transient calculation tool for the evaluation of the phenomena from the molten corium release to debris formation through the corium falling stage of FCI. COCOA was corporately developed to calculate the transient results of the pressure, the released energy, the average water temperature, fragmented particle size distribution, particulate debris fraction etc., during the corium falling stage of the FCI phase. COCOA is validated against FARO, which is a large-scale corium experiment. Comparing to COMETA, which is a 2D severe accident analysis code, COCOA showed better accuracies on the prediction of the pressure, the released energy, and the particulate debris fraction during the corium falling stage of the FCI phase. COCOA gives 16.9% and 26.5% of NRMSD for the pressure and the released energy, while COMETA does 42.2% and 36.0% of NRMSD. Also, COCOA and COMETA predict the particulate debris fraction with 27.22% and 39.25% of RMSE, respectively. Since COCOA is a semi-empirical model-based simple 1-D tool, only small computational effort is required for the calculation. [ABSTRACT FROM AUTHOR]

• Experimental studies of the corium melt oxidation kinetics were carried out. • Various melt structures had been studied: one-, two- and three-liquid. • Melt structures had been studied both in the presence and absence of oxide crusts. • Air, argon-oxygen mixture and steam were used as oxidant. • Melt oxidation control processes had been identified. Prototypic oxidic-metallic molten corium oxidation has been studied in the conditions applicable for a severe accident at the NPP with PWR. Suboxidized oxidic melt and different pool configurations have been examined – one-liquid oxidic, two-liquid oxidic-metallic with surface and bottom positions of metallic liquid, and three-liquid with one oxidic and two metallic layers. The sensitivity of oxidation rate to the presence of oxidic surface and intermediate crusts is discussed. Produced experimental data for different pool structures and oxidants have been summarized. [ABSTRACT FROM AUTHOR]

• Fractional Release Rates of two samples (in-vessel and ex-vessel) were obtained. • Fraction of Inventory in Aqueous Phase of Mo, Ba, Sr, Rh, La, etc. was obtained. • Geochemical modeling was performed to better understand experimental results. On 11 March 2011, the Great East Japan Earthquake occurred and caused a severe nuclear accident at the Fukushima Daiichi Nuclear Power Station (1F). A policy for corium debris retrieval will be decided in the near future. To this end, debris management after removal from reactors must be optimized. From the viewpoint of cost and waiting for definitive solutions, interim or long-term underwater storage is one of the considered options after debris removal. Leaching data on the long term behavior of debris in water are therefore needed before debris removal. In addition, it is important to understand potential differences in leaching behaviors between debris and Spent Fuel (SF) to get comparative elements. For these reasons, simulated in-vessel and ex-vessel corium debris, fabricated in the Colima experimental facility set up in the PLINIUS platform at CEA Cadarache, were selected and leaching experiments were carried out under oxidizing conditions. In parallel, geochemical modeling was performed to better understand the experimental concentrations, pH evolutions and secondary phase's formation. Finally, the Fractional Release Rates of the (U, Zr)O 2 matrix for the two kinds of samples (in-vessel and ex-vessel) were found to be close to or one order of magnitude lower than that of SF under oxidizing conditions (from 10−6 to 10−7 per day) but the release processes are different. Indeed, the Fraction of Inventory in Aqueous Phase of Mo, Ba, Sr, Rh, La, Nd, Ce and Y tended to be higher than that of uranium and zirconium in this study whatever the element considered. The release of simulating PFs, including low volatile element like trivalent rare earths, does not appear to be fully controlled by that of the (U, Zr)O 2 matrix from a macroscopic and average point of view. The FPs could be partially located outside the (U, Zr)O 2 solid solution as a function of the corium history (quenching rate and thermal history, chemical interactions...). The location of inventories and their chemical forms are not very well known moreover uncertainties related to heterogeneity are important and may also impact the calculation of the global released fractions over time. For the ex-vessel sample, releases into the solution of the main concrete elements (Si, Ca, Mg) seem well correlated and dissociated from the behavior of the (U, Zr)O 2 phase. Further studies are needed especially on real or prototypical corium approaching the conditions of the Fukushima accident. The study of simpler and model systems could also provide information on the reactivity of the (U, Zr)O 2 matrix. [ABSTRACT FROM AUTHOR]

• The different ablations regimes for LCS and siliceous concretes are explained. • Ablation instabilities are linked to transitions between these regimes. This paper deals with the extension of the Transient Interface Model (TIM), initially developed for in-vessel and ceramic dissolution situations, to molten core concrete interaction (MCCI) conditions. The application of the TIM model to MCCI leads to evidence of two interaction regimes, depending on the nature of the concrete. For siliceous concrete, there is no crust nor mushy zone at the ablated corium/concrete interface and the interface temperature is close to the liquidus temperature of concrete (Low Interface Temperature – LIT regime). For limestone common sand concrete, a mushy zone or crust can form at the interface between the corium and the concrete and the interface temperature (between mushy zone and melt) is close to the pool liquidus temperature (High Interface Temperature – HIT regime). No intermediate regime exists, but rather regime transitions (HIT to LIT, or LIT to HIT), which may occur on a limited number to all interfaces during MCCI. The work described herein demonstrates that these regime transitions are abrupt and happen within a short time period. These regime transitions are shown to cause ablation instabilities. Many combinations of regime transitions on various interfaces at various pool temperatures (in comparison to pool liquidus) are possible. Only a limited number of these transitions have been calculated; the results indicate that different trends in temperature evolutions observed during MCCI tests can be reproduced. New calculations of the CCI2, CCI3 and M3B tests are presented. [ABSTRACT FROM AUTHOR]