Core Melt Research Articles

Experimental investigation on sodium compatibility of magnesia during corium relocation in SFR for core catcher application

Core melt accident is a very low probability event in Sodium Cooled Fast Reactors (SFRs) due to presence of several inherent and engineered safety features. However, such accidents are considered in the design of reactor to ensure safe mitigation and containment. Core Catcher (CC) is a passive device installed below the core in lower plenum of reactor vessel to retain and cool the degraded core debris in subcritical state. In the earlier reactors, partial core melting (typically seven subassemblies) was considered as design criteria for CC. For future SFRs, retention and cooling of whole core has to be considered as design basis in view of enhanced safety. Whole core relocation would impose significantly higher thermal load on structural material of the CC. A sacrificial lining of refractory ceramic materialis under development for thermal protection of CC during whole core retention. Based on preliminary studies, sintered magnesia has been chosen as candidate material and experiments are conducted to investigate its thermal stability in sodium at high temperature. Standard test specimens of sintered magnesia were exposed to sodium at its boiling point under simulated loading of core debris corresponding to whole core relocation in a medium size SFR. Posttest characterization was carried out using XRD and SEM coupled with ESD techniques. Changes in compressive strength and ultrasonic pulse velocity of specimens after high temperature sodium exposure were measured to assess degradation of the specimens. The experimental results indicated satisfactory performance of high purity sintered magnesia bricks for application to the CC as sacrificial material.

Probability analysis of ex-vessel steam explosion induced pressure load based on analysis code SEINA

In nuclear reactor severe accidents, the interaction of molten core melt and coolant may lead to steam explosion, which seriously threatens the safety of the reactor. The four initial parameters that affect the steam explosion were randomly sampled and calculated multiple times by steam explosion analysis code SEINA, and the probability distributions of steam explosion pressure peak and pressure impulse under uncertain initial conditions are obtained. The results show that the peak pressure of ex-vessel steam explosion is roughly normally distributed, and the pressure peak of 90% of the cases under submerged and non-submerged conditions within the range of 10–60 MPa and 20-80Mpa. respectively. By calculating the four key correlation coefficients between input parameters and output parameters, it is concluded that the influence of melt temperature and breach size on output parameters is greater than that of system pressure and coolant temperature.

Accident Simulation Study for Nuclear Power Plants Impacting Louisiana: Differences in 2017 vs. 1992 Protective Action Guidelines.

Louisiana has aligned its radiological emergency program with the 2017 US Environmental Protection Agency Protective Action Guides Manual but has added a child thyroid dose evacuation threshold in lieu of distributing potassium iodide to the public. The nuclear power plants will continue to align with the 1992 manual for the foreseeable future, which could lead to possible accident scenarios in which state recommendations would differ from those of the utility. The objective of this study is to predict what accident and weather conditions will lead to a differing set of recommendations. This study performs a representative set of simulations of potential nuclear power plant accidents using a combination of the RASCAL software package, provided by the Nuclear Regulatory Commission, and a Software system used by Entergy combining an older RASCAL dose modeling methodology with plant-specific input. Four preliminary results of this study are presented: a spent fuel fire where differences in whole body dose lead to very different evacuations, a loss of coolant accident in which the child thyroid dose is the determining factor, a core melt accident using stack monitors to locate the evacuation threshold point, and a spiked coolant accident that could lead to an evacuation order before the plant declares a General Emergency. Weather plays as great a role as accident conditions in determining whether the evacuation recommendations differ. The completed results of this study can provide guidance to states as they evaluate the transition to the 2017 guidelines.

Studies and cross-comparisons of severe accident prevention and mitigation capabilities of a SFR and a GFR

CEA has developed Generation IV fast neutron reactors with gas and sodium coolants in the two last tenth of years. Namely these reactor projects were the GFR2400 (gas-cooled fast reactor of 2400 thermal megawatts) and ASTRID (sodium-cooled fast reactor demonstrator of 1500 thermal megawatts). The objective of this paper is to provide a cross-comparison of the severe accident prevention and mitigation capability of these reactor concepts based on the work done during a significant time period and taking benefit of the distance since these studies. This comparison can highlight both generic trends resulting from a common study approach used for both concepts and from very detailed results obtained during their conceptual design studies. Despite their power difference and their fuel element design specificities, the main study results and conclusions are quite generical of the two concepts as explained in the paper. Thus they could be extrapolated to other reactor design providing main reactor features are kept (core materials, coolant, neutron spectrum, etc.). The assessment of core melting prevention relies on the study of the natural behavior of the GFR2400 and of ASTRID when facing the various accident sequence families. Then, the efficiency and the features of the systems to be foreseen for core degradation prevention are presented. As far as mitigation is concerned, all the consequences of core melting are investigated (i.e. the induced loadings in terms of nature and of range) by considering various core degraded states. Based on the magnitude of these loadings, the needs of mitigation means are assessed for each concept. Among other trends, the presented work shows the very good prevention capability of the SFR concept but the necessity to mitigate the fast vaporization and expansion of degraded core materials. Conversely, the limited coolant capability of the GFR concept and its low thermal inertia require a pressurized gas circulation into the core, limiting its prevention capability whereas its core melting should not induce substantial mechanical loadings of the reactor structures. However, for this last concept, thermochemical interactions between the core materials are an issue deeply investigated in order to understand and simulate core degradation.

Modeling the process of decay heat imitation in the corium at the «Lava-B» facility

It is known that corium is produced during the development of a severe accident at a nuclear power plant from a melt of core materials [1]. An important feature of the corium formed in an operating reactor is the presence of decay heat. Thus, it is especially important to take into account the presence of decay heat when conducting calculations and physical experiments, since it influences significantly on the nature of interaction between corium melt and structure materials of a reactor facility. For that reason, methods of decay heat simulation are subject to quite serious requirements to the depth and intensity of energy release in the corium when conducting different experiments. This article considers induction and plasmatron heating technologies as methods to simulate decay heat in corium at the Lava-B facility.  Characteristics of the selected methods of heating the corium prototype was analyzed by means of computer modeling. As a result of the work, parameters of melt heating using each of the considered methods were determined, and they were also cmpared. The used thermophysical models were created in the widely known ANSYS software based on the experimental section, which was used in one of the experiments at the Lava-B facility [2]. The main parameters of the corium-heater system were obtained by computer modeling for each of the considered methods and limits of their applicability for simulation of decay heat were determined when conducting the experiments at the facility.

Development of Self-Shielded Flux-Cored Wires for Arc Welding of Low-Alloy Steels

Current aspects of the developing of modern self-shielding flux-cored wires composition for arc welding of low-alloyed steels are considered. Advantages and disadvantages of flux-cored wires of carbonate-fluorite, oxide and oxide-fluoride types of are shown in comparison. The effectiveness of gas shielding of molten metal at welding with self-shielding flux-cored wires of carbonate-fluorite type is analyzed considering the thermal properties of their cores. It is shown that to improve reliability of gas shielding at welding using the wires of this type it is important not only to ensure generation of sufficiently large volume of shielding gases at thermal destruction of the wire core, but also to control this process, providing gas evolution at all stages of heating and melting of the wire. The results of complex thermal analysis of the wire core mixtures containing, for example, lithium carbonate show substantially large heat losses for heating and melting of the wire core, which are accompanied by the development of energy-intensive processes of thermal destruction of core components. It is shown that the limitation of lithium carbonate content in the wire at the level of not more than 2 wt. % allows not only to preserve welding arc burning stability at the acceptable level but also to provide effective gas shielding of molten metal and easy separation of slag crust. The control of thermochemical reactions in the core is achieved by selection of its proper composition to ensure favorable melting of flux-cored wire and electrode metal transfer to the welding pool. Results of metallographic examinations of distribution and composition of non-metallic inclusions in metal of the welds made with wires of the oxide and oxide-fluoride types are presented. Main properties of the developed self-shielding flux-cored wires and recommendations on welding are given in conjunction with prospective fields of their application.

Experimental Investigation of Melt Coolability Behavior in an Ex-Vessel Core Catcher: The Effect of Flooding Time

Abstract Some of the advanced nuclear reactors employ an ex-vessel core catcher to mitigate core melt scenarios by stabilizing and cooling the corium for a prolonged period by strategically flooding it. The side indirect cooling with top flooding strategy described in this study may lead to water ingression through the melt crust, which may lead to interaction between unoxidized metal and water, leading to hydrogen production. In order to avoid this deleterious scenario, water ingression into the interior region of the melt should be avoided. The studies described in this paper show that water ingression depends on the flooding strategy, i.e., the time delay between top flooding and melt relocation. Two experiments under identical conditions of simulant temperature, melt material, and test section geometry were conducted with simulated decay heat of 1 MW/m3. Sodium borosilicate glass was used as the corium simulant. In the first experiment, water was flooded onto the top of melt pool soon after melt relocation. In the second experiment, water flooding at the top of melt pool was made after 30 min of the melt relocation. The results show that a finite time delay of introduction of water onto the top of the melt pool is paramount to engender the development of a stable crust around the melt and therefore eliminating water ingression into melt pool and ensuring controlled coolability of the melt.

High-Temperature Characterization of Melted Nuclear Core Materials: Investigating Corium Properties Through the Case Studies of In-Vessel and Ex-Vessel Retention

During a severe accident in a nuclear reactor, the molten core—or corium—may be relocated into the reactor vessel’s lower plenum in case of core support plate failure. The severe accident management strategy for In-Vessel Retention—or IVR—consists in stabilizing the corium within the reactor pressure vessel by external cooling of the vessel’s lower head. If now, the vessel fails due to excessive thermal loading on its walls, the Ex-Vessel Retention—or EVR—strategy is adopted. In this case, the core melt stabilization can be achieved by effective corium spreading, either in the reactor vessel cavity or in a dedicated “core-catcher”, and cooling by water. The success of both strategies highly depends on the corium behavior at high temperatures, conditioning vessel’s integrity for IVR, and promotion for the spreading of the EVR. This involves a variety of fundamental mechanisms closely related to heat and mass transfer regimes prevailing at the system scale, which requires further analytical and experimental insight to determine the primary mechanisms and feed the modeling tools, allowing the numerical simulations of severe accident scenarios.Within the framework of corium characterization at high temperatures, the present study aims at filling the lack of such fundamental data as density, surface tension, liquidus and solidus temperatures, and viscosity. In order to accurately measure these properties at high temperatures, the VITI facility is designed with various configurations. Concerning IVR, the influence of density and surface tension is particularly highlighted through VITI-SD and VITI-MBP configurations, and practical applications of experimental results are finally discussed, in link with the focusing effect issue at the thin upper metallic layer of the corium pool. Concerning EVR, the properties of interest are solidus/liquidus temperature and dynamic viscosity, and typical experimental results obtained through VITI-VPA and VITI-GFL configurations are discussed in view of characterizing corium spreading.

A numerical investigation into Metallic-Melt continuous drainage in the core support plate region of a BWR for the initial phase of core melt progression

Under a typical uncovered core scenario in a Boiling Water Reactor (BWR), it is not clear that a large core molten pool will result as occurred in the Three Mile Island, Unit 2 (TMI-2) accident. Sandia National Laboratories (SNL) has propounded the alternative “Continuous Drainage” melt progression for the scenario, where molten material drains from the core region without the formation of any stable crusts or blockages. Hence the present simulation study was conducted to investigate the behavior of the metallic-melt continuous drainage and the drainage pathways during the initial phase of core melt progression in a BWR under an uncovered core condition like that encountered in the Fukushima Daiichi Nuclear Power Plants (NPPs) accident experienced by Tokyo Electric Power Company Holdings (TEPCO). Researchers around the world have conducted numerous sophisticated and comprehensive studies using simulation codes and experiments to investigate the Fukushima accident. Melt drainage with deforming components is subject to many uncertainties in a core support plate region possessing rather complicated geometries. Additionally, there is very limited experimental and numerical simulation knowledge on melt drainage in such a region. Thus, in recent years, Japan Atomic Energy Agency (JAEA) has experimented on melt drainage using a simplified apparatus to reproduce the behavior. The present paper provides simulation results on melt drainage in the core support region using SAGITTARIUS, an in-house three-dimensional simulation code capable of consistently simulating melt drainage with crust formation, local blockage, and structural damage. The in-vessel simulation target with a prototypic BWR-composition and -geometry is composed of four fuel assemblies (4 × 4 fuel rods per assembly), an absorber blade, a support piece, a core support plate, an Instrument Guide Tube (IGT), and a Control Rod Guide Tube (CRGT) with a Control Rod Drive Housing (CRDH). We estimated the drainage speed for the JAEA experiment to validate the SAGITTARIUS function. The drainage speed was evaluated to be 3.1 m/s, a speed corresponding to that observed in the core support plate region. We also used SAGITTARIUS to simulate the behavior of the metallic-melt drainage and temperature rise of the structure. The results from SAGITTARIUS on the trend and values of the temperature rise agreed with the measured values. On the basis of this finding, we conducted numerical investigations on the metallic-melt drainage of B4C-Stainless Steel (SS) eutectic from the absorber blade and Zry-melt from the fuel rods. The simulation demonstrated the capacity to describe the following features: (i) visual demonstration of metallic-melt drainage from the absorber blade and fuel rods, (ii) aggressive Zry-melt attack to the inlet orifice, (iii) crust formation with local blockage in the narrow drainage paths, and (iv) sedimentation of solidified melt on the bottom faces. These findings were consistent with the concept of the “Continuous Drainage” proposed at SNL and we supported the SNL scenario by the numerical investigation. These investigations probably provide useful knowledge during the discharge of damaged fuels or crust in the Fukushima-Reactor Pressure Vessels (RPVs).

SEINA: A two-dimensional steam explosion integrated analysis code

In the event of a severe accident, the reactor core may melt due to insufficient cooling. the high-temperature core melt will have a strong interaction (FCI) with the coolant, which may lead to steam explosion. Steam explosion would pose a serious threat to the safety of the reactors. Therefore, the study of steam explosion is of great significance to the assessment of severe accidents in nuclear reactors.This research focuses on the development of a two-dimensional steam explosion integrated analysis code called SEINA. Based on the semi-implicit Euler scheme, the three-phase field was considered in this code. Besides, the influence of evaporation drag of melt and the influence of solidified shell during the process of melt droplet fragmentation were also considered. The code was simulated and validated by FARO L-14 and KROTOS KS-2 experiments. The calculation results of SEINA code are in good agreement with the experimental results, and the results show that if the effects of evaporation drag and melt solidification shell are considered, the FCI process can be described more accurately. Therefore, it is proved that SEINA has the potential to be a powerful and effective tool for the analysis of steam explosions in nuclear reactors.

The effect of micro-encapsulation on thermal characteristics of metallic phase change materials

Micro-encapsulated phase change material with metallic core can work under high temperature for energy storage purpose, making it an appealing candidate for renewable energy technologies, such as thermal energy storage for concentrating solar power plants. The objective of this work is to study the impact of encapsulation on the thermal performance of the micro-encapsulated phase change materials under various operating conditions. To evaluate the thermal performance of the micro-encapsulated phase change material particles, the governing equation of transient heat conduction with phase change in spheres with composite walls is solved numerically. The impact of encapsulation shell on the convective heat transfer rate, total energy absorption, latent heat ratio, energy density, and duration of phase change is presented and analyzed under different operating conditions. Three different core materials including tin, aluminum and copper with distinct Stephan numbers are investigated. In addition to the encapsulation shell effect, this model can be used to study micro-encapsulated phase change materials with different material compositions, core to shell ratios, and it can be used to analyze the impact of air gaps in between materials. Understanding the thermal performance at particle level is essential, and the results of this study could potentially be used as input for thermal energy storage system level analysis, such as the charging and discharging processes. • Encapsulation enhances convection due to the increased surface area • Thicker shell shortens core melting time • Heating time depends on the ratio of heat capacitance and convection rate • Optimal shell thickness exists for achieving high energy storage and fast absorption

Experimental Research for CHF Sensitivity of Heat Flux Distribution under IVR Conditions

In-vessel retention (IVR) through external reactor vessel cooling (ERVC) is one of the most effective severe accident mitigation measures in the nuclear power plants. The most influential issues on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the lower head, and the external cooling of reactor pressurized vessel (RPV). In the molten pool research, there are mainly two different molten pool configurations: two layers and three layers. Based on the different distributions of heat flux in molten pool configurations, a new problem was raised: whether the in-vessel heat flux distribution will affect the CHF on the outer wall of RPV and further affect the effectiveness of IVR measures? A full-height external reactor vessel cooling and natural circulating facility was conducted to study the CHF sensitivity of different heat flux distributions. The experimental results show that the characteristics of natural circulation are similar and the CHF of the RPV lower head external surface is not obviously affected under the different heat flux distributions. The varying heat flux distribution during severe accident process will not threaten significantly the success of IVR strategy.

Analysis of 3-layer corium pool models in CISER2.0 code

In present work, the corium pool stratification models in CISER2.0 code for the validity analysis of In-vessel retention (IVR) strategy were performed, and the heat flux distributed on the outside wall of the lower head were calculated. The 3-layer models in CISER2.0 code consist of three models developed by Esmaili & Khatib-Rahbar (EK), Seiler and Salay & Fichot (SF) respectively and a self-developed New Model. Among these models, Seiler Model is more conservative. The SF Model treats the composition of oxide and metal phases based on the corium thermodynamic properties. Different from the previous three models, the New Model can calculate the transient configuration of the corium pool during the process of the core melt falling into the lower plenum. The heat flux on the outside wall of the lower head was calculated for a typical 1000 MW power plant in China. It was found that the 3-layer pool predicted by Seiler Model was not formed due to the insufficient amount of stainless steel in the corium. Since both EK and SF Model require more user input parameters, when one of them changes, the configuration of the corium pool and the heat flux on the outer wall of the vessel may change significantly. The New Model has the minimum initial input parameters, resulting in consistent results for different users, and the transient corium stratification can be obtained by assuming that the corium is completely mixed and reaches thermodynamic equilibrium as it relocating in the lower plenum.

EXPERIMENTAL STUDIES ON DEVELOPMENT OF THE METHOD FOR DETERMINING THE AMOUNT OF IMPURITATED GASES IN NUCLEAR FUEL

The article is devoted to an issue of estimating the impurity gas amount in nuclear fuel in the aspect of the distracting contribution from released gases to the total pressure inside ampoule of the device in the simulating a severe accident with core melting. The paper presents a method based on measuring the pressure and temperature of gas in a closed values of the fuel elements during the fuel melting. The correctness of the developed methodology is confirmed by the results of experiments on the melting of fuel in a pulsed graphite reactor IGR with the implementation of a controlled neutron pulse.

Numerical evaluation of geometric variations in internally heated melt pool

In Vessel Retention (IVR) is considered as one of the best safety strategy for a hypothetical severe accident in light water nuclear reactors (LWR). Since the strategy involves heat transfer between the corium melt pool and surrounding, it is highly impacted by the natural convection behaviour of such internally heated melt pool. Extensive experiments and numerical study has been performed worldwide to understand the behaviour of corium melt pool. COPRA (COrium Pool Research Apparatus) is one such experiment which uses molten salts of non-eutectic binary mixture (20 mol% NaNO3–80 mol% KNO3) and eutectic binary mixture (50 mol% NaNO3–50 mol% KNO3) as simulant material. Geometric variation in COPRA experiment is investigated to get a better understanding of the natural convection behaviour of the melt pool. Although the influence of geometry is reported in literature, the increase in geometry beyond the conventional size is yet to be explored. Eutectic salt (50 mol% NaNO3–50 mol% KNO3) is used as a simulant material for this study. A commercial CFD code ANSYS FLUENT 20.2 is employed for the numerical modelling of the corium pool. K-epsilon (k-ε) turbulence model and solidification-melting model are used in this simulation. Experimental data is used for validating this model. The geometric effect is investigated by assuming the same height and radius for the pool. Also, the ratio of heating power and cooling length is maintained for all the variations of geometry. The temperature field, liquid fraction contour, and velocity distribution of the melt pool are calculated. This paper provides an extension to the existing data thus further enhances the understanding of the thermal–hydraulic characteristics of the corium melt pool.

A PCTRAN BASED ANALYSIS ON THE EFFECT OF BREAKSIZE AND COMPARATIVE STUDY BETWEEN HOT AND COLD LEG LOSS OF COOLANT ACCIDENTS IN VVER 1200 POWER REACTOR

In this paper, a comparative analysis of loss of coolant accident (LOCA) in hot leg and cold leg of primary circuit in a VVER 1200 nuclear power plant is investigated. The effect of break size on the severity of the accident is observed. The break size was varied in the range 200-11350 cm2. For all the accident scenarios, station blackout (SBO) condition is set up. Additionally, it is assumed that no ECCS (Emergency Core Cooling System) is available due to system malfunction. The whole scenario is simulated in PCTRAN (Personal Computer Transient Analyzer) software. Results reveal that with the increase in the size of the break area, the core uncovering time decreases sharply. However, for a break size of 2800 cm2 or smaller, the water level in the core doesn’t drop to zero, indicating that the core is partially uncovered throughout the accident scenario. In case of hot leg LOCA, the draining of the reactor vessel is observed to be more rapid compared to cold leg LOCA, while the core melting started earlier in case of cold leg.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

Abstract The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

Thermal-electrical coupling characteristic analysis of the heat pipe cooled reactor with static thermoelectric conversion

Heat pipe cooled reactor with static conversion is gaining more and more attention because of its unique advantages such as mobility, high reliability, simple structure, long lifetime, and high-power density. A 100 kW heat pipe reactor with the three-segment type thermoelectric generators namely NUSTER-100 is designed by Xian Jiaotong University. Based on the finite element software COMSOL Multiphysics, a multi-physics coupling analysis platform is established for the static heat pipe reactor NUSTER-100. The heat transfer models of the heat pipe and thermoelectric models of the three-segment type TEGs are supplemented in the simulation model. The thermal-electrical coupling characteristics of the static heat pipe reactor are investigated under different operation conditions. For the steady state condition, the temperature difference of the heat pipe between evaporator and condenser stabilizes at around 110 K, the main temperature gradient about 850 K appears between the two side of TEGs. The total output power of the NUSTER system is 123 kW and the system thermoelectric conversion efficiency is 12.3%. For the three rows unloading condition, the maximum temperature increases to 1483 K which is 133 K higher than that on the steady state condition. The output electrical power of the reactor system is 114.5 kW and the system thermoelectric conversion still could reach 11.5%. For the whole reactor unloading condition, the reactor core temperature increases by 393 K compared with the steady state and the maximum temperature is 1743 K which is lower than the core melting temperature. The operating temperatures of the heat pipes are still within the normal operating temperature range of the sodium heat pipe. The heat transfer capacity of the heat pipe is still sufficient to ensure that the core temperature is kept within the safe limits. In summary, the unloading condition will cause the core temperature to rise significantly. Therefore, the system safety characteristics under accident transient conditions need to be focused coupling with the unloading condition in the further research.

Experiments of melt jet-breakup for agglomerated debris formation using a metallic melt

In severe accidents in a light water reactor, the relocated molten core (so-called corium or melt) can form a debris bed. The debris bed coolability is a critical issue for prevention and mitigation of the molten core–concrete interactions. Agglomeration has a serious impact on assessment of debris bed coolability if agglomeration forms massive debris (so-called agglomerated debris) by merging of melt particles with others when the melt particles accumulate on a floor. This paper presents the results of melt jet-breakup experiments for agglomerated debris formation using a simulant metallic melt. The experiments injected a melt jet of a low-melting point metal through a circular nozzle into a test section filled with coolant water. The particles were generated due to the melt jet-breakup accumulated on to a catcher, which is a flat plate made of stainless steel, installed in the test section. A high-speed video camera imaged particle formation and accumulation on the catcher plate. Agglomerated debris was confirmed by morphological investigation of the recovered debris. The experimental results revealed the effects of the melt jet injection conditions (melt temperature, coolant temperature, and coolant depth) on the mass fraction of agglomerated debris. On the basis of the experimental results, we proposed a simple correlation to estimate the mass fraction. The simple correlation successfully reproduced the mass fraction of agglomerated debris obtained in the DEFOR-A test [Kudinov et al., Nucl. Eng. Des., 301 (2013), 284–295]. The experimental data base presented in this paper makes further contributions to the modeling and validation of mechanistic models or simulation tools for agglomerated debris formation.

Multi-dimensional Simulation of Phase Change by a 0D-2D Model Coupling via Stefan Condition

Considering phase changes associated with a high-temperature molten material cooled down from the outside, this work presents an improvement of the modelling and the numerical simulation of such processes for an application pertaining to the safety of light water nuclear reactors. Postulating a core meltdown accident, the behaviour of the core melt (aka corium) into a steel vessel is of tremendous importance when evaluating the vessel integrity. Evaluating correctly the heat fluxes requires the numerical simulation of the interaction between the liquid material and its solid counterpart which forms during the solidification process, but also may melt back. To simulate this configuration, encountered in various industrial applications, one considers a bi-phase model constituted by a liquid phase in contact and interaction with its solid phase. The liquid phase may solidify in presence of low energetic source, while the solid phase may melt due to an intense heat flux from the high-temperature liquid. In the frame of the in-house legacy code, several simplifying assumptions (0D multi-layer discretization, instantaneous heat transfer via a quadratic temperature profile in solids) are made for the modelling of such phase changes. In the present work, these shortcomings are illustrated and further overcome by solving a 2D heat conduction model in the solid by a mixed Raviart-Thomas finite element method coupled to the liquid phase due to heat and mass exchanges through Stefan condition. The liquid phase is modeled with a 0D multi-layer approach. The 0D-liquid and 2D-solid models are coupled by a Stefan like phase change interface model. Several sanity checks are performed to assess the validity of the approach on 1D and 2D academical configurations for which exact or reference solutions are available. Then more advanced situations (genuine multi-dimensional phase changes and an “industrial-like scenario”) are simulated to verify the appropriate behavior of the obtained coupled simulation scheme.