Corium Pool Research Articles

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.

Numerical investigation of convection heat transfer characteristics in sloshing corium pools

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.

Heat transfer characteristics of single-layer and two-layer corium pool in elliptical lower head

In this study, we investigated the characteristics of natural convection heat transfer of single-layer and two-layer corium pool in the elliptical lower head by the numerical method. Numerical simulations were performed using FLUENT software with the WMLES (Wall-Modeled Large Eddy Simulation) turbulence model, solid-liquid phase change model and the VOF (Volume of Fluid) model. The heat transfer characteristics such as pool temperature field, flow velocity field and heat flux distribution were obtained. For the single-layer corium pool, different degrees of melting occurred on the wall surface from about 1712 mm arc length inside the inner wall to the liquid level. The transient velocity field of single-layer corium pool showed that the natural convection first occurs near the wall surface and then gradually forms in the center of the corium pool as the temperature rises. When the quasi-steady state is reached, the upper part of the pool forms strong turbulence, while the lower part has lower velocity and forms transverse flow. For the two-layer corium pool, the lower head wall is more severely melted than the single-layer configuration. Whereas, the crust at the wall near the stratified interface is only slightly melted due to the lower flow rate and weaker heat transfer capacity. The research revealed the flow and heat transfer mechanisms of corium pool, and provided reference for the technology design, system optimization and safety evaluation of the IVR (In-Vessel Retention) – ERVC (External Reactor Vessel Cooling) severe accident mitigation strategy for the large-scale advanced pressurized water reactor.

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.

Corium Pool Behaviour in the Core Catcher of a Sodium-Cooled Fast Reactor

Corium Pool Behaviour in the Core Catcher of a Sodium-Cooled Fast Reactor

Mass transfer between a continuous oxide phase (U-O-Zr) and a steel droplet at liquid state: Potential impact on corium pool behaviour during in-vessel melt retention

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.

Severe accident analysis of the Qinshan Nuclear Power Plant and evaluation of boundary conditions for ex-vessel heat transfer

Severe accident analysis is critical for the operation and licensing of new generation reactor designs as well as for the prevention and management of severe accidents. The Severe Accident Management Guidelines (SAMGs) are developed and implemented for prevention and mitigation measures of a severe accident. Retention of corium in the reactor vessel via ex-vessel cooling is one of the most promising mitigation strategies. Its success requires a thorough analysis of the reactor core following a severe accident to determine the boundary conditions for external cavity flooding. In this study, two loss of coolant accident cases at the Qinshan nuclear power plant are analyzed using RELAP5/SCDAP simulations. Passive accumulators are credited in the first case (Case-1), but not in the second (Case-2). The important severe accident events are studied for the plant and the sequence of events and melt conditions are captured. Hydrogen generation has been estimated to be 120 kg in Case-1 and 108 kg in Case-2. In both instances, a molten pool is formed. The corium reached a maximum temperature of 3016 K and 3006 K, respectively in Case-1 and Case-2. The metallic/oxidic layer properties and heat transfer coefficient of the corium pool as well as the ex-vessel heat transfer boundary conditions are evaluated analytically for the reference plant. The analytically determined temperatures of oxidic and metallic phases in the corium pool are 3006 K and 1919 K respectively. The upward and sideward heat fluxes (vessel wall) are determined to be 298.3 kW/m2 and 730.75 kW/m2 respectively, while the heat transfer coefficients for oxidic and metallic phases are 2.80 W/m2-K and 4.42 W/m2-K respectively.

A two-phase mathematical model to describe the dissolution of corium crust by molten steel

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.

Numerical simulations of the Rayleigh-Bénard-Marangoni convections in a thin metallic layer

Within the context of nuclear power safety, the In-Vessel Retention (IVR) strategy consists in sustaining the integrity of the reactor pressure vessel by External Reactor Vessel Cooling (ERVC). However, due to the ablation of the vessel wall and the thermochemical effects, a thin metallic layer may be formed on top of the corium pool and lead to the vessel failure. In this article, numerical simulations of the thermal–hydraulic behavior of the thin metallic layer are carried out to reduce the uncertainties in the heat flux evaluation. Specifically, the Rayleigh-Bénard and Bénard-Marangoni convections are considered in the heat transfer models of the metallic layer. The TrioCFD code is validated for the Bénard-Marangoni convection, then applied into the numerical simulations of the metallic layer with different top boundary conditions: pure radiative heat transfer, and radiative heat transfer plus positive or negative Marangoni effects. The temperature profile, mean temperature and focusing effect of the metallic layer are evaluated and compared for various cases. Results show that the Marangoni effects have to be considered for small metallic layer thicknesses. Furthermore, the layer can be divided into two regions, of which the heat flux correlations can be evaluated independently, giving a better estimation of the heat flux profile.

Numerical simulation of CS28-1 experiment by using CANDU severe accident analysis code, CAISER

Due to the demands for an accurate and realistic severe accident code in the PHWR safety analysis, a new severe accident code called CAISER is being developed at Korea Atomic Energy Research Institute (KAERI). The CAISER code simulates core degradation phenomena occurring in a calandria tank, and it consists of two main modules: A fuel rod degradation module and a fuel channel degradation module. The fuel rod degradation module simulates the severe accident phenomena in a fuel channel, which includes a core uncovery, fuel rods heatup, hydrogen generation due to steam-Zr oxidation, fuel rods slumping, fuel rods melting and relocation, and thermal interaction of relocated molten mass with a pressure tube or a calandria tube. The fuel channel degradation module simulates the overall severe accident phenomena in a calandria tank, including the sagging of a fuel channel, debris bed formation caused by a fuel channel failure, the molten pool formation, and the calandria tank failure. Each module is tracking mass and energy changes for the main components, which are fuel, cladding, pressure tube, calandria tube in a fuel rod degradation module, and debris bed, metallic and oxidic corium pools in a fuel channel degradation module. These two modules are closely interconnected to simulate phenomena both in a fuel channel and in a calandria tank at the same time. In this paper, the numerical modelling in the CAISER code is described, including the node system and the governing equations for each main component. For the purpose of CAISER code validation, numerical simulations were conducted for the CS28-1 experiment. The simulation results show that the temperature distributions calculated by CAISER are in line with the experimental data.

Studies on Key Effect Factors of Natural Circulation Characteristics for Advanced PWR Reactor Cavity Flooding System

In order to enhance the ability of severe accident mitigation for Pressurised Water Reactor (PWR), different kinds of severe accident mitigation strategies have been proposed. In-Vessel Retention (IVR) is one of the important severe accident management means by External Reactor Vessel Cooling. Reactor cavity would be submerged to cool the molten corium when a severe accident happens. The success criterion of IVR strategy is that the heat flux which transfers from the corium pool must be lower than the local critical heat flux (CHF) of the reactor pressure vessel (RPV) outside wall and the residual thickness of the RPV wall can maintain the integrity. The residual thickness of RPV is determined by the heat flux transfer from the corium pool and the cooling capability of outer wall of the RPV. There are various factors which would influence the CHF and the cooling capability of outer wall of the RPV. In order to verify the optimized design which is beneficial to the heat transfer and the natural circulation outside the actual reactor vessel, a large-scale Reactor Vessel External Cooling Test (REVECT) facility has been built. A large number of sensitivity tests were carried out, to study how these sensitivity factors affect CHF value and natural circulation. Based on the test results, the structure of the test section flow channel has an obvious effect on the CHF distribution. The flow channel optimized can effectively enhance the CHF value, especially to enhance the CHF value near the “heat focus” region of the molten pool. The water level in the reactor pit has also a great impact on the natural circulation flow. Although natural circulation can be maintained with a low water level, it will lead to a decrease of the cooling capacity. Meanwhile, some noteworthy test phenomena have been found, which are also essential for the design of the reactor pit flooding system.

Modeling of the corium crust of a stratified corium pool during severe accidents in light water reactors

During Severe Accidents (SAs) in Light Water Reactors (LWRs), the core heat up due to the lack of fuel cooling in the Reactor Pressure Vessel (RPV) may lead to the formation of a so called corium pool (molten core oxidic and metallic liquid materials) relocating in the lower head of the RPV. Within the framework of SA, one strategy to minimize risks of containment failure is the In Vessel Retention (IVR) strategy. To demonstrate the feasibility of such a strategy and to ensure the vessel integrity during the accident, one of the key points is to correctly evaluate the thermal load applied on the RPV wall by the lower head corium pool. Among all the complex coupled transient phenomena taking place in the RPV, the possible solidification of the corium pool, forming the so-called corium crust, has been shown by experimental and numerical results to be a key phenomenon for the IVR strategy. However, in most SA codes, the corium crust is only taken into account as a stationary model playing the role of a thermal resistance between the corium pool and the RPV wall. In this paper, a transient corium crust model is presented which explicitly takes into account the mass, temperature, composition and thermodynamic properties of the crust and pool material. Numerical results, in the context of the IVR strategy, highlight the impact of the transient corium crust modeling in comparison to a stationary modeling. The proposed model opens new perspective of modeling such as the dissolution of the crust and/or its thermo-mechanical behavior.

Experimental research on the influence of rolling motions on the convection behaviors inside two-layer corium pools

In realms of nuclear safety analysis, In-vessel retention (IVR) has been a research focus as it is a promising mitigation strategy for the postulated severe accidents in pressurized water reactors (PWRs). However, the recent development of ocean floating reactors for civil use has greatly motivated the pertinent studies under motion conditions, which can be much different from the known investigations conducted under static condition without the additional acceleration forces induced by motions. Thus, through experimental research, this paper extensively studied the influence of rolling motions on the convection behaviors inside two-layer corium pools, which are the major concerns for IVR research. Transient behaviors of the layered corium pools were obtained, and sensitivity analyses were conducted on three critical parameters, i.e. maximum rolling angle, rolling period and the thickness of metal layer. Experimental results show that the two-layer fluids can stay basically stratified under rolling conditions but mixing phenomena can be observed in the interface nearby region when rolling motions become intense enough. Generally, rolling motions can weaken the thermal stratification inside the layered corium pool, and enhance heat removal capacity from the corium pool to the outside cooling boundary. These phenomena indicate a good chance for the efficient decay heat removal from the corium pool under rolling motions. This study is supposed to provide valuable references to assess the success probability of IVR design applied in reactors operated under ocean conditions.

Elaboration of a Phenomena Identification Ranking Table (PIRT) for the modelling of In-Vessel Retention

The strategy for In-Vessel Retention (IVR) of corium is currently considered for several new reactor designs in various countries. One of the issues for the demonstration of the success of this strategy is that there are significant uncertainties in physical modelling of corium in the lower plenum and its transient chemical and thermal interactions with the vessel leading to its significant ablation. Since the initial approaches developed in the nineties for the AP600 and Loviisa VVER-440 plant, knowledge about corium pools (thermochemistry and heat transfer characteristics) and mechanical behaviour of the vessel has improved and it is now possible to model more accurately this phenomenon. The H2020 European project IVMR (In-Vessel Melt Retention) addresses the issue of selecting and improving IVR models for safety evaluation. As a first step, a Phenomena Identification Ranking Table (PIRT) involving the relevant physical processes was developed.The methodology followed to build this PIRT results from the consideration of a few principles: (a) Identifying and separating the risks with respect to which the importance of a physical process is evaluated. (b) Defining physical processes or parameters which can be considered as independent of the other ones. (c) Avoiding expert judgement, as much as possible, and using, instead, the results of previous sensitivity studies to estimate the impact of each physical process or parameter.In order to obtain results representing the currently “shared” state of knowledge, code developers of the most widely used codes were consulted. It includes ASTEC, MELCOR, MAAP (EDF version), SOCRAT, ATHLET-CD integral codes and PROCOR, SIMPLE, HEFEST_URAN, and IVRSYS dedicated codes (specific to IVR calculations).One of the positive outcomes of this PIRT is that the results show significant tendencies, with distinct groups of phenomena or parameters. This allows the identification of the uncertainties and of the phenomena/variables with the highest (or lowest) importance. Some dispersion in the results could also be noticed but it can be understood either because of the variability due to reactor design or the impact of models chosen by experts in the codes they have developed. When the dispersion cannot be explained, it indicates that the phenomenon is poorly understood and may deserve further research. Among the phenomena with highest importance, the heat transfers in the top metal layer and the chemical/thermal interaction with the oxide crust have been identified, as well as the transient formation and stratification of metal and oxide layers, including thermochemical peculiarities of the (U,Zr,Fe,O) system. Another phenomenon of highest importance is the mechanical behaviour of the thin ablated vessel wall, where elasticity, plasticity and creep all play a role. It is concluded that codes should include improved models for those phenomena in order to be able to provide a reasonable assessment of potential for IVR.

Main outcomes from the IVR code benchmark performed in the European IVMR project

In-Vessel Retention (IVR) of corium is one of the possible strategies for Severe Accident (SA) mitigation. Its main advantage lies in the fact that, by maintaining the corium within the vessel, it preserves the last containment barrier from corium aggression. One of the issues for the demonstration of the success of this strategy is the evaluation of the behaviour of the corium relocated in the lower head and how it stabilizes and affects the integrity of the vessel wall. The first modelling was developed in the nineties and assessed the heat transfers in a stratified corium pool with a top metal layer made of steel and Zirconium only. About 10 years later, the results of the MASCA program highlighted the possibility of having more complex stratified configurations, including a dense metal layer. Addressing thermochemical effects in the stratification makes the modelling of the corium pool in the lower head more difficult and, in addition, knowledge of the associated kinetics is still limited. As a consequence, available SA codes, either integral or dedicated to the lower head, can differ significantly in their models, which leads to discrepancies in the results when evaluating the IVR strategy.In order to identify the main modelling issues and to assess the capabilities of the codes, a benchmark exercise for code validation was made in the scope of the European H2020 project IVMR (In-Vessel Melt Retention). It is based on the definition of different IVR configurations at reactor scale with an increase in the complexity of the phenomena involved: starting from a steady-state stratified pool with metal on top, up to consideration of corium phase separation at thermochemical equilibrium and progressive ablation of metallic structures and vessel wall. As a last step, the more challenging transient configuration, with formation of a metal layer heavier than the oxide followed later on by stratification inversion, was also studied. It should be noted that mechanical resistance of the ablated vessel wall and cooling conditions out of the RPV in the cavity are not evaluated and considered out of the scope of this benchmark focused on the thermal load transferred from the corium pool through the vessel wall.Six organisations took part in this benchmark (CEA, EDF, GRS, IBRAE, IRSN, NRC-KI) and 6 different codes were used (ASTEC, ATHLET-CD, MAAP_EDF, PROCOR, HEFEST_URAN and HEFEST – stand-alone version of the corresponding module of the SOCRAT code). The main results and outcomes obtained are presented and discussed in this paper. Sensitivity studies have also been performed and allow obtaining a more consolidated range of results.Thanks to this benchmark exercise and to the approach followed with a progressive increase of complexity, the capabilities of codes to evaluate the heat flux profile applied to the vessel wall in steady-state are demonstrated. Then, for transient configurations, it is shown that different modelling approaches give rather consistent results since dispersion remains limited between code predictions, even in the most challenging configuration with stratification inversion: ±40% for the minimum vessel thickness (2.5 cm ± 1 cm). In addition, all codes predict that transient effects lead to more vessel ablation than in the final steady-state. Hence, the importance of the consideration of the progressive molten steel incorporation in the pool and of the consideration of thermochemical equilibrium to calculate the oxide and metal phases composition was highlighted in the benchmark.Regarding the dispersion of the results obtained, analysis shows that uncertainties on metal layer properties have a significant impact. In addition, differences in modelling assumptions were identified and discussed. The main issues are related to the following physical phenomena: (i) The interaction of the oxide crust with the molten steel; (ii) The kinetics of stratification inversion; (iii) The heat transfer in thin metal layer.

Numerical Study on the Corium Pool Heat Transfer With OpenFOAM

In order to ensure the successful implementation of in-vessel retention strategy to terminate molten corium pool evolution, it is necessary to evaluate the heat flux distribution on the lower head wall of reactor pressure vessel. In the present study, the models of internal heating and natural convention buoyancy, as well as the models of WMLES turbulence and phase changing were applied in the open source CFD software OpenFOAM to perform numerical simulations for the COPRA single-layer molten pool experiment. The distributions of temperature, heat flux and crust thickness were obtained. The simulation results were in good comparison with COPRA experimental data, proving the validity of the developed model for corium pool heat transfer characteristics. The simulation method and results could be applied to further in-depth study of thermal behavior in the corium pool.

Study of online measurements techniques of metallic phase spatial distribution into a corium pool

In the context of in-vessel retention (IVR) strategy in order to better assess the risk of reactor vessel failure, the knowledge related to the kinetics of immiscible liquid phases stratification phenomenon needs to be further improved. So far, only one medium-scale experiment (MASCA-RCW, in the frame of the OECD MASCA program) gives direct information regarding the transient relocation of metal below the oxide phase through post-mortem measurements. No experimental characterization of the stratification inversion kinetics when heavy metal becomes lighter and relocates at the top exists. Further investigation of these hydrodynamic and thermochemical processes could be made possible thanks to on line instrumentation enabling to follow displacement of oxidic and metallic phases into the corium pool. At CEA Cadarache, studies are under progress to set up innovative technologies for corium stratification monitoring which would be integrated to a cold crucible induction melting furnace. Based on space and time resolution specifications, three on-line measurements techniques were selected and studied. The first one is an ultrasonic technique using a refractory material waveguide and based on a time-of-flight measurement. We present the feasibility approach with the preliminary results obtained during experiments at high temperature on VITI facility. The second method consists in electromagnetic characterization of the corium pool thanks to an excitation by a magnetic field induced by surroundings coils and measurement of magnetic response by sensors placed around the crucible. A modelling study has enabled to define an appropriate experimental configuration. An experimental set up has also been tested to verify the calculation results. The third technique is 2D X-rays imaging. A feasibility study for a real-time X-ray imagingwith a framerate of 1 image/s has been performed using home-made simulation software MODHERATO, accounting forscattering, based on corium behavior previsions. Results on thedetection of interfaces between different type of corium phases(oxide, light metal, heavy metal) are shown.

Convection of internal-heated two-layer fluids with deformable interface under swinging motion conditions

Convection inside a layered system has been a significant research focus. Different from the known research under static condition, this study experimentally investigated the convection phenomena of internal-heated two-layer fluids systems under swinging motion conditions. The effects of swinging motions on the major convection phenomena related parameters were studied, including the system temperature fields and heat transfer capacities, together with the sensitivity analyses of maximum swinging angle and swinging period. Additionally, phenomenological research was carried out on the pattern of interface deformation between the two layers under swinging motions. From the experiment, it’s found that swinging motions can weaken the thermal stratification occurring in the internal-heated two-layer system under static condition, and intensify heat transfer from the system to the outer cooling boundary. A wave shape of interface is observed with the declining swinging period, and unstable interface deformations can be seen during swinging motions with a large maximum swinging angle but a small swinging period, with mixing or penetration observed between the two liquids. Moreover, a thicker layer of upper fluid refrains interface deformation and thus helps to stabilize two-layer system during swinging motions. This study can be helpful for the nuclear safety management strategy related to the analysis of stratified corium pools in the marine environment.

Experimental research on heat transfer behavior of large scale two-layer salt melt pool based on COPRA facility

During a core meltdown accident, the significant decay heat within the corium pool will continually impose thermal load on the vessel wall and challenge the integrity of reactor pressure vessel (RPV). External Reactor Vessel Cooling (ERVC) is considered to be an effective strategy to reduce the potential of vessel failure. After long-term cooling, the initially uniform melt pool can reach a final steady state with stable two-layer or three-layer configuration. Based on the COPRA facility, a series of tests were performed to study the natural convection heat transfer phenomenon of stratified corium pools. The test section was a full-scale 2D slice, simulating the hemispherical lower head of the HPR1000 reactor design. Direct electrical heating was applied in the lower part of vessel to simulate the homogenously heated ceramic pool, while no heat was generated in the upper part to model the metallic layer. A binary mixture of 20 mol% NaNO3–80 mol% KNO3 was selected as simulant material due to the similar phase-change characteristics with the prototypical melt. During the experiment, effects of heating power, top boundary condition, corium pool height, decay heat distribution and simulant material on corium pool heat transfer behavior were analyzed. The results implied that the top cooling and decay heat distribution slightly influenced the temperature and heat flux of ceramic pool, but strongly impacted the characteristics of metallic layer. The comparison between the dimensionless parameters of water tests and salt tests also demonstrated the key role of simulant material. Furthermore, the downward heat transfer Nudn and the upward heat transfer Nuup obtained from COPRA two-layer experiment were in accordance with the previous data or empirical correlations. Note that using the same simulant material for oxide layer and metal layer fails to repeat the boundary conditions in the reactor case, however, the current study pays more attention to the qualitative analysis of different influence on natural convection heat transfer behavior. More valuable information about the crust characteristics and stratification effect can be obtained by the following CFD simulation of these tests.

Material characterization of the VULCANO corium concrete interaction test with concrete representative of Fukushima Daiichi Nuclear Plants

In the framework of JAEA-CEA collaboration, experimental studies have been conducted for estimating the material characteristics of corium debris representative of the Fukushima Daiichi nuclear damaged plants. A test has been performed in the VULCANO facility in CEA Cadarache to simulate the concrete corium interaction (CCI) with prototypic corium (using depleted uranium) and concrete of Fukushima Daiichi 1F1 Nuclear Plants. This paper presents the Post Test Analyses on 9 samples representative of the CCI during this test: in the corium pool, in the crusts and at the vertical and horizontal interfaces with the concrete. Analyses have been performed by SEM/EDS, X-Ray Diffraction, complete dissolution and ICP, micro-hardness measurements of the main phases. The pool is very porous, its composition is homogeneous but metallic blocks have been observed at the interfaces. The major phases encountered are uranium rich and zirconium rich oxides forming nodules from micrometers to millimeters size, chromium-iron rich precipitates of several micrometers, metallic Fe–Ni droplets and chromium-silicon rich filaments in a matrix, likely vitreous, rich in concrete elements: Si, Al, Ca, but containing up to 12 cations. The matrix is the softer oxide phase, when the Cr rich precipitates are the harder. The analyses are consistent with the estimated macroscopic ablation ratio, but do not still explain the important axial ablation observed for this specific basaltic concrete. The different phases formation, distribution and solidification path are discussed. First comparisons are proposed with the former CCI tests with European concretes. These results give helpful insights for the future dismantling of the plant and for a deeper understanding of the CCI process for basaltic concrete.