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International audience

International audience

International audience; No individual country has sufficient resources to address all severe accident important phenomena within the framework of a national research programme, therefore optimised use of resources and the collaboration at European and international level are very important. Integrating European severe accident research facilities into a pan-European laboratory for severe accident and corium studies and providing resources to other European partners for better understanding of possible accident scenarios and phenomena is necessary in order to improve safety of existing and, in the long-term, of future reactors.SAFEST (Severe Accident Facilities for European Safety Targets) is a European project networking the European corium experimental laboratories with the objective to establish coordination activities, enabling the development of a common vision and of research roadmaps for the next years, and of the management structure to achieve these goals. One of the main objectives is to address and resolve the variety of the remaining severe accident issues related to accident analysis and corium behaviour. The project is a valuable asset for the fulfilment of the severe accident RetD programmes that are being set up after the Fukushima Daiichi accidents and the subsequent European stress tests, addressing both national and European objectives.Roadmaps on European severe accident experimental research for water reactors and for GenIV technologies will be drafted. Joint RetD is conducted to improve the excellence of the SAFEST facilities this includes measurement of corium physical properties, improvement of instrumentation, consensus on scaling law rationales and cross comparison of material analyses.Joint experimental research is a clear objective in the SAFEST project to provide solutions for stabilisation of severe accident and termination of consequences for the current Gen II and III plants. Consequently, the knowledge obtained in SAFEST shall lead to improved severe accident management measures, which are essential for reactor safety. In addition, it will offer competitive advantages for the nuclear industry and contribute to the long-term sustainability of nuclear energy.

International audience; The introduction of (U,Am)O2±x materials in the Fast Neutron Reactors requires a detailed knowledge of their physico-chemical properties to predict their behaviour under both normal and off-normal conditions. The melting temperature (Tm) is among the fundamental parameters to determine the safety limits of a fuel. We performed Tm measurements under both inert atmosphere (argon) and under air on (U1-y,Amy)O2±x samples with 0.075≤y≤0.7 using a laser-heating coupled with fast pyrometry technique [1]. For all the compositions, the Tm measurements under argon were reproducible. An almost linear decrease from 3080 K to 2900 K as a function of the Am/(Am+U) ratio was found. A different behaviour was observed under air. For the samples with 0.075≤y

International audience

International audience; During a severe accident in a nuclear reactor, the nuclear fuel (UO$_2$ or MOx) may interact with the Zircaloy cladding and with the metallic structures inside the vessel (spacer grids, neutronic absorbers cladding) forming a solid/liquid mixture called corium. At this stage of the accident, the temperature inside the damaged reactor can exceed 2700K. In this framework, the knowledge of the high temperature properties of corium are paramount for the comprehension of how possible accidental scenarios proceed. In particular, the U-(Pu)-Zr-Fe-O system has been considered in the present study, as a simplified version of the real corium. The aim of this study has been the acquisition of high temperature data on some key sub-systems of corium. This paper is focused on the results obtained on the U-Zr-O and Fe-Zr-O ternary systems. Liquidus and solidus temperatures have been obtained by using the laser heating setup at JRC-ITU.

International audience; Severe accident with core meltdown is a threat to the containment integrity. As Chernobyl and Fukushima accidents demonstrate, significant release of radioactive products into the environment can cause severe consequences to health, the environment and the economy. Corium is the molten material formed after meltdown of a nuclear reactor core during a severe accident. In order to improve the understanding and modelling of corium behavior, experiments are needed both for LWRs and GenIV fast reactors. As it is not feasible to simulate all the aspects of corium phenomenology, especially its high temperature physico-chemistry, there is the need for prototypic corium experiments with depleted uranium, even though it implies a series of protective and regulatory measures. Within the European SAFEST consortium on severe accidents and corium experimental research a number of facilities operate with prototypic corium KROTOS, MERELAVA,VITI, VULCANO, FLF, COMETA, SICOPS and CODEX,. Many technological challenges arise in heating, measuring and containing such high temperature material. The limited availability of thermodynamic and thermophysical properties for these materials is also a difficulty whereas it must be stressed that there is not one unique corium composition but a large spectrum of compositions depending on accident scenarios.Nevertheless, prototypic material experimental research has already provided numerous insights to the understanding of corium behavior, such, as among others, data on fuel element and melt oxidation processes and kinetics, material effects affecting steam explosion or molten core concrete interaction, efficiency of corium spreading, In the frame of the SAFEST European project, current experimental capabilities are being upgraded, thanks to improvements in high temperature metrology, corium material analyses, measurement of corium thermophysical properties such as electrical properties as well as insights on scaling laws.New prototypic corium experiments are planned in Europe such as PLINIUS2 (CEA).

International audience; During a severe accident in a PWR reactor, the oxide nuclear fuel (UO2 or MOX) reacts at high temperature with the zirconium alloy clad and the steel vessel to form a partially or fully molten mixture so-called in-vessel corium . In such a case, the corium forms a pool in the bottom of the vessel, constituted of two liquid phases, metallic and oxide. The formation of such a configuration of the corium pool is due to the existence of a miscibility gap in the liquid state. The fractions and compositions of these liquid phases have to be well known in order to model the thermal hydraulic properties of the pool. The aim of the present work is to investigate experimentally the thermodynamic properties of the U-Zr-O-Fe system representative for the in-vessel corium to improve the thermodynamic modelling using the Calphad method. Laser heating techniques as well as heat treatments are used to measure solid/liquid transition temperatures and to highlight the miscibility gap in the liquid state. Experimental data are used to assess the thermodynamic properties of this key system. Both experimental and theoretical results will be presented.

International audience; Severe accident with core meltdown is a threat to the containment integrity. As Chernobyl and Fukushima accidents demonstrate, significant release of radioactive products into the environment can have severe consequences both for people's health and the country's economy. Severe accidents are the focus of considerable research involving substantial human and financial resources worldwide. The research field encompasses many challenging phenomena, complicated by high temperatures and presence of radioactive materials. No individual country has sufficient resources to address all important phenomena within the framework of a national research programme, therefore optimised use of resources and the collaboration at European and international level is very important. One of the main objectives of the SAFEST project of the 7th EU framework programme is integrating European severe accident research facilities into a pan-European laboratory for study of corium behaviour in severe accidents. The resources of this laboratory will be provided to other interested European partners for better understanding of possible accident scenarios and phenomena in order to improve safety of existing and, in the long-term, of future reactors. The SAFEST consortium will be able to address several severe accident issues related to accident analysis and corium behaviour. It will be a valuable asset for the fulfilment of the severe accident RandD programmes that are being set up after Fukushima and the subsequent European stress tests, addressing both national and European objectives.

International audience

International audience; A description of a series of laser heating experiments on MOX with various Pu content under Air. The aim was to determine the melting temeperature dependence on Pu content in the oxidizing atmosphere.

International audience; During a severe accident in a nuclear reactor, extreme temperatures may be reached (T > 2500 K). In these conditions, the nuclear fuel may react with the Zircaloy cladding and then with the steel vessel, forming a mixture of solid-liquid phases called in-vessel corium. In the worst scenario, this mixture may pierce the vessel and reach the concrete underneath the reactor. In the framework of the development of the TAF-ID thermodynamic database (www.oecd-nea.org/science/taf-id) on nuclear fuel and to predict the high temperature behaviour of the corium+concrete system, new high temperature thermodynamic data are needed. The LM2T at CEA Saclay centre started an experimental campaign of measure of phase equilibria at high temperature (up to 2400 K) on interesting corium sub-systems. Furthermore a new laser heating setup has been conceived. This technique allows very high temperature measures (T>3000 K) limiting the interactions between the sample and the surroundings. New phase equilibria data on the U-Zr-O system and on the U-Zr-Al-Ca-Si-O system will be presented. Moreover original melting temperature data on the PuO2-UO2-ZrO2 system obtained in collaboration with the JRC-ITU, will be also shown.

International audience; During a severe accident in a pressurized water reactor, very high temperatures can be reached (T>2500 K). In these conditions, the UO$_2$ or mixed oxide (U,Pu)O$_2$ nuclear fuel shall react with the Zircaloy cladding and then with the steel vessel, forming a complex mixture of solid and liquid phases called in-vessel corium. At very high temperature, the corium can be constituted of two metallic and oxide liquid phases due to the existence of a miscibility gap in the liquid state. The relative fraction and composition of these phases that form in the corium as a function of temperature are key input data for severe accident codes to predict corium thermal-hydraulics properties.The TAF-ID thermodynamic database [1] coupled to software like Thermo-Calc or Open Calphad is a powerful computational tool to predict the thermodynamic data of nuclear fuel materials. The database is developed using the Calphad method on the basis of the available experimental data. In case of severe accident applications, experimental data are missing, especially on systems containing plutonium. New measurements of thermodynamic and phase diagram data on subsystems of the prototypic in-vessel corium U-Pu-Zr-Fe-O are required. Such experiments are particularly difficult to carry out due to radiotoxicity of the materials and to the very high temperature level. New measurements using advanced laser melting techniques will be presented. The models derived from these new experimental data will be described. Finally the influence of the presence of plutonium in the mixed oxide fuel on the thermodynamic properties and phase equilibria of the in-vessel corium will be investigated by thermodynamic calculations.