Materials for Sustainable Nuclear Energy - The Strategic Research Agenda (SRA) of the Joint Programme on Nuclear Materials (JPNM) of the European Energy Research Alliance (EERA)

This Strategic Research Agenda (SRA) has been prepared by the EERA-JPNM, based on a wide consultation with the scientific and industrial community involved, to identify the research lines to be pursued in the EU to ensure that suitable structural and fuel materials are available for the design, licensing, construction and safe long-term operation of GenIV nuclear systems. Three Grand Challenges have been identified, namely: (i) Elaborate design correlations, assessment and test procedures for the structural and fuel materials that have been selected for the demonstrators under the service conditions expected; (ii) Develop physical models coupled to advanced microstructural characterization to achieve high-level understanding and predictive capability; (iii) Develop innovative materials solutions and fabrication processes of industrial application to achieve superior materials properties, to increase safety and improve efficiency and economy. For structural materials, the requirement of 60 years design lifetime for non-replaceable components is in perspective the most demanding requirement, which includes under its umbrella several issues related with the reasonable prediction of long-term degradation processes: high temperature processes (creep, fatigue, thermal ageing), compatibility with –especially- heavy liquid metal and helium coolants, and effects of low flux prolonged irradiation, with emphasis on welded components in all cases. In terms of testing, there is a need for standardization, especially for sub-size and miniature specimens. The modelling, supported by microstructural characterization, has as its main objective the development of suitable microstructure evolution models to be used as input to models for the mechanical behaviour under irradiation and at high temperature, eventually linking with fracture mechanics. Specific developments are required for coolant compatibility models, as well as for models in support of the use of charged particle irradiation for the screening of new materials solutions, such as those listed above. Concerning fuel materials, the properties and processes that govern its behaviour in pile, on which research effort is focused, are: margin to melting (establishment of phase diagrams and evolution of thermal properties), atomic transport properties and ensuing microstructural evolution, fission product (non-gaseous) and helium (gas) behaviour and transport, mechanical properties (their evolution, subsequent fragmentation and cracking, fuel-cladding mechanical interaction), and compatibility with cladding and coolant (internal cladding corrosion, chemical interactions especially in case of severe accident). These are all addressed from both an experimental and a modelling perspective. Besides the obvious need of adequate financial resources in order to address the research problems outlined in this SRA, as well as the necessary corollaries, four recommendations emerge that this document is intended to bring to the attention of stake-holders, particularly decision-makers: R1: Data from materials property measurements after exposure to relevant conditions are the essential ingredient for robust design curves and rules. Plenty of data were produced in the past that are now de facto unusable; this is either because they are covered by confidentiality or because they were not properly archived. Correct data management to guarantee availability for future re-assessment is therefore essential and should be encouraged and fostered. In particular, financially supported policies to foster data sharing and encourage old data disclosure should be implemented. R2: Some infrastructures are absolutely essential to enable the correct qualification of nuclear materials, not only irradiation facilities, but also suitable ‘hot’ cells where active materials can be safely handled and tested, nuclearized characterization techniques, loops and pools for compatibility experiments, etc. They are also crucial for education and training of young researchers and operators. These infrastructures are costly to build and maintain. Other research facilities are, on the other hand, more common and sometimes redundant. A rational and harmonised, pan-European management of infrastructures, based on joint programming, including trans-national infrastructure renewal planning and a scheme for facility sharing and exploitation, would be highly desirable and, at the end of the day, beneficial for all. R3: International cooperation with non-EU countries where research on nuclear materials is pursued can be very valuable for Europe. Quite clearly, the goals of this cooperation are in the end the same as in the case of internal European cooperation, namely coordination of activities, sharing of data, and access to infrastructures. Currently, however, the instruments available in Europe for international cooperation are not sufficiently attractive to motivate significant cooperation with non-EU researchers. Efforts should be made to improve their attractiveness and ease of access. International organisations such as OECD.NEA, IAEA, but also Euratom and JRC for the connection with GIF, have here a crucial role. R4: The nuclear materials research community in Europe is currently strongly integrated and engaged in thriving collaboration, in a bottom-up sense. This is in contrast with the inadequacy of the top-down instruments offered to make this integration efficient and functional. This SRA is largely the result of matching bottom-up research proposals with top-down strategies. The appropriate instrument to allow this community to deliver according to the SRA goals should provide the conditions to implement the agreed research agenda and to set up suitable E&T&M schemes that allow knowledge, data, and facility sharing. Since the financial support of Euratom will never be sufficient, earmarked funding from the MS dedicated to support integrated research on nuclear materials is crucial. In this sense, a co-fund instrument, such as a European Joint Programme, seems to be most suitable.JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

Intra granular precipitation and grain boundary segregation under neutron irradiation in a low purity Fe–Cr based alloy

International audienceA nanoscale description of intra- and inter-granular segregation and precipitation in a Fe–12 at.%Cr model alloy of low purity after neutron irradiation at 300 °C up to 0.6 dpa, has been performed owing to Atom Probe Tomography (APT). Two different populations of clusters have been observed inside the grains: Cr-enriched and NiSiPCr-enriched clusters. As expected with a process of enhanced precipitation, Cr-enriched clusters are homogeneously distributed inside grains. The NiSiPCr-enriched clusters, which are probably formed by radiation induced segregations, are independent of the Cr-enriched clusters. Investigation of a low angle grain boundary has revealed Si, Cr and P segregation on the dislocation network

Intra granular precipitation and grain boundary segregation under neutron irradiation in a low purity Fe–Cr based alloy

International audienceA nanoscale description of intra- and inter-granular segregation and precipitation in a Fe–12 at.%Cr model alloy of low purity after neutron irradiation at 300 °C up to 0.6 dpa, has been performed owing to Atom Probe Tomography (APT). Two different populations of clusters have been observed inside the grains: Cr-enriched and NiSiPCr-enriched clusters. As expected with a process of enhanced precipitation, Cr-enriched clusters are homogeneously distributed inside grains. The NiSiPCr-enriched clusters, which are probably formed by radiation induced segregations, are independent of the Cr-enriched clusters. Investigation of a low angle grain boundary has revealed Si, Cr and P segregation on the dislocation network

Cr precipitation in neutron irradiated industrial purity Fe–Cr model alloys

International audienceThe microstructure of four neutron irradiated Fe-Cr model alloys of industrial purity (Fe-2.5%Cr, Fe-5%Cr, Fe-9%Cr and Fe-12%Cr) has been characterized by atom probe tomography (APT). Irradiation has been performed at 300 degrees C up to 0.6 dpa in MTR reactor. APT investigations confirmed the enhanced precipitation of alpha' clusters as these clusters have only been observed in supersaturated model alloys. In addition a nonexpected family of clusters has been revealed due to irradiation induced segregation of impurities: NiSiPCr-enriched clusters. They might be associated to defect clusters invisible by transmission electron microscopy (TEM). A quantitative description of these objects is presented in this paper and results are compared with TEM and SANS data of the literature obtained on the same model alloy

Atomic scale analysis and phase separation understanding in a thermally aged Fe–20at.%Cr alloy

International audienceFe–Cr model alloys are of interest for understanding of phase separation in structural material for fusion or fission reactors. This motivated the quantitative study of the phase separation in a thermally aged (773 K) Fe–20 at.%Cr alloy using a Tomographic Atom Probe. It is shown that the chromium content in the α′ phase evolves with ageing time synonymous of a non-classical nucleation mechanism. The overlap of the nucleation and coarsening regimes is observed. A non-steady coarsening regime occurs before 1067 h of ageing. The solubility limit at 773 K is found to be 14 at.%. A maximum concentration of (83 ± 1) at.% is observed for the Cr-concentration in α′ precipitates

Atomic scale analysis and phase separation understanding in a thermally aged Fe–20at.%Cr alloy

International audienceFe–Cr model alloys are of interest for understanding of phase separation in structural material for fusion or fission reactors. This motivated the quantitative study of the phase separation in a thermally aged (773 K) Fe–20 at.%Cr alloy using a Tomographic Atom Probe. It is shown that the chromium content in the α′ phase evolves with ageing time synonymous of a non-classical nucleation mechanism. The overlap of the nucleation and coarsening regimes is observed. A non-steady coarsening regime occurs before 1067 h of ageing. The solubility limit at 773 K is found to be 14 at.%. A maximum concentration of (83 ± 1) at.% is observed for the Cr-concentration in α′ precipitates

Cr precipitation in neutron irradiated industrial purity Fe–Cr model alloys

International audienceThe microstructure of four neutron irradiated Fe-Cr model alloys of industrial purity (Fe-2.5%Cr, Fe-5%Cr, Fe-9%Cr and Fe-12%Cr) has been characterized by atom probe tomography (APT). Irradiation has been performed at 300 degrees C up to 0.6 dpa in MTR reactor. APT investigations confirmed the enhanced precipitation of alpha' clusters as these clusters have only been observed in supersaturated model alloys. In addition a nonexpected family of clusters has been revealed due to irradiation induced segregation of impurities: NiSiPCr-enriched clusters. They might be associated to defect clusters invisible by transmission electron microscopy (TEM). A quantitative description of these objects is presented in this paper and results are compared with TEM and SANS data of the literature obtained on the same model alloy

Mécanismes de vieillissement à très longue échéance des aciers inoxydables austénoferritiques

Comprendre l'origine de la fragilisation des aciers austéno-ferritiques utilisés dans les coudes moulés des circuits primaires des centrales nucléaires est une étape clé pour l'anticipation de leur vieillissement. Cette prédiction nécessite une caractérisation et une compréhension du mécanisme de transformation de phase à l'origine de ce constat : la décomposition de la ferrite. Ainsi, de façon duale, des ferrites d'aciers vieillis plus de 20 ans, sur site ou en laboratoire ainsi qu'à différentes températures, ont été analysées par sonde atomique tomographique et un travail de simulation de la décomposition de la ferrite d'alliages modèles Fe-Cr a été initié. Afin de valider les paramètres utilisés en simulation Monte Carlo, une étude expérimentale de la démixtion d'un alliage Fe 20 % at. Cr vieilli à 500C a été réalisée. Cette étude expérimentale a montré qu'un régime de germination non classique (GNC) intervient dans cet alliage. La simulation de la décomposition de la ferrite dans le même alliage, vieilli à la même température, n'a pas révélé l'enrichissement progressif des précipités de phase a' caractéristique de la GNC. L'étude d'aciers vieillis plus de 20 ans a permis de confirmer que les aciers vieillis en laboratoire sont représentatifs de ceux vieillis sur site (pour T < 350C), que la phase G (précipitation intermétallique à l'interface des phases a/a') n'influence pas la fragilisation de la ferrite et que la différence de traitement thermomécanique n'est pas déterminante quant à l'écart de décomposition observé dans ces aciers.Embrittlement study of duplex stainless steels is a very important in order to predict the lifetime of primary circuits of nuclear power plant. Ferrite steels aged over 20 years, on-site, in laboratory and at different temperatures was analyzed by tomographic probe atom to assess the trend of aging of these materials with very long times. A more prospective work was also carried out, the aim was to model the decomposition of ferrite from austenitic-ferritic steels. The simulation of the decomposition of these steels are very complex, we initiated preliminary work in modelling the Fe-Cr alloys, because the decomposition of Fe and Cr in these steels is the main cause of their fragility. To validate the parameters used in simulation, an experimental study of the decomposition of an alloy Fe-20% at. Cr aged at 500 C was performed. This experimental study has shown that a non-classical germination (NCG) is involved in this alloy. The performed simulations on the same alloy at the same temperature, did not reproduce the progressive enrichment of precipitated phase a' (characteristic of NCG). The study of steels, aged over 20 years, has confirmed that the steel aged in laboratory are representative to steel aged in site ( T <350 C). Moreover, it has been shown that the Gphase (intermetallic precipitation at the interface a/a' phases) does not influence the embrittlement of the ferrite and the difference of thermo-mechanical treatment is not determinant of the variance decomposition observed in these steels.ROUEN-BU Sciences Madrillet (765752101) / SudocSudocFranceF