Effect of statistically stored dislocations in tungsten on the irradiation induced nano-hardening analyzed by different methods

Tungsten self-ion irradiation was performed at 800 °C up to 0.01-1 dpa on two different W grades with essentially different dislocation density. Nanoindentation was applied to characterize the radiation hardening in two W grades with different microstructure. Different methods to analyze the indentation curves were applied to extract the bulk equivalent radiation hardening. It was shown that depending on the applied method, different outcomes may occur. The most satisfactory procedure was established and a consistent set of parameters was found. The bulk equivalent radiation hardening was found to saturate above 0.1 dpa. The characteristic distance between irradiation induced defects acting as dislocation pinning points was found to decrease up to 0.1 dpa, and then saturate/increase with irradiation dose. No essential difference in radiation hardening was observed between the studied W grades with essentially different initial dislocation density

Etude par IBA de la réponse de systèmes multicouches Cr-Ta aux effets d'irradiation aux ions lourds et à l'implantation d'hélium

International audienceLe développement de l'industrie de l'énergie nucléaire dépend du développement de nouveaux matériaux capables de supporter une forte dose d'irradiation (plusieurs centaines de déplacements par atome) aux neutrons de grande énergie et une forte accumulation des gaz de transmutation (He et/ou H). Les revêtements nanostructurés apportent un moyen de diminuer l'impact de l'irradiation, d'une part en évacuant par les nombreuses interfaces (joints de grains, interphases) les défauts créés lors de l'irradiation, mais aussi en limitant l'apparition de cavités générées habituellement lors de la nucléation et la percolation des bulles d'hélium. Les systèmes multicouches permettent d'avoir une grande densité contrôlée d'interfaces ce qui permet de réduire la distance de diffusion des défauts vers les puits d'élimination et limite considérablement leur accumulation. Les études faites sur ces systèmes montrent une grande tolérance à l'implantation d'hélium, les systèmes les plus performants souvent faits d'éléments immiscibles peuvent accommoder l'hélium à des concentrations de plus de 20 % at. Cependant, leur réponse à l'irradiation aux ions lourds, représentatifs des effets des neutrons en réacteur, est très peu étudiée. Dans cette étude, nous avons combiné les techniques d'analyse par faisceaux d'ion (RBS et NRA) à la Microscopie Electronique en Transmission (MET) pour étudier le comportement de systèmes multicouches Cr-Ta sous irradiation aux ions lourds et sous implantation d'hélium (implanteur JANNuS-Orsay). L'irradiation aux ions lourds crée des couches de mixing aux interfaces Cr-Ta (Fig. 1), mais le caractère multicouche est préservé même aux très fortes doses d'irradiation (220 dpa dans le tantale). Le mixing est moins important dans les systèmes de plus faible épaisseur. L'application du modèle de la pointe thermique indique une efficacité d'élimination des défauts de 70-80 % dans le système aux couches de 15 nm d'épaisseur. Par ailleurs, ce système peut contenir de l'hélium à des concentrations de 20 % at. avec une diffusion post-implantation très limitée

Etude par IBA de la réponse de systèmes multicouches Cr-Ta aux effets d'irradiation aux ions lourds et à l'implantation d'hélium

International audienceLe développement de l'industrie de l'énergie nucléaire dépend du développement de nouveaux matériaux capables de supporter une forte dose d'irradiation (plusieurs centaines de déplacements par atome) aux neutrons de grande énergie et une forte accumulation des gaz de transmutation (He et/ou H). Les revêtements nanostructurés apportent un moyen de diminuer l'impact de l'irradiation, d'une part en évacuant par les nombreuses interfaces (joints de grains, interphases) les défauts créés lors de l'irradiation, mais aussi en limitant l'apparition de cavités générées habituellement lors de la nucléation et la percolation des bulles d'hélium. Les systèmes multicouches permettent d'avoir une grande densité contrôlée d'interfaces ce qui permet de réduire la distance de diffusion des défauts vers les puits d'élimination et limite considérablement leur accumulation. Les études faites sur ces systèmes montrent une grande tolérance à l'implantation d'hélium, les systèmes les plus performants souvent faits d'éléments immiscibles peuvent accommoder l'hélium à des concentrations de plus de 20 % at. Cependant, leur réponse à l'irradiation aux ions lourds, représentatifs des effets des neutrons en réacteur, est très peu étudiée. Dans cette étude, nous avons combiné les techniques d'analyse par faisceaux d'ion (RBS et NRA) à la Microscopie Electronique en Transmission (MET) pour étudier le comportement de systèmes multicouches Cr-Ta sous irradiation aux ions lourds et sous implantation d'hélium (implanteur JANNuS-Orsay). L'irradiation aux ions lourds crée des couches de mixing aux interfaces Cr-Ta (Fig. 1), mais le caractère multicouche est préservé même aux très fortes doses d'irradiation (220 dpa dans le tantale). Le mixing est moins important dans les systèmes de plus faible épaisseur. L'application du modèle de la pointe thermique indique une efficacité d'élimination des défauts de 70-80 % dans le système aux couches de 15 nm d'épaisseur. Par ailleurs, ce système peut contenir de l'hélium à des concentrations de 20 % at. avec une diffusion post-implantation très limitée

Thermodynamic model for lattice point defect-mediated semi-coherent precipitation in alloys

International audienceThe formation of precipitates with an atomic volume different from their parent phase eventually leads to a loss of the lattice continuity at the matrix/precipitate interface. Here, we show the creation or removal of lattice sites mediated by lattice point defects is an accommodation mechanism of the coherency loss and even a precipitation driving force. We introduce a thermodynamic approach that rationalizes the selection of phases resulting from chemical and crystallographic constraints in relation to point defect properties. The resulting semi-coherent phase diagram and the precipitation kinetic model depend on the equilibrium phase diagram, the eigenstrain of the precipitating phase, and the chemical potential of point defects. From a joint experimental and modeling study, we uncover the prominent role of excess point defects in unforeseen phase transformations of the Fe-Ni metallic system under irradiation. By addressing the fundamental role of lattice point defects in the accommodation mechanisms of precipitation, we provide a step torwards the understanding of semi-coherent phase transformations occurring in solid materials upon synthesis and in use

Drastic influence of micro-alloying on Frank loop nature in Ni and Ni-based model alloys

International audienceNickel and its alloys are f.c.c. model materials to investigate the elementary mechanisms of radiation damage and solute effects. This paper focuses on the drastic influence of micro-alloying (0.4 wt.% Ti or Cr) on the nature of defects after ion irradiation. Ultra-high purity materials are used to avoid impurity effects. For the first time, (i) large stable intrinsic Frank loops are identified in nickel while in alloys extrinsic Frank loops are observed; (ii) an eradication mechanism of intrinsic Frank loops is clearly identified; (iii) the morphology of Frank loops is shown to be characteristic of their nature. IMPACT STATEMENT For the first time, a drastic influence of micro-alloying on the defect nature is shown in irradiated Ni systems. Minor addition of solutes modifies significantly elementary mechanisms of radiation damage

Cr-Ta multilayers as a potential coating material for fuel cladding in Gen III and Gen IV nuclear plants

International audienceThe development of nuclear energy is dependent on the elaboration of advanced materials tolerant to irradiation by high energy neutrons and to gas accumulation (He and/or H transmutation products under neutron irradiation) at high temperature. In this context, research focused on nanostructured materials with multiple interfaces (grain and phase boundaries) acting as (i) defect sinks that contribute to reduce the effects of accumulated radiation damage; (ii) and as traps for implanted species such as helium. Multilayer systems allow well-controlled high density of interfaces (i.e. smalllayer thickness) to reduce the diffusion distance for defects to move from their original location to the nearest sink, thus considerably limiting their transformation into stable aggregates. Previous studies show high tolerance of multilayer systems to helium implantation-the most performant systems can store helium up to 20 at. %. Nevertheless, their behavior under heavy ion irradiation, representative of nuclear reactor neutron irradiation is very less studied

Inversion of dislocation loop nature driven by cluster migration in self-ion irradiated nickel

International audienceNickel is widely used as fcc model material to obtain insight into fundamental mechanisms of radiation damage. This work presents irradiation of high-purity nickel using self-ions at 450°C and fine analysis of dislocation loops in specimens prepared by Focus Ion Beam using Transmission Electron Microscopy. For the first time to our knowledge, a drastic change of both loop nature (vacancy-type in irradiated zones v.s. interstitial-type in unirradiated zones) and loop Burgers vector (1/3 and 1/2 loops v.s. only 1/2 loops) is identified along the implantation direction in irradiated nickel. This change may be attributed to the formation of interstitial crowdions and their long-range 1D migration. A defect-free layer related to the annihilation of vacancy defects by injected atoms is detected. It provides an easy way to identify the injection peak in self-ion irradiated specimens and solid validation for damage calculations

Impact of intragranular misorientation on void swelling and inter-granular cavities after ion irradiation in standard and additive manufacturing 316 L austenitic steels

International audienceAdditive manufacturing (AM) is a promising technology for the design of materials with complex geometries with reduced cost and material waste. In order to be used in the nuclear industry, the capability of AM materials, in term of radiation resistance must be compared with materials elaborated in conventional ways. In this work, the radiation resistance of 316L austenitic stainless steels (ASSs) elaborated by AM is compared to a solution-annealed 316L ASS after irradiation with 5 MeV Fe$^{5+}$ for 3 dpa at 550 °C (873 K). After irradiation, cavities are mainly located near grain boundaries for all studied alloys. Intra-granular cavities are only found in the AM material after heat treatment and are likely to be remaining porosity already present before irradiation. No cavities in intra-granular position are found in the conventional 316L ASS or in the AM material after hot isostatic pressing (HIP) at 1100 °C. It suggests that the void swelling ASSs starts by the formation of cavities at grain boundaries followed by a formation of cavities in intra-granular position, conventionally studied. Loops in the AM material with a hot treatment at 700°C are heterogeneously distributed due to a bimodal distribution of grains in terms of intra-granular misorientation. The intra-granular misorientation drastically reduces the loop density after irradiation. Frank and perfect loops are found to be only interstitial-type loops. The larger cavities' size and the more advanced dislocations network in the AM HIP sample suggests a slightly reduced radiation swelling resistance for the AM material but further investigations at higher irradiation dose have to be done

Free surface impact on radiation damage in pure nickel by in-situ self-ion irradiation: can it be avoided?

International audienceA major issue of in-situ radiation damage studies is related to the influence of free surfaces. Free surfaces in thin foils are indeed strong sinks for radiation-induced defects. Nevertheless, in-situ irradiation is a powerful tool to study real-time microstructural evolution and obtain insight into dynamic mechanisms of radiation damage. Thus, a detailed evaluation of surface effects is essential to validate existing results and provide guideline for future comparative experiments. In this work, nickel is chosen as model material to conduct systematic studies on surface effects due to the high mobility of its self-interstitials. Ultra-high purity Ni thin foils are in-situ irradiated by 2 MeV Ni2+ ions at high temperatures (400-700°C). Microstructural evolution analysis and detailed characterization of dislocation loops are performed in function of specimen thickness.The present work shows: (i) a drastic influence of thickness on the microstructural evolution and irradiated microstructure with the existence of a critical thickness depending on temperature; (ii) a good prediction of an adequate irradiation thickness with a vacancy concentration calculation model; (iii) an impact of free surfaces on the fine distribution of loop Burgers vectors even above the critical thickness; (iv) a first determination of migration energy of vacancies in Ni considering the temperature dependence of the loop-depleted zones; (v) a production bias model showing that a loss of 10% interstitials favors the growth of vacancy loops observed for the first time in Ni at 510°C