Twinning is a displacive deformation mechanism characterized by a continuous deformation of the material. Although widely studied for other industrial materials such as titanium alloys, this inelastic mechanism remains poorly understood and incompletely modeled for complex crystallographic structures. However, due to the reduced number of symmetries in these structures, dislocation slip is insufficient to accommodate deformation in certain loading directions, requiring the activation of twinning. This is the case for tin, which has a tetragonal structure. In particular, twinning contributes significantly to the mechanical response of tin at high strain rates and low temperatures. At intermediate temperatures and strain rates, a competition between dislocation plasticity and twinning plasticity can occur, making it crucial to describe the coupling between these two phenomena. Proposing a better description of this coupling will shed new light on the experimental data available at CEA DAM. The objective of the thesis is to develop a multiscale approach, from molecular dynamics to continuum mechanics, validated by experiments, to converge on a model that describes the behavior of tin over a wide range of temperatures and strain rates.
Solid oxide cells (SOCs) are electrochemical devices operating at high temperature that can directly convert fuel into electricity (fuel cell mode – SOFC) or electricity into fuel (electrolysis mode – SOEC). In recent years, the interest on SOCs has grown significantly thanks to their wide range of technological applications that could offer innovative solutions for the transition toward a renewable energy market. However, despite of all their advantages, the large-scale industrialization of this technology is still hindered by the durability of SOCs. Indeed, the SOCs remain limited by various degradation phenomena including mechanical damage in the electrodes. For instance, the formation of micro-cracks in the so-called ‘hydrogen’ electrode is a major source of degradation. However, the precise mechanism and the full impact of the micro-cracks on the electrode performances are still unknown. By a multi-physic modelling approach, it is proposed in this thesis (i) to simulate the damage in the microstructure of the electrode and (ii) to calculate its impact on the loss of performances. Once the model validated on dedicated experiments, a sensitivity analysis will be conducted to provide relevant guidelines for the manufacturing of improved robust and performant electrodes.
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In a context where material durability is essential for the safety of infrastructures and the promotion of a sustainable energy transition, mastering corrosion phenomena represents a major challenge for key sectors such as decarbonized energy transport through buried pipelines and civil engineering (hydrogen, nuclear, underground infrastructures). The CORPORE project addresses this issue by proposing the development of advanced numerical simulation models to study corrosion in porous media using COMSOL Multiphysics.
The main scientific and technological objective is to establish an integrated multiphysics modeling approach for the electrochemical and transport mechanisms within porous materials: studying the coupled influence of chemistry, pore network properties, and material–environment interactions on the initiation and propagation of corrosion.
This approach will help optimize anticorrosion protection strategies, reduce maintenance costs, and extend the service life of structures. From a state-of-the-art perspective, most current models focus on homogeneous media and compartmentalized approaches. Our project stands out by integrating a multi-scale mechanistic modeling framework combined with the use of archaeological data for long-term validation.
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In France, nuclear power-plants used for electricity production generate high-level long-lived radioactive wastes through spent fuel reprocessing. These wastes are confined within a borosilicate glass matrix, whose structure allows for the incorporation at the atomic scale of a large number of elements, and which displays excellent long-term properties. The industry challenges are leading to changes in the nuclear fuel composition, which can thus induce a modification of the spent-fuel composition to be vitrified.
Chromium is as such an element of interest: its relatively low solubility in borosilicate glasses as well as its tendency to crystallise with other elements, such as iron, nickel and zinc, needs to be further investigated. This thesis aims to study the synergetic effect of Cr, Ni, Fe and Zn on crystallisation in simplified peralkaline glasses of nuclear interest in order to better comprehend affinities between these elements, thus identifying both the nature and quantity of the several crystalline phases which may form.
The PhD student will benefit from the recognised skills of the host laboratory in glass formulation as well as the study of their physico-chemical properties. All of the resources made available will enable a global approach to the subject, working on a fast-growing topic with major industrial and societal implications. The experience acquired during this interdisciplinary work will be useful in the field of materials.
The CEA and the CNRS have launched an initiative to design a new neutron source using low-energy proton accelerators, the ICONE project [1]. The goal is to build a facility that will provide an instrumental suite of about ten spectrometers available to the French and European scientific community. Alongside ICONE, the LLB is also participating in HiCANS R&D work on the construction of a platform in Bilbao to facilitate European collaborations.Neutron scattering experiments require thermal and cold neutrons. The design of the moderator is therefore a crucial component of the project to maximize the source's performance.
One avenue for improving the moderator performances is to enhance the efficiency of the reflector, and more specifically, the cold neutron reflector. In this study, we propose to investigate the specific scattering properties of cold neutrons on nanostructured materials. Indeed, cold neutrons have long wavelengths (> 0.4 nm) and can therefore be coherently scattered by nanostructured materials. Scattering efficiency is not only amplified by coherent scattering effects, but it is potentially possible to direct this scattering if the reflecting material is anisotropic. This control over the scattering direction can further increase the moderator's brightness.
The first part of the work will consist of identifying the most promising nanostructured materials and modeling their cold neutron reflectivity performance. In a second step, these materials will be shaped and their properties characterized using neutron scattering devices at neutron scattering facilities such as the ILL in Grenoble or the PSI in Switzerland.
Nuclear fuel cladding made of zirconium alloys constitute the first safety barrier in pressurized water reactors. The microstructure of these alloys not only controls mechanical properties, but also phenomenon such as corrosion or growth under irradiation. Enabling a more flexible use of nuclear energy in the mix while maintaining the structural integrity of fuel cladding under both operating and accidental conditions, we must understand the detailed mechanisms of microstructure evolution under irradiation. Numerous studies point toward the center part played by Nb in such microstructural evolution. For instance, diffusion flux coupling between solutes (Nb) and point defect created by irradiation gives rise to local Nb segregation, as well as precipitates which are not seen in non-irradiated samples. Atomic scale modeling brings in information that complements that obtained from experimental observations, allowing to confirm or disprove the evolution scenarios found in the literature. The aim of this Ph.D. work is to use the tools which have been developed to study irradiation effects in ferritic steels, and apply them to Zr alloys, with a focus on radiation induced segregation. Electronic structure calculations in the density functional theory approximation will be used to study the interactions between niobium atoms and point defects. From this data, we are able to compute transport coefficients, from which we can discuss quantitatively solute/point defect flux coupling and radiation induced segregation effects.
Zirconium-based alloys are used as fuel cladding material for pressurized water reactors due to their low thermal neutron absorption cross-section, good mechanical strength, and excellent corrosion resistance. However, despite decades of research, the mechanisms governing the evolution of their microstructure and microchemistry under irradiation are still not fully understood. These phenomena strongly influence the in-reactor performance and lifetime of the materials
Neutron irradiation generates displacement cascades in crystalline material, producing large numbers of point defects (vacancies and interstitials) that can cluster and drive atomic redistribution. The high concentration of point defects promotes radiation-induced segregation and precipitation of alloying elements. In Zr1%Nb alloys, irradiation leads to the unexpected formation of high density Nb-rich nanoprecipitates. This phenomenon has significant implications on the macroscopic properties of the material, notably its post-irradiation creep and corrosion behavior in reactors.
This PhD project aims to elucidate the mechanisms responsible for the precipitation of Nb-rich nanoprecipitates under irradiation. A Zr1%Nb alloy will be irradiated with ions at various doses and temperatures, followed by advanced nanoscale characterization using transmission electron microscopy (TEM) and atom probe tomography (APT). These complementary techniques will provide detailed information on the spatial distribution of alloying elements and the nature of point defect clusters at the atomic scale. Based on these results, a comprehensive mechanism for irradiation-induced precipitation will be proposed, and its implications for the macroscopic properties and in-reactor performance of zirconium alloys will be assessed. By improving the fundamental understanding of irradiation-induced microstructural evolution, this research aims to contribute to the development of more radiation-resistant zirconium alloys for nuclear applications.
The cladding of nuclear fuel rods used in Pressurized Water Reactor, made of zirconium alloys, is the first barrier for the confinement of radioactive nuclei. In-reactor, the cladding is subjected to radiation damage resulting in a change of its mechanical properties. After in-reactor use, the fuel rods are transported and stored. During these various steps, the radiation damage is partially annealed, leading to another evolution of the material properties. All these evolutions are still not well understood.
The objective of this PhD work is to better understand the deformation mechanisms and the mechanical behavior of zirconium alloys after irradiation, and after a partial annealing of the radiation damage. This will help to better predict the behavior of the cladding tube after use and thus guaranty the confinement of radioactive nuclei.
In order to achieve this goal, original experimental methods and advanced numerical simulations will be used. Ion irradiations will be conducted in order to reproduce the radiation damage. Heat treatments will then be done on the specimens after irradiation. Small tensile samples will be strained in situ, after annealing, inside a transmission electron microscope, at room temperature or at high temperature. Deformation mechanisms observed at nanometer scale and in real time will be simulated using dislocation dynamics, at the same time and space scales. Large scale dislocation dynamics simulations will then be conducted in order to deduce the single-crystal behavior of the material. In parallel with this study at the nanometric scale, a study will also be conducted at the micrometric scale. Nanoindentation and micropillar compression tests will be performed to assess the mechanical behavior after irradiation and annealing. The results of mechanical tests will be compared with large-scale dislocation dynamics numerical simulations.
This study will allow a better understanding of the special behavior of zirconium alloys after irradiation and annealing and then help to develop physically based predictive models. In a future prospect, this work will contribute to improve the safety during transport and storage of spent nuclear fuel.
This PhD project focuses on developing nuclear fuels with improved properties through the addition of a dopant, for use in pressurized water reactors.
In nuclear reactors, the fuel consists of uranium dioxide (UO2) pellets stacked inside zirconium alloy cladding. These pellets, in contact with the cladding, must withstand extreme conditions of temperature and pressure. One of the challenges is to limit chemical interactions that may occur during the migration of fission products from the center to the periphery of the pellet and with the cladding. A notable example of such a phenomenon is the stress corrosion assisted by iodine, which can occur during accidental transients.
One strategy is to dope the UO2 ceramic with a metal oxide in order to control the material’s microstructure and also to modify its thermochemical behavior, thereby limiting both the mobility and corrosive nature of fission gases. Among the possible dopants, manganese oxide (MnO) represents a promising option and a potential alternative to chromium oxide (Cr2O3), which is currently a mature solution for the industry.
This PhD will explore the role of manganese in the sintering of UO2, particularly the microstructure and final properties of the fuel. The work will take place at the CEA Cadarache center, within the Institute for research on nuclear systems for low-carbon energy production (IRESNE).
During these three years, you will be hosted in the Laboratory for the study of uranium-based fuels (LCU) within the fuel study department (DEC), in close connection with the Laboratory for fuel behavior modeling (LM2C).
This research, combining experimentation and modeling, will be structured around three main topics:
• Study of the influence of manufacturing conditions on the microstructure of Mn-doped UO2,
• Investigation of the impact of doping on defect formation in UO2 and the associated properties,
• the contribution to the thermodynamic modelling of the system, based on experimental tests.
During this PhD, you will gain solid experience in the fabrication and advanced characterization of innovative materials, particularly in the field of ceramics for the nuclear industry. Your work could lead to publications, patents, and participation in national and international conferences.
You will also acquire numerous technical skills applicable across various research and industrial fields, including energy, microelectronics, chemical and pharmaceutical industries.