Nickel-based alloys are structural materials of choice for Molten Salt Reactors (MSRs). They offer excellent mechanical properties and good corrosion resistance. In these materials, helium production, mainly caused by the transmutation of nickel by fast neutrons, can reach levels sufficient to strongly embrittle the material or cause it to swell under irradiation. Helium is hardly soluble in the material, and condenses in the form of bubbles or segregates at grain boundaries. To limit these phenomena and successfully trap the helium, one solution is to introduce into the material to be irradiated a high density of nanoprecipitates, whose interfaces will serve as germination sites for nanometric bubbles capable of trapping the helium atoms, preventing the latter from migrating to the grain boundaries and degrading the material's performance. Corrected transmission electron microscopy will be used to study the precipitation kinetics of the thermodynamically expected phases, as well as the atomic structure of the interfaces formed between the precipitates and the matrix. A phase-field simulation of precipitation will also be considered. Finally, the He trapping mechanisms at the interfaces will be studied using electron energy loss spectroscopy (EELS).

Uranium oxide (UO2) is the usual fuel used in nuclear fission power plants. As such, its behaviour under irradiation has been extensively studied. Irradiation creates vacancies or interstitial defects that control the evolution of the material's microstructure, which in turn impacts its physical (e.g. thermal conductivity) and mechanical properties. Interstitial clusters in particular play a major role.
On the one hand, at the smallest sizes, the diffusion of interstitials in UO2 is still relatively poorly understood. Experimentally, we observe the appearance of dislocation loops made up of interstitials as large as ten nanometres. Conversely, no cavities are observed and the vacancy defects remain sub-nanometric in size. This indicates that interstitials diffuse more rapidly than vacancies, with diffusion allowing interstitials to agglomerate and form loops. However, atomic-scale calculations show no major difference between the diffusion coefficients of vacancies and interstitials in UO2. One hypothesis to explain this apparent contradiction is that interstitial clusters diffuse rapidly (Garmon, Liu et al. 2023).
On the other hand, the three-dimensional interstitial clusters are expected to be the seeds of the dislocation loops observed by transmission electron microscopy in irradiated uranium oxide. However, the mechanisms by which the aggregates transform into loops and the nature of the loops changes remain poorly understood in uranium oxide. These mechanisms have very recently been elucidated for face-centred cubic metals (Jourdan, Goryaeva et al. 2024). It is possible that comparable mechanisms are at work in UO2 with the complication induced by the existence of two sub-lattices.
We therefore propose to study interstitial clusters in UO2 using atomic-scale simulations.
We will first study the structure of these three-dimensional subnanometric clusters. To do this, we will use artificial intelligence tools for classifying defect structures developed in the laboratory (Goryaeva, Lapointe et al. 2020). We will study the diffusion of these objects using molecular dynamics and automatic searches for migration saddle points using kinetic-ART type tools (Béland, Brommer et al. 2011). Secondly, we will study the relative stability of 3D clusters and loops of faulted and perfect dislocations and the transformations between these different objects.
This study will be based on interatomic interaction potentials. We will start by using empirical potentials available in the literature before turning to Machine Learning-type potentials (Dubois, Tranchida et al. 2024) under development at the CEA Cadarache Fuel Studies Department.

Béland, L. K., et al. (2011). ‘Kinetic activation-relaxation technique.’ Physical Review E 84(4): 046704.

Chartier, A., et al. (2016). ‘Early stages of irradiation induced dislocations in urania.’ Applied Physics Letters 109(18).

Dubois, E. T., et al. (2024). ‘Atomistic simulations of nuclear fuel UO2 with machine learning interatomic potentials.’ Physical Review Materials 8(2).

Garmon, A., et al. (2023). ‘Diffusion of small anti-Schottky clusters in UO2.’ Journal of Nuclear Materials 585: 154630.

Goryaeva, A. M., et al. (2020). ‘Reinforcing materials modelling by encoding the structures of defects in crystalline solids into distortion scores.’ Nature Communications 11(1).

Jourdan, T., et al. (2024). ‘Preferential Nucleation of Dislocation Loops under Stress Explained by A15 Frank-Kasper Nanophases in Aluminum.’ Physical Review Letters 132(22).

Heterogeneous borosilicate-based glasses contain crystalline or amorphous precipitates forming secondary phases embedded within the glass matrix. These materials are valued for their high thermal shock resistance and excellent chemical durability, making them ideal for various applications such as cookware and laboratory equipment. In particular, within the nuclear industry, many wasteforms effectively function as glass-ceramics due to the presence of elements that form precipitates.

It is well known that secondary phases can significantly affect mechanical properties, particularly fracture toughness. However, the specific mechanisms by which they influence mechanical properties at the atomic scale remain poorly understood. In particular, whether they are crystalline or amorphous and the structure of their interface with the bulk glass are expected to play a crucial role.

The primary aim of this project is to investigate the specific mechanisms by which precipitates influence mechanical properties at the atomic scale.
Additionally, it seeks to understand how these precipitates affect crack propagation.
For this purpose, numerical modelling tools based on molecular dynamics will be employed.
This technique simulates the behaviour of individual atoms over time under different testing conditions.
Thus, it enables probing the local structure of crack tips and how they interact with precipitates at the atomic level, providing valuable insights into the mechanisms underlying crack resistance in heterogeneous glasses.

Alloy 718, a nickel-based alloy, is used in fuel assemblies for pressurized water reactors (PWRs). In service, these components are subjected to high mechanical stress, neutron irradiation and exposure to the primary environment. Usually, this alloy shows very good resistance to intergranular cracking. However, there are microstructural and/or irradiation conditions which, by modifying the mechanical properties and plasticity mechanisms, make the material susceptible to intergranular cracking in the primary PWR environment.

In this context, the aim of this thesis will be to study the influence of microstructure (via different heat treatments) and irradiation on deformation localization and susceptibility to intergranular cracking in primary PWR media.

To this end, two grades will be tested, one deemed sensitive and the other not. In-situ SEM tensile tests on samples whose microstructure has been previously characterized by EBSD will be carried out to identify the types of intra- and intergranular deformation localization and their evolution. The non-irradiated state will be characterized as the reference state. In addition, exposure and intergranular cracking tests in the primary medium (coupons, slow tension, etc.) will be carried out on both grades and at different irradiation levels. The microstructure as well as surface and intergranular oxidation of the specimens will be characterized by various microscopy techniques (SEM, EBSD, FIB and transmission electron microscopy).

This thesis constitutes for the candidate the opportunity to address a problem of durability of metallic materials in their environment following a multidisciplinary scientific approach combining metallurgy, mechanics and physico-chemistry and based on the use of various cutting-edge techniques available at the CEA. The skills that he will thus acquire can therefore be valued during the rest of his career in the industry (including non-nuclear) or in academic institutions.