From angstroms to microns: a nuclear fuel microstructure evolution model whose parameters are calculated at the atomic scale
Controlling the behavior of fission gases in nuclear fuel (uranium oxide) is an important industrial issue, as fission gas release or precipitation limit the use of fuels at extended burn-ups. The gas behavior is strongly influenced by the material’s microstructure evolution due to the aggregation of irradiation-induced defects (gas bubbles, dislocation loops and lines). Cluster dynamics (CD) (a kind of rate theory model) is relevant for modelling the nucleation/growth of the defect clusters, there gas content and the gas release. The current model has been parameterized following a multiscale approach, based on atomistic calculations (ab initio or empirical potentials). This model has been successfully applied to annealing experiments of UO2 samples implanted with rare gas atoms and has emphasized the impact of the irradiation damage on gas release. The aim of this PhD thesis is now to improve the model, particularly the damage parameterization, and to extend its validation domain through in depth comparison of simulation with a large set of recently obtained experimental results, such as gas release measurement by annealing of sample implanted in ion beam accelerator, bubble and loop observation by transmission electrons microscopy of implanted or in-pile irradiated samples. This global analysis will finally yield an improved parameterization of the CD model.
The research subject combines a “theoretical” dimension (improving the model) with an “experimental” one (interpreting existing experiments or designing some new ones). The variety of techniques will introduce you into the experimental world and thus broaden your scientific skills. You will be welcomed at the Fuel Behavior Modeling Laboratory (part of the Institute for Research on Nuclear Systems for Low-Carbon Energy Production, IRESNE, CEA Cadarache), where you will benefit from an open environment rich in academic collaborations. You also have to manage collaborations for the experiments analysis, for the model development and for the specification of additional atomistic calculations. You will be at the interface of atomistic techniques, large-scale simulation and various experimental techniques. Therefore, You will develop a broad view of irradiation effects in materials and of multi-scale modelling in solids in general.
This project is an opportunity to contribute to the overall development of numerical physics applied to multi-scale modeling of materials, occupying a pivotal position and adopting a global viewpoint. This will allow experiencing yourself the way computed fundamental microscopic data finally helps solving complex practical issues.
Further readings:
Skorek et al. (2012). Modelling Fission Gas Bubble Distribution in UO2. Defect and Diffusion Forum, 323–325, 209.
Bertolus et al. (2015). Linking atomic and mesoscopic scales for the modelling of the transport properties of uranium dioxide under irradiation. Journal of Nuclear Materials, 462, 475–495.
Study of the behaviour of mixed oxide fuels with degrade isotopy at the beginning of life.
France has decided to adopt a 'closed' nuclear fuel cycle. This involves processing spent fuel to recover valuable materials such as uranium and plutonium, while other compounds such as fission products and minor actinides constitute final waste. UO2 fuel irradiated in pressurised water reactors (PWRs) is currently reprocessed to produce plutonium (PuO2), which is then reused in the form of mixed oxide (MOX) fuel. This fuel is then irradiated in PWRs, a process known as plutonium monorecycling. The CEA is currently studying the multi-recycling of materials using fuels containing Pu from the processing of spent MOX assemblies. However, this multi-recycled plutonium contains a higher proportion of highly alpha-active isotopes (Pu238, Pu240 and Pu241/Am241), resulting in more severe alpha self-irradiation than current MOX fuels experience [1]. This exacerbates certain physical phenomena [2-5], such as fuel swelling due to helium precipitation and the creation of crystal defects and decreased thermal conductivity [6-8], which can alter its behaviour in the reactor.
The proposed thesis will study the impact of these phenomena on the behaviour of MOX fuels at the beginning of the irradiation, using a combination of experimentation and modelling. Heat treatments will be employed to analyse the mechanisms of crystal defect healing and helium behaviour. Various experimental techniques will be employed to characterise the structure and microstructure (X-ray diffraction, scanning electron microscopy (SEM), Raman spectroscopy and microprobe analysis), defect densities (transmission electron microscopy (TEM)), helium release (KEMS), thermal gradient reproduction (CLASH laser) and thermal conductivity (LAF laser). The results will inform simulations modelling the microstructure and thermal properties.
This cross-disciplinary study will improve our understanding of the phenomena involved in the initial power-up of fuels damaged by alpha self-irradiation, particularly the impact of helium produced by decay.
You will be based at the Multi-Fuel Design and Irradiation Laboratory (LECIM) within the Research Institute for Nuclear Systems for Low-Carbon Energy Production at CEA/Cadarache. For the experimental part of the project, you will collaborate with the Chemical Analysis and Materials Characterisation Laboratory (LMAT) at CEA/Marcoule and the European Research Centre (JRC) in Karlsruhe. You will have the opportunity to publish your results through scientific publications and conference presentations. This role offers the chance to develop your expertise in a variety of techniques that can be applied across multiple fields of materials science and engineering.
[1]O. Kahraman, thésis, 2023.[2]M. Kato et al., J Nucl Mater, 393 (2009) 134–140.[3]L. Cognini et al., Nuclear Engineering and Design 340 (2018) 240–244.[4] T. Wiss et al., Journal of Materials Research 30 (2015) 1544–1554.[5]D. Staicu et al., J Nucl Mater 397 (2010) 8–18.[6] T. Wiss et al.,Front. Nucl. Eng. 4 (2025) 1495360.[7]E.P. Wigner, J. Appl. Phys. 17 (1946) 857–863.[8]D. Staicu et al., Nuclear Materials and Energy 3–4 (2015) 6–11.
Parallel simulation and adaptive mesh refinement for 3D solids mechanics problems
The challenge of this PhD thesis is to implement adaptive mesh refinement methods for non-linear 3D solids mechanics adapted to parallel computers.
This research topic is proposed as part of the NumPEx (Digital for Exascale) Priority Research Programs and Equipment (PEPR). It is part of the Exa-MA (Methods and Algorithms for Exascale) Targeted Project. The PhD will take place at CEA Cadarache, within the Institute for Research on Nuclear Energy Systems for Low-Carbon Energy Production (IRESNE), as part of the PLEIADES software platform development team, which specializes in fuel behavior simulation and multi-scale numerical methods.
In finite element simulation, adaptive mesh refinement (AMR) has become an essential tool for performing accurate calculations with a controlled number of unknowns. The phenomena to be taken into account, particularly in solids mechanics, are often complex and non-linear: contact between deformable solids, viscoplastic behaviour, cracking, etc. Furthermore, these phenomena require intrinsically 3D modelling. Thus, the number of unknowns to be taken into account requires the use of parallel solvers. One of the current computational challenges is therefore to combine adaptive mesh refinement methods and nonlinear solid mechanics for deployment on parallel computers.
The first research topic of this PhD thesis concerns the development of a local mesh refinement method (of block-structured type) for non-linear mechanics, with dynamic mesh adaptation. We will therefore focus on projection operators to obtain an accurate dynamic AMR solution during the evolution of refined areas.
The other area of research will focus on the effective treatment of contact between deformable solids in a parallel environment. This will involve extending previous work, which was limited to matching contact meshes, to the case of arbitrary contact geometries (node-to-surface algorithm).
The preferred development environment will be the MFEM tool. Finite element management and dynamic re-evaluation of adaptive meshes require assessing (and probably improving) the efficiency of the data structures involved. Large 3D calculations will be performed on national supercomputers using thousands of computing cores.
his will ensure that the solutions implemented can be scaled up to tens of thousands of cores.
Design artificial intelligence tools for tracking Fission Product release out of nuclear fuel
The Laboratory for the Analysis of Radionuclide Migration (LAMIR), part of the Institute for Research on Nuclear Systems (IRESNE) at CEA Cadarache, has developed a set of advanced measurement methods to characterize the release of fission products from nuclear fuel during thermal transients. Among these innovative tools is an operando in situ imaging system that enables real-time observation of these phenomena. The large amount of data generated by these experiments requires dedicated digital processing techniques that account for both the specificities of nuclear instrumentation and the underlying physical mechanisms.
The goal of this PhD project is to develop an optimized data processing approach based on state-of-the-art Artificial Intelligence (AI) methods.
In the first phase, the focus will be on processing thermal sequence images to detect and analyze material movements, aiming to identify an optimal image-processing strategy defined by rigorous quantitative criteria.
In the second phase, the methodology will be extended to all experimental data collected during a thermal sequence. The long-term objective is to create a real-time diagnostic tool capable of supporting experiment monitoring and interpretation.
This PhD will be carried out within a collaborative framework between LAMIR, which has recognized expertise in nuclear fuel behavior analysis and imaging, and the Institut Fresnel in Marseille, known for its strong background in image analysis and artificial intelligence.
The candidate will benefit from a multidisciplinary and stimulating research environment, with opportunities to present and publish their work at national and international conferences and in peer-reviewed journals.
Development of manganese-doped uranium oxide fuel: sintering mechanisms and microstructural changes
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.
Atomistic investigation of the thermophysical properties of metallic nuclear fuel UMo
Uranium – molybdenum alloys UMo present excellent thermal properties and a good uranium density. For those reasons, they are considered as nuclear fuel candidates for research reactors. It is therefore crucial for the CEA to deploy new computational methodologies in order to investigate the evolution of their thermo-physical properties under irradiation conditions.
This project is centered on the application of atomistic methods in order to investigate the stability and diffusion of intra-granular xenon clusters within the metallic nuclear fuel UMo.
The first step of your work will involve continuing the development of atomic-scale computational models for UMo, as initiated within the host laboratory. These models use machine learning methods to develop interatomic potentials and will be validated by comparison with existing experimental data for this material. They will then be used to assess the temperature-dependent evolution and the impact of defect accumulation (both point and extended defects) on several thermophysical properties critical to fuel modeling, such as elastic properties, density, thermal expansion, as well as thermal properties like specific heat and thermal conductivity. In collaboration with other researchers in the department, you will format these results for integration into the Scientific Computing Tools used to simulate the behavior of nuclear fuels.
In a second phase, you will be responsible for extending the validity of your models to account for the formation of fission gases such as xenon within UMo single crystals. This will enable you to simulate the stability of xenon clusters in UMo crystals. These calculations, performed using classical molecular dynamics methods, will be systematically compared with experimental observations obtained via transmission electron microscopy.
The results obtained during the various stages of this project will be completely innovative and will be the subject of scientific publications as well as presentations at international scientific conferences. Besides, this work will enable you to complement your training by acquiring skills applicable to many areas of materials science, including ab initio calculations, machine learning-based interatomic potential fitting, classical molecular dynamics, use of CEA supercomputers, and key concepts in statistical physics and condensed matter physics—fields in which the supervising team members are recognized experts.
You will join the Fuel Behavior Modeling Laboratory at the Research Institute for Nuclear Systems for Low-Carbon Energy Production (IRESNE, CEA Cadarache), a dynamic research team where you will have regular opportunities to interact with fellow PhD students and researchers. This environment also provides extensive opportunities for national and international collaboration, including with:
• Developers and users of the MAIA fuel performance code (dedicated to research reactor fuel studies),
• Experimental researchers from the Nuclear Fuel Studies Department,
• Teams from other CEA centers (Saclay, CEA/DAM),
• International partners.
This rich and multidisciplinary context will enable you to fully engage with the scientific community focused on nuclear materials science.
[1] Dubois, E. T., Tranchida, J., Bouchet, J., & Maillet, J. B. (2024). Atomistic simulations of nuclear fuel UO2 with machine learning interatomic potentials. Physical Review Materials, 8(2), 025402.
[2] Chaney, D., Castellano, A., Bosak, A., Bouchet, J., Bottin, F., Dorado, B., ... & Lander, G. H. (2021). Tuneable correlated disorder in alloys. Physical Review Materials, 5(3), 035004.
UO2 Powders: Morphological Characterization of Aggregates and Study of Their Interactions Using a Combined Experimental and numerical Approach
This PhD thesis is part of the optimization of nuclear fuel fabrication processes, which rely on the powder metallurgy of uranium dioxide (UO2) and plutonium dioxide (PuO2). These powders exhibit a hierarchical microstructure composed of crystallites forming rigid aggregates, themselves agglomerated into larger structures. The morphology and interactions between aggregates play a key role in the macroscopic behavior of the powders—particularly their flowability, compressibility, and agglomeration capacity—and directly influence the quality of the fuel pellets obtained after pressing and sintering. However, the experimental characterization of these aggregates remains complex and does not yet allow for the establishment of a predictive link between synthesis processes and morphological properties.
The objective of this thesis is to combine experimental and numerical approaches to achieve a detailed characterization of the aggregates in a reference powder. Experimentally, techniques such as Scanning Electron Microscopy (SEM), specific surface area measurement (BET), and laser granulometry will be used to determine particle size, roughness, and size distribution. In parallel, numerical simulations based on the Discrete Element Method (DEM) will be employed to construct a granular digital twin consistent with the experimentally measured properties. This digital twin will allow the reconstruction of the internal structure of the aggregates, the evaluation of interparticle adhesion forces, and the analysis of agglomeration and densification phenomena under controlled conditions.
The PhD will take place at CEA Cadarache within the Institute for Research on Nuclear Systems for Low-Carbon Energy Production (IRESNE). The student will be assigned to the PLEIADES Fuel Development Laboratory (LDOP), which specializes in simulating nuclear fuel behavior (from fabrication to in-reactor performance) and in multi-scale numerical methods. The work will be carried out in collaboration with the CNRS/LMGC in Montpellier, internationally recognized for its research on granular materials, and with the Uranium Fuel Laboratory (LCU – CEA Cadarache), which has extensive experience in the experimental characterization of uranium powders.
The PhD candidate is expected to demonstrate strong skills in numerical simulation and in the physical analysis of results. He will share its results through publications and conference presentations and will have the opportunity to learn or further develop various experimental and numerical techniques that can be applied in other contexts.In particular, the issues related to the physics of granular media — which constitute the core of this PhD — are of significant industrial relevance and are common to many other sectors handling powders, such as pharmaceuticals, agri-food, and powder metallurgy.
[Hebrard2004] S.Hebrard, Etude des mécanismes d’évolution morphologique de la structure des poudres d’UO2 en voie sèche, thèse de doctorat, CEA-LSG2M-COGEMA), 2004.
[Pizette2010] P. Pizette, C.L. Martin a, G. Delette, P. Sornay, F. Sans, Compaction of aggregated ceramic powders: From contact laws to fracture and yield surfaces, Powder Technology, 198, 240-250, 2010.
[Tran2025] T.-D. Tran , S. Nezamabadi , J.-P. Bayle, L. Amarsid, F. Radjai , Effect of interlocking on the compressive strength of agglomerates composed of cohesive nonconvex particles, Advanced Powder Technology 36, 2025.
Micromechanical Modeling of the Behavior of Polycristals with Imperfect Interfaces: Application to Irradiated UO2 Fuel
This thesis aims to analyze the thermomechanical properties of UO2 fuel used in pressurized water reactors (PWRs), accounting for the effects of microscopic defects. It focuses particularly on intergranular decohesion phenomena observed at various stages of fuel evolution, notably prior to crack initiation and propagation. The objective of this thesis is to clarify the impact of decohesion on both the local and effective properties of UO2 during irradiation. To this end, intergranular decohesion is modeled at the local scale by means of imperfect interface models, which ensure traction continuity while allowing for a displacement jump at grain boundaries. This modeling approach enables the development of homogenization models incorporating innovative theoretical and numerical advances, capable of capturing the behavior of the fuel at very high temperatures, under off-normal and accidental conditions. This work will be conducted at CEA Cadarache,in the Institute for Research on Nuclear Systems for Low-Carbon Energy Production (IRESNE), in close collaboration with national and international research teams. The tools developed will contribute to improving our understanding of the fuel's properties and to enhancing the accuracy and reliability of existing models, particularly those implemented in the PLEIADES simulation platform developed by the CEA in collaboration with French nuclear industry partners.
Nuclear fuel fragmentation under thermal gradient of fuel during laser heating: correlation, numerical simulation and and adaptation of the experimental setup.
The aim of this thesis is to simulate the cracking of nuclear fuel, which consists of a brittle ceramic material, uranium dioxide, during laser heating experiments. The objective is to compare the numerical results with experimental data through image correlation. This comparison will make it possible to optimize the experimental setup, improve the quality of the experimental results, and move toward a quantitative validation of the gradient damage models used in the simulations.
The starting point of this work is a campaign of uranium dioxide pellet fragmentation by laser heating, carried out as part of the PhD of Hugo Fuentes [1] in one of the experimental laboratories of the Institute for Research on Nuclear Energy Systems for Low-Carbon Energy Production (IRESNE) at CEA Cadarache (DEC/SA3E/LAMIR). This heating technique reproduces temperature gradients representative of reactor conditions. For each test, films showing the evolution of cracks and surface temperature changes in the pellet are available.
These films will be analyzed by digital image correlation (DIC) [3] using an in-house software tool to determine optimal boundary conditions for the numerical simulations and extract relevant data for model validation. The experiments will then be modeled using gradient damage models developed in the PhD theses of David Siedel and Pedro Nava Soto [2]. Based on the results obtained, the PhD candidate will be able to optimize and/or adapt the setup to study other operating conditions and conduct a new experimental campaign.
The PhD student will work in close collaboration between a simulation laboratory and an experimental laboratory within the IRESNE Institute at CEA Cadarache. The proposed work is open-ended and may be promoted through participation in national or international conferences and the publication of scientific articles in high-impact journals.
References
[1] Fuentes, Hugo, Doualle, Thomas, Colin, Christian, Socié, Adrien, Helfer, Thomas, Gallais, Laurent, and Lebon, Frédéric. Numerical and experimental simulation of nuclear fuel fragmentation via laser heating of ceramics. In: Proceedings of Top Fuel 2024, Grenoble, 29 September 2024.
[2] Nava Soto, Pedro, Fandeur, Olivier, Siedel, David, Helfer, Thomas, and Besson, Jacques. Description of thermal shocks using micromorphic damage gradient models. European Solid Mechanics Conference, Lyon, 2025.
[3] Castelier Etienne, Rohmer E., Martin E., Humez B. Utilisation de la dimension temporelle pour ameliorer la
correlation d'images. 20 eme Congres Francais de Mecanique, 2011.
DEM-LBM Coupling for simulating the ejection of immersed granular media in compressible Fluid under High Pressure Gradients
In Pressurized Water Reactors (PWRs), the fuel consists of uranium oxide (UO2) pellets stacked in metallic cladding. During a Loss of Coolant Accident (LOCA) scenario, the rapid temperature increase can cause deformation and sometimes rupture of these claddings. This phenomenon can potentially lead to the ejection of fuel fragments into the primary circuit. This phenomenon is known as FFRD (Fuel Fragmentation, Relocation, and Dispersal). Since the cladding is the first safety barrier, it is crucial to evaluate the amount of dispersed fuel. Experimental studies have shown that the size, shape of the fragments, shape of the breach, and internal pressure significantly influence the ejection. However, the speed of the initial depressurization phase makes direct measurements difficult. Numerical approaches, particularly through fluid-grain coupling (LBM-DEM), offer a promising alternative. The IRESNE Institute at CEA Cadarache, through the PLEIADES software platform, is developing these tools to model the behavior of fragments. However, the compressibility of the gas needs to be integrated to accurately reproduce the initial depressurization. In this context, the laboratory M2P2 of the CNRS, a specialist in modeling compressible flows with the LBM method and developer of the ProLB software, brings its expertise to integrate this effect. The thesis therefore aims to design and improve a compressible model in the LBM-DEM coupling, to conduct a parametric study, and to develop a 3D HPC demonstrator capable of leveraging modern supercomputers.
This CEA thesis will be conducted in close collaboration between the Fuels Research Department (DEC) of the IRESNE Institute at CEA Cadarache and the laboratory M2P2 (CNRS). You will be primarily located at M2P2 but will make regular visits to CEA within the Fuel Simulation Laboratory, to which you will be affiliated. The approaches developed in this thesis ensure a high scientific level with numerous potential industrial applications both within and outside the nuclear field.