Study of plutonium oxalate formation mechanisms – Application to molten salt reactors
Molten salt reactors (MSRs) offer a promising alternative for sustainable nuclear energy production, thanks to their intrinsic safety and their ability to close the nuclear fuel cycle, notably through the use of a fast neutron spectrum. This type of reactor can use liquid chloride salts containing plutonium and other actinides as fuel. As part of the development of this nuclear pathway, the CEA supports the development of a PuCl3 production process. The synthesis of this chloride has already been carried out at small scale at the CEA and elsewhere in the world. Several starting materials can be used for the synthesis of the trichloride, notably plutonium metal, plutonium oxide and plutonium oxalate. The most industrially promising synthesis route is the oxalate route, because it can be transferred to the equipment already present at the La Hague site. This process consists of converting the oxalate into plutonium chloride via a gas–solid reaction with a chlorinating agent, such as HCl for example. However, the reaction mechanism and the decomposition of the oxalate in a chlorinated environment are still poorly understood. A detailed understanding of this transformation would make it possible to optimize operating conditions and facilitate the scale-up of this synthesis. The topic will initially focus on determining the reaction mechanism of Ce oxalate (a surrogate for Pu) to the chloride. Small-scale studies will be performed to identify the various reaction intermediates using analytical techniques such as X-ray diffraction (XRD), thermogravimetric analysis / differential thermal analysis (TGA/DTA) and analysis of the gases produced during the reaction. The kinetics as well as the enthalpy changes will also be studied in order to obtain key data for modelling a large-scale process. Subsequently, an optimization of the PuCl3 synthesis at the scale of a few tens of grams will be carried out. These studies will first be conducted under non-radioactive conditions on a surrogate to validate the experimental approach, before being transposed to radioactive conditions.
Effects of alpha decay on the alteration of nuclear glasses: simulation, understanding, and consideration in geochemical models
This Ph-D at the CEA on the alteration of nuclear glass is central to the challenges of sustainable radioactive waste management. The doctoral student will acquire expertise in materials and modeling, paving the way for exciting careers in research, engineering, or the nuclear industry. In deep geological storage, contact with groundwater can cause glass alteration, which is the main source of radionuclide release. The CEA is developing a multi-scale model that needs to be adapted to take into account the effects of glass self-irradiation. The aim of the thesis is to identify the mechanisms modified by irradiation and to parameterize the model. The doctoral student will conduct controlled irradiation experiments on non-radioactive glasses and compare them to ²44Cm-doped active glass. The structural and physicochemical changes induced will be characterized using various techniques (Raman, IR, NMR, SEM, TEM, DSC, etc.). Targeted alteration tests will be used to observe the impact of the level of damage on the kinetics of alteration. The results will be used to adjust and validate the predictive model under conditions representative of geological storage. The work will be carried out both in an active environment (shielded cells) and in an inactive laboratory. An M2 internship is available on the same subject. Profile: M2 or materials engineer, physical chemistry.
Optimization of the conditions for the electrolytic synthesis of metallic uranium
Reprocessed uranium (RepU), derived from the reprocessing of spent nuclear fuel, represents a material whose reuse in power plants would allow for the sustainable management of energy resources. Accordingly, the CEA is supporting the nuclear industry to evaluate the feasibility of enriching this RU via the laser route. This technology requires, as a process input, uranium in the form of a metallic alloy. Consequently, an upstream process for the synthesis of metallic uranium must be developed and optimized to build a sovereign RepU sector.
One of the routes under study for synthesizing metallic uranium is the electrolysis of uranium oxide, previously dissolved in high-temperature molten fluoride salt media. This synthesis, which was previously implemented in the United States using the aluminum synthesis process, now requires a re-appropriation and optimization of experimental conditions.
In a first phase, the PhD student will conduct a systematic study of the electrolyte, in order to understand the influence of key parameters—salt composition, temperature range, redox environment, material compatibility, and oxide solubility—on the behavior of the electrolysis bath. For each parameter, targeted tests will be conducted: thermochemical characterization of the salt (melting point, volatility, purification, etc.), evaluations of the kinetics and the solubility limit of uranium oxide in the bath (a crucial point of the process), electrochemical tests aimed at identifying redox systems of interest, as well as studies on the resistance of materials when in contact with the molten salt and liquid metal. All of these investigations will make it possible to define the optimal experimental conditions for the controlled implementation of metal synthesis by oxide electrolysis.
In a second phase, once these conditions are established, the work will focus on the formation of the metal at the electrode, its recovery, and its characterization. The quantity and quality of the metal produced after electrolysis will be the major criteria for validating the selected experimental parameters.
All acquired data will be utilized for the design of pilot and industrial scale electrolyzers, and will feed into future digital models that will be developed. The results obtained may be the subject of presentations at international conferences and publications.
These studies will be carried out at the laboratory scale using active material, with work phases on simulants to grasp the implementation of the process and scaling up. The host laboratory, which operates in both these environments, specializes in the implementation of thermal processes and pyrochemical tests.
The candidate should ideally have a Master 2 or engineering school degree in chemistry or physics.
At the end of this thesis work, the PhD student will have acquired expertise in experimental techniques related to metallic synthesis by electrolysis, from the design of electrochemical devices to the multi-scale characterization of materials. Furthermore, their involvement in a sovereign project focused on strategic metals will open up numerous employment prospects in academic research or industrial R&D, both in the nuclear sector and in other fields of chemistry and materials.
Impact of fission products and microstructure on the thermophysical properties of LWR (U,Pu)O2-x fuel
In France, mixed oxide fuel (MOX, (U,Pu)O2) is currently deployed in several pressurized water reactors (PWRs) operated by EDF. To ensure continued low-carbon electricity production, a broader use of MOX fuel across the French nuclear fleet is expected to become essential in the near future. During reactor operation, U1??Pu?O2?? fuels undergo significant changes in their physical properties and microstructure, primarily due to the accumulation of dozens of lighter elements generated by plutonium’s fission, commonly referred to as fission products (FPs). Because of the high radiotoxicity of irradiated fuel, surrogate materials known as SIMMOX have been developed. In a previous PhD project, we established a synthesis route enabling the production of SIMMOX doped with up to twelve fission products, successfully reproducing the microstructure of irradiated PWR MOX fuel.
To maintain an adequate margin to fuel melting during irradiation, it is crucial to understand how the thermophysical and thermodynamic properties of MOX fuel evolve under these conditions. This PhD project aims to measure these properties on a representative MOX composition currently used in EDF reactors. The key properties of interest include thermal conductivity, heat capacity, and melting temperature. These measurements will be carried out at the JRC-Karlsruhe (Germany) during a research stay of approximately 12 months. Subsequently, the samples will be returned to CEA-Marcoule, where the impact of high-temperature exposure on actinide and fission product speciation, as well as on the microstructural evolution of the MOX fuel, will be investigated. In parallel, the experimental work will be complemented by thermodynamic modeling using the CALPHAD approach, in order to identify the mechanisms and phase equilibria governing high-temperature behavior during property measurements.
Investigation of geopolymer durbility for radioactive wastewater treatment
The reprocessing of spent nuclear fuel generates radioactive effluents that require appropriate treatment. To meet industrial and regulatory challenges, the CEA is developing geopolymer-based adsorbent materials that are robust, cost-effective, and efficient for capturing Cs-137 and Sr-90. Their performance can be enhanced through the incorporation of selective adsorbents (such as zeolites) and through innovative shaping processes (3D printing, beads, foams) optimized for column adsorption.
The durability of these materials remains a critical issue, as their leaching and ageing mechanisms in column systems are still poorly understood. This PhD project will focus on studying these phenomena in order to assess the impact of effluent chemistry on the stability and efficiency of geopolymers. The work will include material synthesis, batch and column sorption tests, and the use of modelling tools to interpret alteration mechanisms. The scientific challenge is to identify the key physicochemical markers of geopolymer degradation in the targeted liquid effluents and to link them with column sorption performance.
The PhD candidate will join the Laboratory for Supercritical Processes and Decontamination (LPSD), renowned for its expertise in column-based ion extraction and adsorbent characterization. He/she will collaborate with specialists at CEA Marcoule and with the laboratory teams, and will regularly present project progress to the industrial partner. Upon completion of the PhD, the candidate will have developed recognized expertise at the interface of materials science, chemistry, and column adsorption processes. This work will open a wide range of opportunities: R&D positions in the nuclear sector, waste management, and functional materials; academic pathways (postdoctoral research, academia, teaching); or contributions to major energy and environmental challenges.
Effect of gamma-ray irradiation on ferroelectric, hafnia-based, non-volatile memory for use in extreme environments
The emergence of hafnia-based ferroelectric (FE) memories has opened a new paradigm for ultra-low-power edge computing. Hafnia is fully compatible with CMOS technology and is ultra low-power—three orders of magnitude less than other emerging memory technologies.
These advantages align with strategic applications in space, defense, medical, nuclear safety, and heavy-duty transport, where electronics face harsh radiation environments.
Imprint induces a shift of the Polarization-Voltage (P-V) curve along the voltage axis and is attributed to charge trapping/detrapping, domain pinning and charged defects. All may be accentuated under irradiation.
The project will use advanced photoelectron spectroscopy techniques including synchrotron radiation induced Hard X-ray photoelectron spectroscopy and complementary structural analysis including high-resolution electron microscopy, X-ray diffraction and near field microscopy. The experimental characterization will be accompanied by theoretical calculations to simulate the material response to irradiation
The work will be carried out in the framework of close collaboration between the CEA/Leti in Grenoble providing the samples, integrated devices and wafer scale characterization and the CEA/Iramis in Saclay for the fundamental analysis of the material properties, irradiation experiments and device scale characterizations.
Coarse-grained simulations and viscoelastic behavior of polymeric photoactuators: a bottom-up strategy
Mechanical actuators, like muscles, are materials that can change their own macroscopic shape to perform mechanical work when submitted to an external stimulus, such as light irradiation. The resulting photoactuators (PA) are based on a variety of photoactive materials including gels, crystals, liquid crystal elastomers (LCE) or polymer films forming polymeric PA (PPA). This project focuses on PPAs, usually made of elastomers in which photoactive molecules are inserted. To optimize the PPA properties, a precise understanding of the behavior of these materials at all scales is necessary. PPAs are viscoelastic by nature and therefore the continuous scale modeling of their behavior requires the knowledge of some specific mechanical properties, like the time-dependent relaxation moduli G(t) and K(t). At the supramolecular scale, these relaxation moduli can be obtained by Molecular Dynamics (MD) simulations using the Green-Kubo relation [3]. However, for these materials, the G(t) and K(t) timescales far exceed the accessible timescales of MD (on the order of thousands of seconds vs. microseconds). This PhD work has thus two main objectives to reduce this gap :(i) temperature-accelerated dynamics, (ii) anisotropic coarse-grained (CG) simulations.
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.
In situ and real-time characterization of nanomaterials by plasma spectroscopy
The objective of this Phd is to develop an experimental device to perform in situ and real time elemental analysis of nanoparticles during their synthesis (by laser pyrolysis or flame spray pyrolysis). Laser-Induced Breakdown Spectroscopy (LIBS) will be used to identify the different elements present and their stoichiometry.
Preliminary experiments conducted at LEDNA have shown the feasibility of such a project and in particular the acquisition of a LIBS spectrum of a single nanoparticle. Nevertheless, the experimental device must be developed and improved in order to obtain a better signal to noise ratio, to increase the detection limit, to take into account the different effects on the spectrum (effect of nanoparticle size, complex composition or structure), to automatically identify and quantify the elements present.
In parallel, other information can be sought (via other optical techniques) such as the density of nanoparticles, the size or shape distribution.
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.