Behavior of matter under isothermal dynamic compression: displacement of chemical reactivity; synthesis of new metastable materials; phase transition mechanisms.
The Diamond Anvil Cell equipped with piezoelectric actuators, or d-CED, is an innovative device that can generate dynamic compressions and decompressions over a wide range of pressure variation rates. The d-CED thus enables finely controlled dynamic stresses to be applied, with (de)compression rates that can vary over several orders of magnitude along isothermal paths. This paves the way for the creation of reference databases for the validation of microscopic mechanisms. Furthermore, the compression or decompression rates can be equated to ultra-fast heating or cooling rates of the sample, offering the possibility of exploring, in a highly controlled manner, certain phenomena still debated in the literature, such as the maximum stability of a solid beyond its melting point.
The objective of this thesis is to exploit the new possibilities offered by d-CED to demonstrate new phenomena or gain a detailed understanding of certain effects discussed in the literature, by performing ultra-fast temperature variations. A first application will consist of studying the nucleation kinetics of rare gases (Ar, Ne, Kr) as a function of the compression rate, and comparing them with recent measurements made at the XFEL in cryogenic jets. A second objective will be to study chemical changes, with an initial study focusing on the modification of the reactivity of nitromethane, a reference explosive. Another area of study will concern the synthesis of new molecular compounds from mixtures of dense molecular fluids (N2, H2, O2).
Development of a new method for analyzing the manufacturing range of cladding tubes for fourth-generation nuclear reactors
Austenitic steel AIM1 is considered as benchmark alloy for fuel cladding in fourth-generation lead (RNR-pb) or sodium (RNR-Na) reactors. This alloy is currently undergoing qualification testing. The solution treatment of titanium carbides is a key point to obtaining a microstructure that is resistant to irradiation and, in particular, to the phenomenon of irradiation swelling (condensation of vacancies that form cavities in the material). It depends mainly on the quality of the thermomechanical treatments carried out during industrial manufacturing. New approaches to fine characterization (combining electron microscopy, atom probe tomography (APT), and thermoelectric power (TEP)) make it possible to specify microstructural changes during the manufacturing process.
In this thesis, we propose to study a new criterion for assessing the manufacturing quality of AIM1. The primary objective is to determine to which extent the variations in the material's thermoelectric power (TEP) can contribute to the implementation of an acceptance test that can be applied industrially. We will seek to acquire the knowledge that will enable us to perform a simple measurement to validate the metallurgical state of the tubes by having a precise understanding of the microstructures that produce the TEP signal intensity.
This study, which will combine experimental work and modeling, will enable to acquire skills in transmission electron microscopy, atom probe tomography, behavior under ion irradiation, and cluster dynamics modeling.
Atomic scale modeling of radiation induced segregation in Zr(Nb) alloys
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.
Experimental study of Nanometric-Scale Microstructural and Microchemical Evolution in Zirconium Alloys under Irradiation
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.
Impact of irradiation parameters on the alpha’ phase formation in oxide dispersion strengthened steels
Ferritic-martensitic oxide dispersion strengthened steels (ODS steels) are materials of great interest in the nuclear industry. Predominantly composed of iron and chromium, these materials can become brittle due to the precipitation of a chromium-rich phase, called a', under irradiation. This phase, known to be sensitive to irradiation conditions, provides an ideal topic for a deeper exploration of the capability to emulate neutron irradiation with ions. Indeed, while ion irradiations are frequently used to understand phenomena observed during neutron irradiations, the question of their representativeness is often raised.
In this thesis, we aim to understand how the irradiation parameters can affect the characteristics of the a' phase in ODS steels. To do so, various ODS steels will be irradiated under different conditions (flux, dose, temperature, and type of particles, such as ions, neutrons, electrons), and subsequently analyzed at the nanoscale. The a' phase (size, chromium content) obtained for each ion irradiation condition will be compared to the one after neutron irradiation.
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.
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.
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.