Thermodynamic and Structural properties of alkaline and/or alkaline-earth, actinyl(VI) with carbonates
The triscarbonato uranyl solid phases containing alkaline (Alk = Na, K) alkaline earth (Ae = Mg, Ca) are known since mid-19th century. The evidence of equivalent complexes in solution Ca-UO2-CO3 has only occurred during the 1990s. Their importance is recognised as major in certain natural environments, from surface waters down to deep underground disposal sites of nuclear waste disposal. Using theoretical calculations, the formation of Alk-Ae-U(VI)-CO3 complexes has been proposed but without any experimental confirmation. This PhD work, in collaboration between Laboratory for Analytical, Nuclear, Isotopic, and Elementary development (DES/ISAS/DRMP/SPC/LANIE) in the Saclay centre, and the Laboratory for the Ligand Actinide Interactions (DES/ISEC/DMRC/SPTC/LILA) in the Marcoule centre, is proposing to couple the advantages of time-resolved fluorescence (or luminescence) of uranium(VI) complexes (LANIE) and theoretical calculations and X-ray spectroscopy (LILA) to guide the experiments to evidence the eventual Alk-Ae-U(VI)-CO3 complexes in solution.
The candidate must already dominate, or quickly acquire, a competence in solution chemistry with a strong tendency upon the acquisition of thermodynamic constants and functions of reactions, with a knowledge of the concepts linked to the ionic strength correction and SIT. Adaptability to, of even better knowledge, of different spectroscopic techniques will be an undeniably asset. The inclination to multidisciplinary team work will be essential.
Hydrogen transport and trapping in austenitic alloys coupling experiments and simulations.
Molecular hydrogen H2 is an alternative energy carrier to traditional fossil fuels, gas or oil. It meet the current energy and environmental challenges, i.e. the need to store greenhouse gases free energy produced by intermittent means such as wind turbines or solar panel. Nevertheless, its safe storage and transportation is one of the keys to its use. The containers or pipes that carry the hydrogen must be leaktight and maintain their integrity over time, for both economical and safety reasons. Understanding and predicting the behavior of hydrogen in container/pipeline alloys and the associated mechanical degradation – such as embrittlement – is therefore crucial for the development of the hydrogen industry. These issues are also generic to all alloys exposed to a source of hydrogen, in corrosion or in the metallurgical industries where the hydrogen simply comes from contact with water, or in the oil&gas industry where hydrogen comes from hydrogen sulphides present in hydrocarbons.
If many experimental works have identified hydrogen embrittlement as the origin of the degradation of alloys exposed to hydrogen, large gray areas still remain on the mechanisms at work due to experimental difficulties and the great variability of the observed phenomena. In addition, the transport and trapping of hydrogen prior to mechanical degradation are poorly known and poorly documented at the nanoscale.
The objective of the thesis is to explore the mechanisms of hydrogen trapping / transport in austenitic materials, as well as its distribution in volume, prior to cracking in order to be able to report and explain the experimental observations.
To achieve this objective, the thesis work will be dedicated to the study of pure nickel, a model system for the austenite phase. The study will be carried out in three stages: (i) thermodesorption measurements and (ii) atomic scale simulations using molecular dynamics, both feeding chemical kinetics modeling coupled with Fick's law at the mesoscopic scale.
Development of a modeling tool for corrosion in porous media
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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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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.
Experimental study and numerical simulation of deformation mechanisms and mechanical behavior of zirconium alloys after irradiation
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.
Chemical and mechanical properties of N-A-S-H aluminosilicates of geopolymer
Management of low- and medium-level nuclear waste relies primarily on cements, but their limitations with regard to certain types of waste (reactive metals, oil) require the exploration of new, more effective materials. Geopolymers, particularly those composed of hydrated sodium aluminosilicates (Na2O–Al2O3–SiO2–H2O, or N–A–S–H), appear to be a promising alternative thanks to their chemical compatibility with certain types of waste.
However, despite the growing interest in geopolymers, scientific obstacles remain: 1) The available thermodynamic data on N-A-S-H is still incomplete, making it difficult to predict their long-term stability via modeling, 2) The role of their atomic structure in regard to their reactivity remains unclear, and 3) The links between chemical composition (in terms of Si/Al ratio) and mechanical properties are not established, limiting the representativeness of the models created.
By combining experimentation and modeling in order to link atomic structure and properties, this thesis aims to obtain robust and novel data on the chemical and mechanical properties of N-A-S-H. The thesis is organized around three major objectives: 1) determining the impact of N-A-S-H composition on dissolution and establishing thermodynamic solubility constants, 2) characterizing their atomic structure (aluminols, silanols, and hydrated environments) using advanced NMR spectroscopy, and 3) linking their mechanical properties, measured by nanoindentation, to their structure and composition using molecular dynamics modeling.
Study of oxygen and hydrogen diffusion processes in pre- and post-transitional oxide layers formed on zirconium alloys
The corrosion mechanisms of zirconium alloys in pressurised water reactors are still a subject of debate more than half a century after the first research on this material. The literature reports two distinct mechanisms for the transport of diffusing species in oxide layers: one favours the molecular diffusion of oxygen and hydrogen through interconnected nanopore channels during the pre-transient regime, while the other favours diffusion via short circuits (grain boundaries, etc.) in the oxide layer. In the latter case, the oxide layer is considered to be relatively homogeneous and impermeable to the oxidising medium, in this case the water in the primary circuit. On the other hand, the first interpretation is based on the principle that there is a layer that is permeable to the medium due to an interconnected network of nanopores, even during the pre-transient regime, with the density of percolated nanopores increasing over time.
Technically speaking, how can we decide between these two divergent interpretations in terms of the diffusion mechanism, which consequently leads to different solutions for protection against degradation? What is the reaction mechanism that ultimately leads to the hydration of Zr alloys and their oxidation?
To address this challenge, we will explore diffusion processes by studying the dissociation-recombination rates of molecular species at different temperatures in equi-isotopic gas mixtures such as H2/D2, 18O2/16O2, H218O/D216O, H218O/D2, etc., using an experimental device equipped with a mass spectrometer that tracks the molecular species of interest in real time.