Investigation of Very High Cycle Fatigue Behavior of 13-4 Martensitic Stainless Steel Manufactured by Laser Metal Deposition: Influence of Microstructure, Post-treatments and Temperature Project
Recent research on 13-4 martensitic stainless steel manufactured by metal additive manufacturing, particularly using the Laser Metal Deposition (LMD) process, has made it possible to obtain materials with good mechanical properties. Following this optimization phase, current work is now focused on studying their Very High Cycle Fatigue (VHCF) behavior, which is a critical criterion for components subjected to repeated loading under severe operating conditions.
Fatigue is one of the main causes of failure in metallic components during service. This thesis therefore aims to understand and model the fatigue behavior of LMD-produced 13-4 steel. The work will investigate the influence of microstructure, thermomechanical treatments, and testing conditions on crack initiation and propagation during mechanical loading.
Experimental investigations will be carried out using ultrasonic fatigue testing devices. Failure mechanisms will be analyzed through multi-scale characterization techniques such as EBSD, SEM, and TEM. The final objective is to develop a predictive model capable of estimating the service life of components under operating conditions.
Characterization and Understanding of Degradation Mechanisms in Encapsulation Materials Used in New-Generation Silicon Photovoltaic (PV) Modules under Humidity and UV Stress
New-generation photovoltaic technologies (TOPCon, SHJ, tandems) are particularly sensitive to environmental stressors, including humidity (Damp Heat, DH), UV radiation, and thermal cycling. These stresses accelerate the degradation of encapsulation materials (EVA, POE, TPO), leading to performance losses in modules—such as reduced transparency, delamination, metal contact corrosion, and Potential-Induced Degradation (PID). Despite their increasing adoption, these novel encapsulants lack long-term durability data, while widely used EVA exhibits premature aging (degradation after 10–15 years of exposure). The combined degradation mechanisms (DH + UV + temperature) remain understudied, yet they reflect real-world exposure conditions
This thesis aims to identify and understand the physicochemical degradation mechanisms of polymer encapsulants under coupled stressors, focusing on:
- Multi-scale analysis (chemical structure, optical properties, microstructure) of materials during accelerated aging
- Development of an experimental protocol replicating real-world conditions (coupled DH/UV stress) to assess material resilience
- Study of additive roles in encapsulant degradation, including UV absorbers, peroxides etc.
An electrochemical flow microreactor for a greener synthesis of gold nanoparticles
Gold nanoparticles (AuNPs) possess unique electronic, photonic, and chemical properties of invaluable interest in a variety of medical and technological applications. They are typically produced by controlled chemical precipitation from a salt solution to achieve the precise size control critical for most applications. Continuous flow microreactors, which efficiently mix the salt solution and the reducing agent, are known to offer improved size control. However, even in these reactors, the smallest AuNPs can only be formed using powerful reducing agents that are harmful to human health or the environment. We propose to minimize their impact and to develop a more resource-efficient process by inserting an electrochemical cell into the reactor to form the reducing agent in-situ in the adjusted amount necessary to produce the desired AuNPs.
Your goal will be to test and adapt continuous-flow electrochemical cells for the synthesis of AuNPs, exploring various electrochemical reactions and cell designs. You will also explore the use of several capping agents of biological interest. A careful examination of AuNPs characteristics (size, interfacial and optical properties, etc.) will guide you in this research.
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).
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.
Characterization of radiolytic mechanisms in tritiated water–zeolite systems under storage conditions
The operation of the tritium facilities at Valduc generates low-activity tritiated liquid effluents, which are stored in an adsorbed form on 4A zeolite for operational reasons. Understanding the mechanisms of self-radiolysis of this confined water is essential for optimizing storage conditions.
Several PhD projects have already investigated these mechanisms by combining experiments and modelling. Early work showed that below 13% hydration, the radiolytic gases H2 and O2 can recombine within the zeolite. Subsequent studies, based on DFT calculations and molecular dynamics, identified the adsorption sites and the mobility of the gases. They revealed a hydration threshold (13–15%) above which gas diffusion becomes very low, consistent with the experimentally observed cessation of recombination. However, these simulations rely on idealized models.
The new proposed PhD aims to shift the project back toward experimental work in order to better reflect real storage conditions. It will begin with a detailed characterization of the zeolite used industrially. Water–zeolite reservoirs will then be irradiated to simulate the effect of tritium, and analyzed by NMR and possibly by Electron Spin Resonance (ESR) to detect reactive species. The experimental results may feed into a macroscopic model (Kinetic Monte Carlo, KMC), also developed previously, to predict the evolution of the system and identify possible optimizations for storage. The work will be carried out mainly at the NIMBE laboratory (CEA-CNRS), with simulation collaboration in Besançon and regular exchanges with CEA Valduc.
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.
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.