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.

Effect of gravity on agitation within a turbulent bubbly flow in a channel

Understanding two-phase flows and the boiling phenomenon is a major challenge for the CEA, for both the design and safety of nuclear power plants. In a Pressurized Water Reactor (PWR), the heat generated by the nuclear fuel is transferred to the water in the primary circuit. Under accident conditions, the water in the primary circuit can enter a nucleate boiling regime, or even evolve to a boiling crisis. While the phenomenon of boiling is the subject of numerous studies, the dynamics of the generated bubbles also receive special attention at the CEA. This thesis will focus on the coupling between the turbulence generated by a shear flow and the agitation induced by the bubbles. Its originality lies in the study of the effect of gravity, achieved by tilting the channel, a parameter that can generate complex flow regimes.
This experimental work will be based on the new CARIBE facility at CEA Saclay. The PhD student's mission will be to characterize the different flow regimes and then to conduct a detailed study of the flow by implementing specific metrology (including Particle Image Velocimetry (PIV), hot-film anemometry, and optical probes). Conducted within the LE2H laboratory, the project will benefit from a close collaboration with the LDEL (CEA Saclay) and the IMFT (Toulouse). The PhD student will work in a dynamic environment with other PhD students and will present their work at national and international conferences.
We are looking for a candidate with a background in fluid mechanics and a strong interest in experimental work (a Master's thesis internship is possible). This PhD offers the opportunity to develop expertise in instrumentation, data analysis, and turbulent two-phase flows—skills that are highly valued in the energy, industrial, and academic research sectors.

Localised solidifications in Molten Salt Reactors

In a Molten Salt Reactor (MSR), the nuclear fuel is a liquid, high-temperature salt which acts as its own coolant. Some accidental transients (over-cooling of the fuel, leak) may cause localised solidifications of the fuel salt in the core. These solidifications will have in turn an impact on the salt flow in the core, as well as its neutronic behavior, and could lead to a localised over-heating of the core vessel. Such transients are not well studied, although they have a major impact on the safety and design of an MSR.
The objective of the PhD is to study different accidental transients that would lead to localised solidifications, and to study their impact on the neutronics and thermal-hydraulics of the core. These analyses will require the use of multiphysics, MSR-adapted numerical tools, such as the CFD code TrioCFD and its extensions TRUST-NK (neutronics) and Scorpio (reactive transport), as well as the deterministic neutronic code APOLLO3. In order to balance precision and computation time, different models will be tested, depending on the transient studied: 1D/ turbulent 3D (RANS, LES) models for thermal-hydraulics ; diffusion / SPn transport / Sn transport for neutronics.

Numerical modelling of large ductile crack progagation and assessment of margins comparing to engineering approach

Predicting failure modes in metal structures is an essential step in analyzing the performance of industrial components where mechanical elements are subjected to significant stress (e.g., nuclear power plant components, pipelines, aircraft structural elements, etc.). To perform such analyses, it is essential to correctly simulate the behavior of a defect in ductile conditions, i.e., in the presence of significant plastic deformation before and during propagation.
Predictive numerical simulation of ductile tearing remains an open scientific and technical issue despite significant progress made in recent years. The so-called local approach to fracture, notably the Gurson model (and its modified version GTN), is widely used to model ductile tearing. However, its use has limitations: significant computation time, simulation stoppage due to the presence of completely damaged elements in the model, and non-convergence of the result when the mesh size is reduced.
The aim of this thesis is to develop the ductile tear simulation model used at LISN so that it can be applied to large crack propagation on complex structures. It also aims to compare the results obtained with engineering methods that are simpler to implement.

Fatigue crack growth modelling with residual stress - Improvement of the Gtheta method

Residual stresses are self-balanced stress fields found in certain mechanical components in the absence of external loading. Caused by welding, for example, these stresses can potentially affect the behaviour of the structure and its resistance to fracture. When demonstrating the integrity of a mechanical component, particularly in the context of nuclear safety, it is crucial to precisely understand the role of these stress fields on the component's resistance. In the case of fatigue crack propagation, to accurately model all the phenomena involved (stress redistribution, evolution of plasticity, closure effect), it will be necessary to improve numerical tools, such as meshing and crack propagation methods (AMR, X-FEM...) and the J-integral interpolation in the case of through-cracks (Gtheta method). The thesis will consist of two complementary parts: (a) numerical development aimed at improving the Gtheta method in Castem FE code, associated with a 3D crack propagation modelling using AMR, and (b) continuation of component scale tests on fatigue crack propagation in different configurations of residual stress fields.

Effect of water radiolysis on the hydrogen absorption flux by austenitic stainless steels in the core of a nuclear pressurized water reactor

In pressurized water nuclear reactors, the core components are exposed to both corrosion in the primary medium, pressurized water at around 150 bar and 300°C, and to neutron flux. The stainless steels in the core are damaged by a combination of neutron bombardment and corrosion. In addition, radiolysis of the water can have an impact on the mechanisms and kinetics of corrosion, the reactivity of the medium and, a priori, the mechanisms and kinetics of hydrogen absorption by these materials. This last point, which remains unexplored, may prove problematic, as hydrogen in solid solution in steel can lead to changes in (and degradation of) the mechanical properties of the steel or induce premature cracking of the part. The pioneering work developed in this highly experimental thesis will focus on the impact of radiolysis phenomena on the mechanisms and kinetics of corrosion and, above all, hydrogen pick-up in 316L stainless steel exposed to the primary environment under irradiation. Hydrogen will be traced by deuterium, and neutron irradiation simulated by electron irradiation on particle accelerators. An existing permeation cell will be modified into a unique setup to allow in operando measurement by mass spectrometry of the deuterium permeation flux through a sample exposed to the simulated primary water under radiolysis conditions. The distribution of hydrogen in the material, as well as the nature of the oxide layers formed, will be analysed in detail using state-of-the-art techniques available at the CEA and in partner laboratories. The doctoral student will ultimately be required to (i) identify the mechanisms involved (corrosion and hydrogen entry), (ii) estimate their kinetics and (iii) model the evolution of hydrogen flux in the steel in connection with radiolysis activity.

Kinetics of segregation and precipitation in Fe-Cr-C alloys under irradiation : coupling magnetic, chemical and elastic effects

Ferritic steels are being considered as structural materials in future fission and fusion nuclear reactors. These alloys have highly original properties, due to the coupling between chemical, magnetic and elastic interactions that affect their thermodynamic properties, the diffusion of chemical species and the diffusion of point defects in the crystal. The aim of the thesis will be to model all of these effects at the atomic scale and to integrate them into Monte Carlo simulations in order to model the segregation and precipitation kinetics under irradiation, phenomena that can degrade their properties in use. The atomic approach is essential for these materials, which are subjected to permanent irradiation and for which the laws of equilibrium thermodynamics no longer apply.

The candidate should have a good background in statistical physics or materials science, and be interested in numerical simulations and computer programming. The thesis will be carried out at CEA Saclay's physical metallurgy laboratory (SRMP), in a research environment with recognised experience in multi-scale modelling of materials, with around fifteen theses and post-doctoral contracts in progress on these topics.

A Master 2 internship on the same subject is proposed for spring 2025 and is highly recommended.

Numerical simulation of turbulence models on distorted meshes

Turbulence plays an important role in many industrial applications (flow, heat transfer, chemical reactions). Since Direct Simulation (DNS) is often an excessive cost in computing time, Reynolds Models (RANS) are then used in CFD (computational fluid dynamics) codes. The best known, which was published in the 70s, is the k - epsilon model.
It results in two additional non-linear equations coupled to the Navier-Stokes equations, describing the transport, for one, of turbulent kinetic energy (k) and, for the other, of its dissipation rate (epsilon). ). A very important property to check is the positivity of the parameters k and epsilon which is necessary for the system of equations modeling the turbulence to remain stable. It is therefore crucial that the discretization of these models preserves the monotony. The equations being of convection-diffusion type, it is well known that with classical linear schemes (finite elements, finite volumes, etc ...), the numerical solutions are likely to oscillate on distorted meshes. The negative values of the parameters k and epsilon are then at the origin of the stop of the simulation.
We are interested in nonlinear methods allowing to obtain compact stencils. For diffusion operators, they rely on nonlinear combinations of fluxes on either side of each edge. These approaches have proved their efficiency, especially for the suppression of oscillations on very distorted meshes. We can also take the ideas proposed in the literature where it is for example described nonlinear corrections applying on classical linear schemes. The idea would be to apply this type of method on the diffusive operators appearing in the k-epsilon models. In this context it will also be interesting to transform classical schemes of literature approaching gradients into nonlinear two-point fluxes. Fundamental questions need to be considered in the case of general meshes about the consistency and coercivity of the schemes studied.
During this thesis, we will take the time to solve the basic problems of these methods (first and second year), both on the theoretical aspects and on the computer implementation. This can be done in Castem, TrioCFD or Trust development environments. We will then focus on regular analytical solutions and application cases representative of the community.

Staggered schemes for the Navier-Stokes equations with general meshes

The simulation of the Navier-Stokes equations requires accurate and robust numerical methods that
take into account diffusion operators, gradient and convection terms. Operational approaches have
shown their effectiveness on simplexes. However, in some models or codes
(TrioCF, Flica5), it may be useful to improve the accuracy of solutions locally using an
error estimator or to take into account general meshes. We are here interested in staggered schemes.
This means that the pressure is calculated at the centre of the mesh and the velocities on the edges
(or faces) of the mesh. This results in methods that are naturally accurate at low Mach numbers .
New schemes have recently been presented in this context and have shown their
robustness and accuracy. However, these discretisations can be very costly in terms of memory and
computation time compared with MAC schemes on regular meshes
We are interested in the "gradient" type methods. Some of them are based on a
variational formulation with pressure unknowns at the mesh centres and velocity vector unknowns on
the edges (or faces) of the cells. This approach has been shown to be effective, particularly in terms of
robustness. It should also be noted that an algorithm with the same degrees of freedom as the
MAC methods has been proposed and gives promising results.
The idea would therefore be to combine these two approaches, namely the "gradient" method with the same degrees of freedom as MAC methods. Initially, the focus will be on recovering MAC schemes on regular meshes. Fundamental
questions need to be examined in the case of general meshes: stability, consistency, conditioning of
the system to be inverted, numerical locking. An attempt may also be made to recover the gains in
accuracy using the methods presented in for discretising pressure gradients.
During the course of the thesis, time will be taken to settle the basic problems of this method (first and
second years), both on the theoretical aspects and on the computer implementation. It may be carried
out in the Castem, TrioCFD, Trust or POLYMAC development environments. The focus will be on
application cases that are representative of the community.

Development of a transport chemistry model for spent fuel in deep geological disposal under radiolysis of water

The direct storage of spent fuel (SF) represents a potential alternative to reprocessing as a means of managing nuclear waste. The direct storage of spent fuel in a deep geological environment presents a number of scientific challenges, primarily related to the necessity of developing a comprehensive understanding of the processes involved in the dissolution and release of radionuclides. The objective of this thesis is to develop a comprehensive scientific model that can accurately describe the intricate physico-chemical processes involved, such as the radiolysis of water and the interaction between irradiated fuel and its surrounding environment. The objective is to propose an accurate reactive transport model to enhance long-term predictions of storage performance. This thesis employs a back-and-forth process between modeling and experimentation, with the goal of refining the understanding of alteration mechanisms and validating hypotheses with experimental data. Based on existing models, such as the operational radiolytic model, the work will propose improvements to reduce the current simplifying assumptions. The candidate will contribute to major industrial and societal issues related to nuclear waste management and will help to provide solutions to the associated safety issues.

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