Description of the evolution of grain size and dislocation density during ODS steels consolidation
Steels reinforced with a dispersion of nanometric oxides (generally referred to as ODS steels) are currently considered especially as potential material for combustible cladding for 4th generation reactors. Up to now, these materials are conventionally produced by powder metallurgy. The evolution of the microstructure during processing is not well described, yet. Recent work in the laboratory has focused on the evolution of nano-precipitation during processing. The objective of the post-doctoral work is therefore to refine the description of this evolution, more precisely with regard to the grain size and the density of dislocations. This subject combines an experimental approach, through analyses in electron microscopy and X-ray diffraction, and a numerical approach, aiming to define an optimized method for the treatment of the evolution of dislocations.
Stability of the oxide/metal interface of a coated 6061-T6 aluminium alloy
The aluminium alloy, named 6061-T6, is used as core component for the Jules Horowitz French experimental reactor (RJH). In order to improve the corrosion resistance, and to prevent the alloy from wear degradation, a coating is deposited at the surface of the alloy. The coating layer that is 50 µm thin is obtained by oxidation of the aluminium alloy.
The RJH core component will be subjected to neutron irradiation that may modify the microstructure of both the 6061-T6 alloy and the coating layer. Concerning the 6061-T6 alloy, the irradiation damages are well known: neutron irradiation induces the formation of dislocation loops, and causes the dissolution of the nano-precipitates. However, the effect of irradiation on both the coating layer and the interface metal/oxide remains unknown. One of the deleterious effect that may occur in reason of irradiation could be the peeling of the oxide coating and consequently the loss of the corrosion properties. Thus, the understanding of the irradiation response of the coating layer remain a key issue to guarantee a safe use of the coated aluminium alloy. Therefore, the aim of the study is to characterize the irradiation damage of ion irradiated coated aluminium alloys.
Time-resolved in-situ study, by X-ray diffraction under synchrotron radiation, of structural evolutions in a high temperature oxidized zirconium alloys
In certain hypothetical accident situations in pressurized-water nuclear reactors (PWRs), the zirconium alloy cladding of fuel pallets, which constitutes the first barrier for the containment of radioactive products, can be exposed for a few minutes to water vapor. at high temperature (up to 1200 ° C), before being cooled and then quenched with water. The cladding material then undergoes numerous structural and metallurgical evolutions. In order to study these structural evolutions in a precise way, a first experiment campaign was carried out on the BM02 line of the ESRF on a prototype furnace allowing to perfectly control the atmosphere and the temperature. Two tasks will be entrusted to the candidate: continue and finish the analysis of the first experiment(phase fraction determination, residual constraints ...) and prepare a new complementary experimental proposal by mid 2020.
Strudy and processing of C/SiC composites
For different applications, we are looking for materials having superior mechanical properties at high temperature (1000 ° C or higher) and that are resistant to oxidation. The family of ceramic matrix composite materials (CMC), especially C / SiC, seems the most relevant to our needs. However, it is necessary to conduct studies to determine the most efficient solutions among the wide variety of fibrous architectures and possible matrix microstructures, while taking into account the constraints related to available processes and targeted geometries. This work will be conducted in collaboration with other CEA laboratories.
Cluster dynamic simulations of materials under irradiation
Alloys used in nuclear applications are subjected to neutron irradiation, which introduces large amounts of vacancy and interstitial defects. Over time, these defects migrate, recombine and agglomerate with minor alloying elements to form small clusters. This affects the mechanical properties of ferritic steels and weakens them. In this context, the microstuctural evolution is to be simulated using the rate equation cluster dynamic method. However, this approach becomes ineffecient when several minor alloying elements need being taken into account. The difficulty comes from the huge number of cluster variables to describe. The project aims at optimizing the code efficiency on a distributed parallel architecture by implementing parallelized vector and matrix functions from SUNDIALS library. This library is used to integrate the ordinary differential equation describing the reactions between clusters. Another aspect of the work is more theoretical and involves reformulating the non-linear root-finding problem by taking advantage of the reversibility of most chemical reactions. This property should facilitates the implementation of direct and gradients iterative sparse solvers for symmetric definite positive matrices, such as the multi-frontal Cholesky factorization and the conjugate gradient methods, respectively. One avenue of research will consists of combining direct and iterative solvers, using the former as a preconditioner of the latter.
Multiscale Modelling of Radiation Induced Segregation
Irradiation produces in materials excess vacancies and self-interstials that eliminate by mutual recombination or by annihilation at sinks (surfaces, grain boudaries, dislocations).
It sustains permanent fluxes of point defects towards those sinks. In case of preferential transport of one componant of an alloy, the chemical composition is modified in the vicinity of the sinks: a Radiation Induced Segregation (RIS). Its modelling requires a good description of the alloy properties: its driving forces (derived from the thermodynamics) and its kinetic coefficients (the Onsager matrix). The objectif on this project is to combine (i) atomic models (Kinetic Monte Carlo simulations and Self-Consistent Mean Field), fitted on ab initio calculations, that provide the Onsager coeffcients and the driving forces and (ii) a Phase-Field modelling that will give a description of the evolution of the alloy under irradiation at much larger time- and space-scales. The approach will be applied to Fe-Cr and Fe-Cu alloys, already modelled at the atomic scale. RIS will be first modelled near grain boundaries, then near dislocation loops. Special attention will be paid to the effect of elastic stresses on the RIS.
Modelling of interstitial cluster evolution in bcc metals after helium implantation
Under irradiation, structural materials inside nuclear reactors undergo changes in mechanical properties, which result from the formation of point defect clusters, such as cavities (clusters of vacancies) and interstitial dislocation loops (clusters of self-interstitial atoms). Understanding the formation processes of such clusters is thus of prime importance. Recently, three-dimensional interstitial clusters, known as C15 clusters, have been shown theoretically to be highly stable in iron . In order to detect such clusters experimentally, an idea is to make them grow, as shown for dislocation loops after helium implantation .
This approach will be carried out experimentally in various bcc metals in the framework of the ANR project EPigRAPH, in collaboration with Chimie ParisTech, GEMaC and LPS.
In this project, the following modelling tasks will be performed by the postdoc:
- DFT calculations will be done to obtain the energetic properties of point defects and point defect clusters in the bcc metals envisaged in the project.
- These data will then be used to parameterize a kinetic model based on cluster dynamics . This formalism is particularly well adapted to simulate the evolution of point defect clusters over long physical times.
The modelling work will be performed in close collaboration with another postdoc working on the experimental part.
 M. C. Marinica, F. Willaime, J.-P. Crocombette, Phys. Rev. Lett. 108 (2012) 025501
 S. Moll, T. Jourdan, H. Lefaix-Jeuland, Phys. Rev. Lett. 111 (2013) 015503
 T. Jourdan, G. Bencteux, G. Adjanor, J. Nucl. Mater. 444 (2014) 298
Modelling of interstitial cluster evolution in body-centered cubic metals after helium implantation
Under irradiation, structural materials inside nuclear reactors undergo changes in mechanical properties, which result from the formation of point defect clusters, such as cavities (clusters of vacancies) and interstitial dislocation loops (clusters of self-interstitial atoms). Understanding the formation processes of such clusters is thus of prime importance. Recently, three-dimensional interstitial clusters, known as C15 clusters, have been shown theoretically to be highly stable in iron. In order to detect such clusters experimentally, an idea is to make them grow, as shown for dislocation loops after helium implantation. This approach will be carried out experimentally in various bcc metals in the framework of the ANR project EPigRAPH, in collaboration with Chimie ParisTech, GEMaC and LPS.
In this project, the following modelling tasks will be performed by the postdoc:
- Electronic structure calculations will be done to obtain the energetic properties of point defects and point defect clusters in the bcc metals envisaged in the project.
- These data will then be used to parameterize a kinetic model based on cluster dynamics. This formalism is particularly well adapted to simulate the evolution of point defect clusters over long physical times.
Fabrication and characterization of high thermal conductivity SiCf/SiC composites
SiCf/SiC ceramic matrix composites are foreseen candidates for structure materials and claddings in fast neutron reactor of 4th generation. However, their use may be limited because of their too low thermal conductivity in the operating conditions (< 10 W/mK).
SiCf/SiC ceramic matrix composites are now elaborated by chemical vapour infiltration (CVI). In order to improve their thermal conductivity (reduced porosity), it is planned to develop a hybrid elaboration process combining CVI and liquid routes.
The objective of this study is to determine the conditions of elaboration of a SiC matrix by liquid routes and then to characterize the thermo-mechanical behaviour of the hybrid composites, particularly in relation to CVI references.
Multi-scale modelling of the structure and mobility of small defect clusters in metals
Recently, we have proposed a three dimensional periodic structure for self-interstitial clusters in body-centered-cubic metals, as opposed to the conventional two dimensional loop morphology . The underlying crystal structure corresponds to the C15 Laves phase. Using Density Functional Theory and interatomic potential calculations, we have demonstrate that in a–iron these C15 aggregates are highly stable and immobile and that they exhibit large antiferromagnetic moments. They form directly in displacement cascades and they can grow by capturing self-interstitials. They thus constitute an important new element to account for when predicting the microstructural evolution of iron base materials under irradiation.
Despite their low concentration, these clusters are expected to play a crucial role in the behavior of iron and ferritic steels under irradiation and many questions remain to be elucidate: which clusters are the most stable in intermediate sizes, which are the reaction pathways which link the traditional clusters to new ones, how the new clusters interact with the dislocation loops, which are the effects of finite temperatures etc