Silicon carbide and nanostructured ferritic alloy (SiC-NFA) composites have the potential to maintain the outstanding high temperature corrosion and irradiation resistance and enhance the mechanical integrity for nuclear cladding. However, the formation of detrimental silicide phases due to reaction between SiC and NFA remains a major challenge. By introducing a carbon interfacial barrier on NFA (C@NFA), SiC-C@NFA composites are investigated to reduce the reaction between SiC and NFA. In a similar way, the effect of chromium carbide (Cr3C2) interfacial barrier on SiC (Cr3C2@SiC) is also presented for Cr3C2@SiC-NFA composites. Both the coatings were successful in suppressing silicide formation. However, despite the presence of coatings, SiC was fully consumed during spark plasma sintering process. TEM and EBSD investigations revealed that spark plasma sintered SiC-C@NFA and Cr3C2@SiC-NFA formed varying amounts of different carbides such as (Fe,Cr)7C3, (Ti,W)C and graphite phases in their microstructure. Detailed microstructural examinations after long term thermal treatment at 1000oC on the microstructure of Cr3C2@SiC-NFA showed precipitation of new (Fe,Cr)7C3, (Ti,W)C carbides and also the growth of existing and new carbides. The results were successfully explained using ThermoCalc precipitation and coarsening simulations respectively. The oxidation resistance of 5, 15 and 25 vol% SiC@NFA and Cr3C2@SiC-NFA composites at 500-1000oC temperature under air+45%water vapor containing atmosphere is investigated. Oxidation temperature effects on surface morphologies, scale characteristics, and cross-sectional microstructures were investigated and analyzed using XRD and SEM. SiC-C@NFA showed reduced weight gain but also showed considerable internal oxidation. Cr3C2@SiC-NFA composites showed a reduction in weight gain with the increasing volume fraction of Cr3C2@SiC (5, 15 and 25) without any indication of internal oxidation in the microstructure. 25 vol% SiC-C@NFA and 25 vol% Cr3C2@SiC-NFA showed over 90% and 97% increase in oxidation resistance (in terms of weight gain) as compared to NFA. The results were explained using the fundamental understanding of the oxidation process and ThermoCalc/DICTRA simulations. Finally, the irradiation performance of SiC-C@NFA and Cr3C2@SiC-NFA composites was assessed in comparison with NFA using state-of-the-art TEM equipped with in-situ ion irradiation capability. Kr++ ions with 1 MeV energy was used for irradiation experiments. The effect of ion irradiation was recorded after particular dose levels (0-10 dpa) at 300oC and 450oC temperatures. NFA sample showed heavy dislocation damage at both 300oC and 450oC increasing gradually with dose levels (0-10 dpa). Cr3C2@SiC-NFA showed similar behavior as NFA at 300oC. However, at 450oC, Cr3C2@SiC-NFA showed remarkably low dislocation loop density and loop size as compared to NFA. At 300oC, microstructures of NFA and Cr3C2@SiC-NFA show predominantly 1/2<111> type dislocation loops. At 450oC, NFA showed predominantly <100> type loops, however, Cr3C2@SiC-NFA composite was still predominant in ½<111> loops. The possible reasons for this interesting behavior were discussed based on the large surface sink effects and enhanced interstitial-vacancy recombination at higher temperatures. The molecular dynamics simulations did not show considerable difference in formation energies of ½<111> and <100> loops for NFA and Cr3C2@SiC-NFA composites. The additional Si element in the SiC-NFA sample could have been an important factor in determining the dominant loop types. SiC-C@NFA composites showed heavy dislocation damage during irradiation at 300oC. At 450oC, SiC-C@NFA showed high dislocation damage in thicker regions. Thinner regions near the edge of TEM samples were largely free from dislocation loops. The precipitation and growth of new (Ti,W)C carbides were observed at 450oC with increasing irradiation dose. (Fe,Cr)7C3 precipitates were largely free from any dislocation damage. Some Kr bubbles were observed inside (Fe,Cr)7C3 precipitates and at the interface between α-ferrite matrix and carbides ((Fe,Cr)7C3, (Ti,W)C). The results were discussed using the fundamental understanding of irradiation and ThermoCalc simulations. / Doctor of Philosophy / With the United Nations describing climate change as 'the most systematic threat to humankind', there is a serious need to control the world's carbon emissions. The ever increasing global energy needs can be fulfilled by the development of clean energy technologies. Nuclear power is an attractive option as it can produce low cost electricity on a large scale with greenhouse gas emissions per kilowatt-hour equivalent to wind, hydropower and solar. The problem with nuclear power is its vulnerability to potentially disastrous accidents. Traditionally, fuel claddings, rods which encase nuclear fuel (e.g. UO2), are made using zirconium based alloys. Under 'loss of coolant accident (LOCA) scenarios' zirconium reacts with high temperature steam to produce large amounts of hydrogen which can explode. The risks associated with accidents can be greatly reduced by the development of new accident tolerant materials. Nanostructured ferritic alloys (NFA) and silicon carbide (SiC) are long considered are leading candidates for replacing zirconium alloys for fuel cladding applications. In this dissertation, a novel composite of SiC and NFA was fabricated using spark plasma sintering (SPS) technology. Chromium carbide (Cr3C2) and carbon (C) coatings were employed on SiC and NFA powder particles respectively to act as reaction barrier between SiC and NFA. Microstructural evolution after spark plasma sintering was studied using advanced characterization tools such as scanning electron microscopy (SEM), electron backscattered diffraction (EBSD), transmission electron microscopy (TEM) and energy dispersive spectroscopy (EDS) techniques. The results revealed that the Cr3C2 and C coatings successfully suppressed the formation of detrimental reaction products such as iron silicide. However, some reaction products such as (Fe,Cr)7C3 and (Ti,W)C carbides and graphite retained in the microstructure. This novel composite material was subjected to high temperature oxidation under a water vapor environment to study its performance under the simulated reactor environment. The degradation of the material due to high temperature irradiation was studied using state-of-the-art TEM equipped with in-situ ion irradiation capabilities. The results revealed excellent oxidation and irradiation resistance in SiC-NFA composites as compared to NFA. The results were discussed based on fundamental theories and thermodynamic simulations using ThermoCalc software. The findings of this dissertation imply a great potential for SiC-NFA based composites for future reactor material designs.

La fragilisation des aciers de cuve des réacteurs nucléaires sous irradiation aux neutrons est le facteur limitant la durée de vie des centrales nucléaires françaises. Ceci est dû au mouvement des dislocations qui se trouve être entravé par des amas de Cu, P, Si, Mn et Ni. Plus particulièrement, les amas induits de Mn et de Ni sont à l'origine d'un durcissement significatif à forte dose. Afin de prédire la dégradation des propriétés mécaniques, les expériences sont généralement réalisées à l'aide d'accélérateurs de particules. Cependant, les flux d'irradiation atteints sont compris entre 10⁻⁴ 10 ⁻ ⁶ dpa/s⁻ ¹, tandis qu'il est limité à 10⁻ ¹ ⁰ dpa/s⁻ ¹ dans les réacteurs de puissance actuels. Ce point est essentiel étant donné que le dommage d'irradiation dépend du flux de particules incidentes. La transférabilité ion/neutron constitue donc la problématique centrale. Celle-ci a été étudiée dans les alliages austénitiques seulement. Ce travail de thèse se propose donc d'étudier, dans des alliages ferritiques, l'effet du flux d'irradiation sur l'endommagement dans deux alliages différents : le Fe-Ni et le Fe-Mn, dans le but d'évaluer également l'effet de chaque soluté sur la microstructure obtenue après irradiation.Les alliages ont été analysés expérimentalement par Microscopie Electronique en Transmission (TEM), Microscopie Electronique à Balayage par Transmission (STEM) couplée à l'Analyse Dispersive en Energie des Rayons-X (EDS) et à la Spectroscopie de Perte d'énergie des Electrons (EELS), ainsi que par Sonde Atomique Tomographique (APT).Les irradiations ont été réalisées avec des ions Fe³⁺ de 2 MeV et des ions Fe⁹⁺ de 27 MeV, à 400°C, à des taux de dommage de 10⁻⁴ et 10⁻ ⁶ dpa/s⁻ ¹ respectivement, jusqu'à un même dommage de 1.2 dpa.Les résultats obtenus montrent que le Ni et le Mn ont des comportements sous irradiation très différents en termes de nature de nano-défauts créés.Des irradiations aux particules légères ont également été réalisées de manière à apprécier l'effet des cascades de déplacement.Enfin, une irradiation séquentielle, en deux étapes, a été effectuée à l'aide d'ions Fe⁹⁺ à température ambiante, puis de protons à 400°C, dans le but d'isoler la contribution au durcissement des amas de défauts ponctuels de celle des zones enrichies en soluté. / Reactor pressure vessel (RPV) steel embrittlement under neutron irradiation is the main lifetime limiting factor of nuclear reactors. This is due to the impeding of dislocation glide by nanometric clusters composed of Cu, P, Si, Mn and Ni. More specifically, radiation induced Mn and Ni enriched clusters cause a significant hardening at high dose. To predict this change in mechanical properties, particle accelerator based experiments are conducted. However, the achieved flux ranges between 10⁻⁴ and 10 ⁻ ⁶ dpa/s⁻ ¹, whereas it is limited to 10⁻ ¹ ⁰ dpa/s⁻ ¹ in modern nuclear power technologies. This point is of high importance since radiation damage highly depends on irradiation flux. The reproducibility ion-neutron is thus the key point. It has been studied in austenitic steels but little is known regarding any dose rate dependence in ferritic alloys. Therefore, this thesis focuses on the effect of ion fluxes on radiation damage in two different alloys : Fe-Ni and Fe-Mn in order to investigate, additionally, the effects of each solute on the microstructure after irradiation.The alloys were experimentally investigated using conventional Transmission Electron Microscopy, Scanning Transmission Electron Microscopy coupled to Energy Dispersive X-ray Spectroscopy and Electron Energy Loss Spectroscopy and by Atom Probe Tomography.Irradiations were performed with 2 MeV Fe³⁺ ions and 27 MeV Fe⁹⁺ ions at 400°C at a nominal damage rate of 10⁻⁴ and 10⁻ ⁶ dpa/s respectively, up to a nominal displacement damage of 1.2 dpa. The detailed analysis shows that Ni and Mn behave in a very different way in terms of nano-defects formed under irradiation.Besides, light particle irradiations were also performed in order to ascertain the cascade effects.Finally, a two-series irradiation was carried out using Fe ions at room temperature and protons at 400°C, to isolate the contribution of point defect clusters to hardening from that of solute enriched zones.

<p>An investigation of the influence of three alloying elements Chromium, Phosphorus, and Nitrogen with the solute types of oversized substitutional, undersized substitutional, and interstitial on the irradiation induced microstructural evolution and hardening</p>

Le développement à grande échelle des projets de véhicules électriques à pile àcombustible nécessite le déploiement d’infrastructures de transport et de stockaged’Hydrogène gazeux. La conception de ces structures et la sélection des matériaux nécessitede s’affranchir des risques liés à la fragilisation par l’Hydrogène des alliages métalliques. Cephénomène est bien décrit depuis plusieurs décennies, mais les mécanismes élémentaires àl’origine de ce mode d’endommagement restent controversés, notamment par manque demodèles quantitatifs. Plus précisément, le rôle de la déformation (micro-)plastique en pointede défaut sur le piégeage et l’endommagement par l’hydrogène, s’il est bien démontréexpérimentalement dans de nombreux systèmes, reste mal pris en compte dans les modèlesmicro-mécaniques. Le centre SMS de l’ENSM.SE a proposé des approches originales demodélisation des interactions hydrogène – dislocations, qui ont pu être validéesexpérimentalement dans des matériaux modèles de structure cubique à faces centrées. Cette thèsese propose d’appliquer une démarche semblable dans des alliages de structure cubiquecentrée. On mettra en oeuvre des essais de déformation sur des matériaux modèles pré-chargésen hydrogène, des modèles semi-analytiques et des observations des structures de déformationen microscopie électronique à transmission. / The development of electrical vehicles powered by hydrogen fuel cells requires the large scaledeployment of hydrogen storage and transport infrastructures. This in turn requires theassessment of the sensitivity of structural materials to hydrogen embrittlement phenomena.These damage modes, while being well described experimentally for since several decades,are still highly debated when it comes to elementary physical processes, mainly because of thelack of quantitative models for these elementary processes. More precisely, the role of the(micro-)plasticity developing at the tip of structural defects, while being well establishedexperimentally, is still poorly accounted for in the available micro-mechanical models. TheScience of Materials and Structures division of ENSM.SE already proposed originalmodelling approaches for hydrogen – dislocation interactions, that have been experimentallyvalidated in face-centred cubic materials. This project aims at applying the same type ofapproach to body-centred cubic metals. This will be achieved by means ofdeformation tests on hydrogen-charged model body centred cubic alloys, investigations of thedislocation microstructures by transmission electron microscopy and the development ofsemi-analytical models of hydrogen-dislocation interactions.