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International audience; SiC/SiC ceramic matrix composites are promising cladding candidates to improve the accident tolerance of the fuel in pressurized water reactors. Indeed, their excellent mechanical properties at high temperatures (above 1200°C) would give them additional margins to face a loss of coolant. In this work, the mechanical behavior of a filament wound tube under cyclic tensile stress is performed in order to investigate the impact of two types of manufacturing defects on the mechanical behavior. The defects studied correspond to local modifications of the composite microstructure. For both classes of defects, the same methodology is used, consisting in performing mechanical cyclic tensile tests in the tube axis direction with post-analysis to assess fine parameters. The results allow characterizing the elastic behavior, as well as, the behavior during matrix microcracking through the consideration of unload-reload cycles. Microstructiral analyses are performed to establish the mechanical properties/microstructures relationship.

The purpose of this guide is to outline the basic technical requirements for performing thermal diffusivity measurements of solid materials at high temperature (between 1500 ��C and 3000 ��C) by applying the laser flash method., The project 17IND11 Hi-TRACE has received funding from the EMPIR programme co-financed by the Participating States and from the European Union's Horizon 2020 research and innovation programme.

International audience; SiC/SiC composites are promising candidates for nuclear fuel cladding due to their outstanding mechanical resistance and properties at high temperatures (>1200°C). During the manufacturing process, a grinding phase of both the external and internal surfaces is required to respect the dimensional specifications which results in local modifications. In this work, the impact of the grinding process on the mechanical behavior is studied. For that purpose, several batches have been produced with various surface finishes, differing in their level of internal and external grindings (raw CVI tubes, tubes ground at the inner and outer sides with a fine control of grinded thickness). On the one hand, the mechanical behavior was investigated by performing uniaxial cyclic tensile tests and monitoring several inherent parameters. On the other hand, the microstructure was studied in the initial and final states by cross-sectional microscope observations and scanning electron microscopy performed on fracture surfaces. To establish microstructure/properties relationships, a unidirectional model is used. It has been enriched to take into account the investigated CVI-SiC/SiC tubes specificities. Today, the implemented model allows understanding the impact of local defects induced by the different levels of grinding on the tubes mechanical behavior.

International audience; SiC/SiC composites have aroused interest for fusion and fission nuclear applications due to their thermomechanical properties at high temperature and behavior under irradiation (low swelling, neutron transparency, …), essentially as core materials. In this aim, the CEA has been working on the manufacture and the development of tubular SiC/SiC composites cladding elements for more than a decade. During this period, an optimization of the mechanical properties and the mastering of the manufacturing process have been conducted. Third generation SiC fibers, Hi-Nicalon S and Tyranno SA3, and high-purity CVI pyrocarbon interphase and SiC matrix have been employed for their expected behavior under irradiation. However, the latter consequences have to be properly evaluated. Different types of materials have then been produced for several irradiations. Firstly, the manufacturing process and the materials characterization will be described as the irradiations conditions. The CROCUS irradiation was aimed to determine the mechanical behavior of SiC/SiC minicomposites and of various brazing fillers while the CEDRIC irradiation had as objective to follow their creep at high temperature (600 – 1000°C). The TIRAMISU purpose was to evaluate the effect of high-doses (> 100 dpa) at the Sodium Fast Reactor temperature (550°C). The aftermath of these irradiations on the materials dimensions (swelling) as well as on its structure (fibers, interphase) and mechanical properties will then be evaluated.

International audience; In response to Fukushima Daiichi accident, the nuclear industry across the world initiated research programs aimed at developing a new generation of safer fuel systems which will maximize safety and provide economic benefits to utilities. Framatomes strategy regarding Enhanced Accident tolerant Fuel (EATF) relies on a two-phased approach enabling the balancing of benefits with speed to market. The first phase consists of a near term evolutionary solution improving safety and fuel cycle economics while being compliant with insertion of batches by 2025. A second, longer-term solution, SiC/SiC fuel rods, aimed to offer drastic improvements during beyond design basis accidents.In order to contain radioactive material in fuel rods even in beyond-design basis accidents, where the core temperature can rise in excess of 1200°C, Framatome is developing a Silicon Carbide (SiC) based cladding solution, PROtect SiC. This advanced cladding offers high strength and reduced oxidation, along with other favorable attributes in both normal and accident conditions, thus proving to be a potential replacement to traditional zirconium based cladding. In the past year, Framatome and CEA jointly made systematic improvements to their 2014 sandwich design to address critical technical challenges such as hydrothermal corrosion, hermeticity and sealing the ends of the cladding tube.Further efforts are focused on continuing testing and evaluation of candidate solutions in test reactors. This paper outlines the irradiation testing being planned on SiC for accelerated development and implementation in the current fleet of commercial light water reactors. The plan includes loop testing of SiC fuel rods in Advanced Test Reactor ATF-2, followed by transient testing of both fresh and irradiated fuel at Transient Reactor Test facility (TREAT). Additionally, this paper presents qualitative analysis of safety benefits associated with SiC cladding in both design basis and beyond design basis accidents.

International audience; To improve the tolerance of the fuel towards accidental scenarios of light water reactors (LWR), an extensive R&D is performed by the French Alternative Energies and Atomic Energy Commission (CEA) in the framework of the French Nuclear Institute with FRAMATOME and EDF. In particular, fuel cladding constituted of silicon carbide continuous fiber ceramic matrix composite materials (SiCf/SiC) is considered as a long-term option besides evolutionary designs (e.g. Cr-coated cladding).The outstanding mechanical behaviour of SiCf/SiC cladding tubes at high temperature and their low oxidation rate, resulting in reduction in hydrogen generation, has been previously reported both in LWR operating conditions and in high temperature steam conditions. However, after exposure in autoclaves, some recession of the SiC was detected. Even if it remains limited, this raises the question of the cladding dissolution, the resulting pollution of the primary loop by silicium-base species and the effect of the water flow on these phenomena.The present work reports the preliminary result obtained in a corrosion loop especially developed at CEA to address the flow accelerated corrosion of SiCf/SiC cladding tubes. The device FACETT (Flow Accelerated Corrosion Experiments for Tube Testing) allows to investigate four 50-mm-long tubes in primary loop conditions (350°C, appropriate chemistry) and with water flow of 4 m.s-1. Sampling of the media can be performed to follow the chemistry change. The results will include weight loss measurements, surface examination and chemical analyses after a 5-month-exposure of CEA-made SiCf/SiC cladding tubes. A comparison with the results obtained in static media will be done.

International audience; Le concept de tube hexagonal mixte avec tronçons en composite SiC/SiC et armure métallique, destiné à contenir le faisceau d’aiguilles de combustible, constitue une option séduisante sur le plan fonctionnel pour un réacteur nucléaire de quatrième génération à spectre rapide et à caloporteur gaz et/ou sodium. Il autorise dans son principe une utilisation à haute température, acquise grâce à un comportement mécanique acceptable jusqu’au-delà des plages habituelles de fonctionnement, tout en satisfaisant les contraintes de transparence neutronique et de résistance à l’irradiation. En s’appuyant sur l’expertise acquise dans le cadre du développement des gaines de combustible réfractaires, des tronçons hexagonaux prototypes en composites ont pu être récemment élaborés. Une solution privilégiée pour le renfort fibreux a consisté en la mise au point d’une architecture hybride associant la technologie d’enroulement filamentaire pour la partie interne à un tressage interlock en seconde paroi. Compte tenu des dimensions conséquentes de l’objet comparativement à celles des gaines, une voie nouvelle fondée sur l’utilisation d’un fil plat multibrins, préalablement préparé à partir à partir de fils unitaires a été expérimentée. En première approche, ces essais on été conduits à partir de fibres de carbone. La matrice de carbure de silicium a été obtenue par le procédé d’infiltration en phase gazeuse. Ce travail restitue l’état des travaux engagés pour concevoir ces nouvelles structures composites. Les microstructures élaborées sont ici décrites en termes de recouvrement, de régularité et de porosité aux différentes échelles d’observation du composite. Une analyse fine par tomographie X permet d’accéder aux chemins de tressage. Enfin, le comportement mécanique est évalué à température ambiante à partir d’essais de flexion 3 points et de traction uni-axiale réalisés avec cycles de charge et décharge sur des éprouvettes prélevées dans l’axe (0°). Au-delà des amplitudes modestes des contraintes atteintes qui dépendent de la nature du renfort, les résultats obtenus démontrent la pertinence de la voie proposée pour concevoir un matériau « défragilisé ».