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International audience; SiC/SiC composites have aroused interest for nuclear applications thanks to their high decomposition temperature, strong thermomechanical properties, neutron transparency and low swelling under irradiation. Given these properties, SiC/SiC composites are mainly considered as cladding material either for generation IV systems (such as Gas-cooled Fast Reactors) or current nuclear power plant (Light Water Reactors). Two types of fibers are currently available to manufacture nuclear grade composites Hi-Nicalon S (HNS) and Tyranno SA3 (TSA3). Unfortunately, CVI matrix composites made of each fiber do not exhibit similar mechanical behavior. The HNS-based highlight higher ultimate tensile strains and stresses than the TSA3-based ones. Thus, both fibers have similar properties and identical interphase and matrix, which should have lead to similar behavior. The fiber/matrix coupling must have a strong influence on the final mechanical behavior. Hypotheses were made that those differences came from discrepancies in the fibers surface roughness but its decrease is not the only factor at stake. The understanding of the mechanisms controlling the fiber/matrix were then conducted in order to improve the mechanical properties of SiC/SiC composites made of TSA3 fibers. The fiber/matrix interface was characterized by a combination of mechanical testing and TEM investigations. The differences in the composition of the fiber surface were quantified by XPS and IGC. Results have pointed out that the carbonaceous structure of the fiber extreme surface has to be responsible of those mechanical behaviors.

International audience; The interest of SiC/SiC composites as structural materials for nuclear applications has grown over the past decade. Due to their high thermomechanical properties over a wide range of temperature as well as their stability under neutron irradiation, SiC/SiC were initially considered for both the fuel cladding element and the hexagonal sub-assembly body for the fast fission neutron reactors. In 2011, the major accident at Fukushima Daiichi highlighted the need to improve the safety of current facilities and future planned designs of light water reactors (LWRs). The use of a refractory and less reactive material under oxidizing conditions such as a SiC-based fuel cladding could be an alternative solution to the conventional zirconium alloys to improve the accident tolerance.The present work primarily focuses on the thermomechanical performances of SiC/SiC composites under severe environments including high temperature (> 1000DC) in neutral then steam atmospheres. A first objective is to specify the relevant and suitable working range for using the composites under normal operating conditions. The determination of temperature limits before occurring significant degradation, as well as the underlying causes of such a decrease were investigated on representative composite structures. A second objective addresses the thermal shock tolerance of SiC/SiC composites to demonstrate their ability to remain structurally sound after water quenching during the re-flooding phase in situation of accidents for LWRs. Preservation of the geometry which maintain core coolability up to very high temperature are very positive results. Close examinations of the oxidized surfaces evidence the efficiency of the passivation layer in protecting substrate. Finally, a correlation between the observations and the residual mechanical properties was developed to propose a mechanism.

International audience; The present work reports the results from experiments on SiC/SiC clad segments demonstrating their outstanding ability to preserve a coolable geometry during the reflood phase of a postulated design basis Loss Of Coolant Accident (LOCA) and beyond. Experimental assessments of thermal shock performance were investigated for this purpose by quenching specimens from 1400DC and 1500DC under steam environment into room temperature water. SiC/SiC composites remain structurally sound and retain their mechanical properties. Post-quenched SiC/SiC tubes were subjected to internal helium gas pressure to test the permeation of the structures. These tests established the link between the leak-tightness measurements and damages in the microstructure through micro crack initiation. Close examinations of the oxidized surfaces evidence the efficiency of the passivation layer in protecting substrates. Finally, a correlation of the observations with the oxide growth rate is developed to support a better understanding of the mechanisms as well as the prediction of the acceptable limits of application. These positive results reinforce the interest of a SiC-based fuel cladding concept to enhance the accident tolerance of fuel for future reactors.

International audience; Silicon carbide based continuous fiber ceramic matrix composite materials (SiCf/SiC) are considered by the French Nuclear Institute as a long-term option for Gen III/III+ light water reactor cladding to improve the accident tolerance of the fuel (ATF). Consistent with this ambition, the extensive RetD activities over the post-Fukushima period has resulted in significant progress in the fabrication of representative and functional specimen, removing some of technological barriers that prevent such advanced ceramics materials from using in a nuclear environment. In addition to on-going basic research, a collaborative program was launched to assess the thermo-mechanical performances of SiC/SiC composites produced at CEA and to collect the required data for a preliminary conceptual design.The present work reports the results from experiments on SiC/SiC clad segments demonstrating their outstanding ability to preserve a coolable geometry during the reflood phase of a postulated design basis Loss of Coolant Accident (LOCA) and beyond. Experimental assessments of thermal shock performance were investigated for this purpose by quenching specimens from 1200DC to 1500DC under steam environment into room temperature water. SiC/SiC composites remain structurally sound and exhibit a slightly affected residual mechanical behavior depending on the oxidation time and the quenching temperature. Post-quenched SiC/SiC tubes were subjected to internal helium gas pressure to test the permeation of the structures. These tests clearly established the link between the leak-tightness retention and damages in the microstructure through micro crack initiation. Close examinations of the oxidized surfaces evidence the efficiency of the passivation layer in protecting substrates. Finally, a correlation of the observations with the oxide growth rate is developed to support a better understanding of the mechanisms as well as the prediction of the acceptable limits of application. These positive results reinforce the interest of a SiC-based fuel cladding concept to enhance the accident tolerance of fuel for future reactors.

National audience

International audience; The mechanical behavior of 2D-braided SiC/SiC composite tubes depends strongly on the braiding angle. Tubes of three different braiding angles (30°, 45°, 60°) have been observed during in situ tensile tests using synchrotron X-ray tomography. In order to detect the slightly open cracks from heterogeneous microstructures, various image processing techniques have been used and developed. Digital Volume Correlation (DVC) allows the 3D displacement field within the material under loading to be estimated. The crack networks are extracted by a DVC-based image subtraction. The detected cracks are first classified into two groups according to their orientations, and then are characterized in qualitative and quantitative ways. The crack analyses on the three different tubes provide not only a study on the braiding angle effects but also a better understanding on the damage mechanisms of the composite.

International audience; The behavior of fiber/matrix bonding leading to poor/fair damage tolerance of SiC/SiC composites was investigated through the observation of the local damage mechanisms in this interface region.SiCf/SiCm composites reinforced with last generation small diameter SiC fibers are promising candidates for thermomechanical applications in nuclear environments. The ability of these SiCf/SiCm composites to sustain damage is dictated by the fiber/matrix coupling mode, achieved through a pyrolytic carbon (PyC) interphase. In this study, opposite coupling behaviors were compared. Push-out tests were conducted on SiC/SiC composites. Indented fibers were extracted by FIB (Focused Ion Beam) and the microstructural organization of the carbon interface around a dislodge fiber were analyzed by HRTEM (High Resolution Transmission Electronic Microscopy).The experimental results have demonstrated the key role of the characteristics of SiC fibers surface and of its structure of carbon rich multi-layers on the macroscopic mechanical behavior of the SiC/SiC composite. It has also been proved that crack deflection occurred near the fiber/PyC region rather than near the matrix/PyC one. This technique is promising for the understanding and the improvement of the interfacial properties of the ceramic composite materials.

International audience; Increasing the sustainability and the safety ofnuclear reactors require the development of new types ofreactors (GEN-IV systems) which can work as breeder(producing more fuel that it consumes) and can offer thepossibility of burning minor actinides to reduce the waste,such as the Sodium- (SFR) or the Gas-cooled (GFR) FastReactors. Therefore, there is a need to assess materialswhich can withstand these very harsh core conditions. Inthis aim, SiC/SiC composites are promising candidatesthanks to their high decomposition temperature(> 2000°C), low swelling and creep under irradiation andgood neutron transparency. A recent CEA patent hashighlighted that SiC/SiC-based hexagonal tubes wouldincrease the resistance to melting and, as a consequence,the safety of the SFR core. In this way, techniques havebeen developed to manufacture a SiC/SiC hexagonal tubewith given dimensions, which has a relatively low level ofporosities and a pseudo-ductile mechanical behavior(tolerance to deformation). Besides, the chemicalcompatibility between SiC and SiC/SiC compositestowards liquid sodium and its impurities (in particularoxygen) was investigated. For this purpose, two sets ofexperiments were conducted in the CORRON$^2$facility(CEA). On the one hand, immersions up to 2000h in anoxygen-purified ([O]

International audience; Tubular nuclear-grade SiC/SiC composites produced by CEA were investigatedin the present work as cladding material to enhance the accident tolerance ofLight Water Reactors (LWRs). A first part reports a preliminary assessment of thebehavior of SiC/SiC composite tubes in simulated relevant LWR conditionsrelated with the manufacturing parameters. The dependence of the corrosion rateon the chemical composition of water is also investigated for a betterunderstanding of the hydrothermal SiC corrosion mechanisms. Additionally, theability of SiC/SiC composite tubes to preserve their geometry and maintaincoolability under high temperature steam environment is examined. Oxidationtests are performed to confirm the favorable behavior and to evaluate theoxidation kinetics of the composite exposed to high-temperature steam (up to1400°C) under pressure and under dynamic gas flow conditions.

International audience; The mechanical behavior of architectured SiC/SiC composites is driven by different damage mechanisms whose understanding is required for building micromechanics-based models able to reproduce and predict its complexity. The kinematics of the surface, precisely analyzed using DIC at the textile pattern scale, exhibit a fiber realignment unexplained by the cracks observed at the surface. The missing mechanism, tracked by tomography in-situ testing (SOLEIL synchrotron), appears to be in-plane microcracking which does not emerge at the free surface of the composite.

International audience