Microstuctural Characterization of the Radiation Effects in ZrC, a Potential Material for Next Generation Nuclear Plants

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1043-T02-01

Microstuctural Characterization of the Radiation Effects in ZrC, a Potential Material for Next Generation Nuclear Plants Gianguido Baldinozzi1,2, Dominique Gosset1,2, David Simeone1,2, Mickael Dollé1,2, Lionel Thomé3, and Suzy Surblé1,2 1 SPMS, CNRS, Ecole Centrale Paris, Chatenay-Malabry, 92295, France, Metropolitan 2 DEN/DMN/SRMA/LA2M, CEA, CEN Saclay, Gif-sur-Yvette, 91191, France, Metropolitan 3 CSNSM, CNRS, Université Paris XI, Orsay, 91405, France, Metropolitan

ABSTRACT The development of a new generation of nuclear reactors (Gen-IV), with improved thermodynamic yield and a reduction of waste production, makes necessary to consider materials able to withstand high operating temperatures. Transition metal carbides, like ZrC, are then under consideration. Despite their good thermal and neutron properties, they have unfortunately a brittle mechanical behaviour. This is the reason why it is important to investigate the properties of these systems with sub-micrometric grains and as a function of their composition. Therefore, samples having micrometric and nanometric grain sizes (and different oxygen content) were irradiated by low energy ions at room temperature to simulate their behaviour in a neutron flux. The irradiation effects in these materials were studied by grazing X-ray diffraction and transmission electron microscopy. INTRODUCTION The fabrication of reliable, high performance ceramics is an important industrial challenge and it has long been recognized that further progress will be made by an increased understanding of the role played by the microstructure and the surface chemistry of the starting powders. Rather surprisingly, little experimental work has been directed towards the powder surface characterization [1]. This may be due to experimental difficulties encountered until recently in characterizing the surface of bulk nanostructured materials. Zirconium carbide is one of the candidate materials considered as a component for the fuel elements of some nuclear reactors in the Gen-IV international project [2]. Gen IV materials must be highly refractory, have good thermal conductivity [3], low neutron absorption or scattering cross sections, low damage under irradiation (swelling, thermal conductivity) [4]. Unfortunately, most of the materials presenting these properties also have a brittle mechanical behaviour and a high sensitivity to oxidation. Nanostructuration of these materials may improve their mechanical properties but it can also affect their chemical sensitivity to oxidation. ZrC chemistry is complex and the actual material presents large departures from stoichiometry (vacancies and O substitution), mainly affecting the carbon subnetwork. Little information is known about the behaviour under irradiation of ZrC. Zirconium carbide was irradiated in nuclear reactors to study the macroscopic evolution of the material properties (swelling, thermal and mechanical properties). ZrC is also a component of the Triso fuel particles. Most of those studies show a moderate damage consisting o

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