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Helium mobility in advanced nuclear ceramics

Abstract : While the current second and third generation nuclear plant designs provides an economically, technically, and publicly acceptable electricity supply in many markets, further advances in nuclear energy system design can broaden the opportunities for the use of nuclear energy. The fourth generation of nuclear reactors is under development. These new reactors are designed with the following objective in mind: sustainability, safety and reliability, economics, proliferation resistance. Out of six Generation IV systems namely, Gas-Cooled Fast Reactor (GFR), Lead-Cooled fast reactor (LFR), Molten Salt Reactor (MSR), Sodium-Cooled Fast Reactor (SFR), Supercritical-Water-Cooled Reactor (SCWR), Very-High-Temperature Reactor (VHTR), this work is dedicated to identify specific fuel type that is compatible with gas-Cooled fast reactor (GFR) in-Core service conditions and could be extended to diagnose potential cladding material for SFR. The French strategy is mainly oriented towards the development of sodium-Cooled fast reactors (SFR) and very slightly focused on GFR. This dissertation is focused on the study of transition metal ceramics which are candidates for fuel coatings in GFR and have been considered as potential cladding materials for SFR. The specific fuel type in GFR should consists of spherical fuel particle made up of UC or UN, surrounded by a ceramic coating which provides structural integrity and containment of fission products. The most promising candidates for ceramic coatings are ZrN, ZrC, TiN, TiC & SiC due to a combination of neutronic performance, thermal properties, chemical behavior, crystal structure, and physical properties. It is obvious that these ceramics would be exposed to energetic fission products from fuel such as heavy ions and neutrons. These high-Energy neutron will knock the atoms in the surrounding materials and can induce (n, α) reactions, thus producing high concentration of helium atoms during and after reactor operation. The helium atoms produced are energetic and can easily penetrate into the surrounding material. Helium atoms are considered to be highly insoluble in previously studied structural nuclear materials. The accumulation of helium into solid matrix, can lead to the formation of bubbles, cavity, swelling, embrittlement etc. Helium can strongly induce grain boundary cavitation that can produce formation of inter-Granular channels, which may serve as pathways for release of radioactive elements to the environment or lead to grain-Boundary weakening and de-Cohesion. Particularly in ceramics, large quantities of helium can also lead to dimensional changes and cracks due to over-Pressurized helium bubbles. Therefore, study of helium behavior in advanced nuclear ceramics under high operating temperatures and extreme radiation conditions predicted for GFRs is viewed as crucial. In this thesis, ion-Implantation technique and material characterization techniques are used to study diffusion of helium in transition metal ceramics under thermal and extreme irradiation environments. Our main aim during this thesis is: 1) To calculate diffusion and migration energies of helium under different experimental conditions by applying theoretical models on experimental data.2) To investigate the role of microstructure such as grain boundaries, native vacancies and porosity on helium accumulation and its evolution after helium accumulation.3) To know the role of helium introduction conditions on helium diffusion. 4) To establish and validate an approach to calculate pressure built by helium gas inside the bubbles and to verify if the pressure approaches mechanical stability limit.
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Shradha Agarwal. Helium mobility in advanced nuclear ceramics. Materials Science [cond-mat.mtrl-sci]. Université Paris Sud - Paris XI, 2014. English. ⟨NNT : 2014PA112197⟩. ⟨tel-01138454⟩

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