Author: Hao Jiang

Publisher:

Published: 2017

Total Pages: 0

ISBN-13:

Because of various excellent properties, SiC has been proposed for many applications in nuclear reactors including cladding layers in fuel rod, fission products container in TRISO fuel, and first wall/blanket in magnetic controlled fusion reactors. Upon exposure to high energy radiation environments, point defects and defect clusters are generated in materials in amounts significantly exceeding their equilibrium concentrations. The accumulation of defects can lead to undesired consequences such as crystalline-to-amorphous transformation1, swelling, and embrittlement, and these phenomena can adversely affect the lifetime of SiC based components in nuclear reactors. It is of great importance to understand the accumulation process of these defects in order to estimate change in properties of this material and to design components with superior ability to withstand radiation damages. Defect clusters are widely in SiC irradiated at the operation temperatures of various reactors. These clusters are believed to cause more than half of the overall swelling of irradiated SiC and can potentially lead to lowered thermal conductivity and mechanical strength. It is critical to understand the formation and growth of these clusters. Diffusion of these clusters is one importance piece to determine the growth rate of clusters; however it is unclear so far due to the challenges in simulating rare events. Using a combination of kinetic Activation Relaxation Technique with empirical potential and ab initio based climbing image nudged elastic band method, I performed an extensive search of the migration paths of the most stable carbon tri-interstitial cluster in SiC. This research reveals paths with the lowest energy barriers to migration, rotation, and dissociation of the most stable cluster. Based on these energy barriers, I concluded defect clusters are thermally immobile at temperatures lower than 1500 K and can dissociate into smaller clusters and single interstitials at temperatures beyond that. Even though clusters cannot diffuse by thermal vibrations, we found they can migrate at room temperature under the influence of electron radiation. This is the first direct observation of radiation-induced diffusion of defect clusters in bulk materials. We show that the underlying mechanism of this athermal diffusion is elastic collision between incoming electrons and cluster atoms. Our findings suggest that defect clusters may be mobile under certain irradiation conditions, changing current understanding of cluster annealing process in irradiated SiC. With the knowledge of cluster diffusion in SiC demonstrated in this thesis, we now become able to predict cluster evolution in SiC with good agreement with experimental measurements. This ability can enable us to estimate changes in many properties of irradiated SiC relevant for its applications in reactors. Internal interfaces such as grain boundaries can behave as sinks to radiation induced defects. The ability of GBs to absorb, transport, and annihilate radiation-induced defects (sink strength) is important to understand radiation response of polycrystalline materials and to better design interfaces for improved resistance to radiation damage. Nowadays, it is established GBs' sink strength is not a static property but rather evolves with many factors, including radiation environments, grain size, and GB microstructure. In this thesis, I investigated the response of small-angle tilt and twist GBs to point defects fluxes in SiC. First of all, I found the pipe diffusion of interstitials in tilt GBs is slower than bulk diffusion. This is because the increased interatomic distance at dislocation cores raises the migration barrier of interstitial dumbbells. Furthermore, I show that both the annihilation of interstitials at jogs and jog nucleation from clusters are diffusion-controlled and can occur under off-stoichiometric interstitial fluxes. Finally, a dislocation line model is developed to predict the role of tilt GBs in annihilating radiation damage. The model predicts the role of tilt GBs in annihilating defects depends on the rate of defects segregation to and diffusion along tilt GBs. Tilt GBs mainly serve as diffusion channel for defects to reach other sinks when defect diffusivity is high at boundaries. When defect diffusivity is low, most of the defects segregated to tilt GBs are annihilated by dislocation climb. Up-to-date, the response of twist GBs under irradiation has been rarely reported in literature and is still unclear. It is important to develop atom scale insight on this question in order to predict twist GBs' sink strength for a better understanding of radiation response of polycrystalline materials. By using a combination of molecular dynamics and grand canonical Monte Carlo, here I demonstrate the defect kinetics in {001} and {111} twist GBs and the microstructural evolution of these GBs under defect fluxes in SiC. I found due to the deep potential well for interstitials at dislocation intersections within the interface, the mobility of defects on dislocation grid is retard and this leads to defect accumulation at GBs for many cases. Furthermore, I conclude both types of twist GBs have to form mixed dislocations with edge component in order to absorb accumulated interstitials at the interface. The formation of mixed dislocation is either by interstitial loop nucleation or by dislocation reactions at the interface. The continuous formation and climb of these mixed dislocations make twist GBs unsaturatable sinks to radiation induced defects.