SUBJECT: Ph.D. Proposal Presentation
BY: Elton Chen
TIME: Tuesday, January 30, 2018, 9:00 a.m.
PLACE: Boggs Building, 3-47
TITLE: Atomistic Investigation of Irradiation Resistance of SiC Phase Boundaries
COMMITTEE: Dr Chaitanya S. Deo, Chair (NRE)
Dr Remi Dingreville (Sandia)
Dr David L. McDowell (ME)
Dr Olivier Pierron (ME)
Dr Hamid Garmestani (MSE)


Long-term radiation exposure is certain for materials with applications in next-generation nuclear
reactors, and consequentially, the study of aging of materials has been a popular topic of research
for decades, as a crucial part of safer reactor designs. At the atomic scale, radiation damages
are the results of numerous displacement cascade events, initiated by high energy particle collisions. An individual displacement event creates vacancy-interstitial point defect pairs known
as Frenkel Pair/defect (FP) in the material lattice. The diffusion and accumulation of the vacancy
point defects form larger voids; similarly, the interstitial defect buildup forms various structural dependent dislocation loops and, in f.c.c. lattice, stacking-fault tetrahedra. With
sufficient defect concentrations in the bulk, amorphization can also be observed in materials
with weakened lattices. In most materials, crystalline grain and phase boundaries act as potent
defect sinks, trapping defects and fission gases alike. The presence and accumulation of
void defects and fission gases tend to weaken grain/phase boundary bonding, eventually leading to
boundary decohesion or formation of micro-cracks. In turn, the growth rates of existing
micro-cracks are affected by the emission of dislocation defects and microstructure variations on
the boundary, although the net effects are still ambiguous. Obviously grain and phase
boundaries interactions with the various aforementioned processes play key roles in the aging process. Atomistic scale models are the logical tools to study the boundary aging phenomena, having
the capability to directly simulate boundary microstructures, collisional cascade damages and point
defect evolutions.

For this project, variations of Silicon Carbide phase boundaries are chosen as the case study.
The high melting point and low oxidation rate of SiC makes it a favorable choice for fuel cladding
in next-gen reactors and structural material in fusions operations. Both applications result
in harsh radiation exposure during its lifetime. Although considered a brittle ceramic material,
SiC has been observed to exhibit metallic plasticity behaviors in nano-scale deformations. This
allows access to the atomistic study of decohesion crack growth mechanisms