|About this Abstract
||2018 TMS Annual Meeting & Exhibition
||Accident Tolerant Fuels for Light Water Reactor
||Modeling Radiation Defect Cluster Accumulation in Neutron Irradiated FeCrAl
||Dwaipayan Dasgupta, Brian Wirth
|On-Site Speaker (Planned)
This presentation will focus on modeling neutron induced radiation effects in FeCrAl alloys proposed as accident tolerant fuel clad. Key questions about FeCrAl clad performance relate to the extent of radiation hardening and irradiation creep expected at temperatures around 340°C and for neutron dose levels up to about 15 dpa. We present the results of a cluster dynamics reaction-diffusion based rate theory model to predict the size and density evolution of a/2<111> and a<100> prismatic dislocation loops, as well as the fluxes of defects that partition to network dislocations driving irradiation creep. The modeling predictions are compared to recent experimental results, and indicate a good agreement of the a<100> loop size and a/2<111> loop density. Future model extensions to improve the agreement with the experimental observations will be discussed, as well as the anticipated performance of accident tolerant fuel cladding.
||Planned: Supplemental Proceedings volume