Ceramic Materials for Nuclear Energy Research and Applications: Poster Session
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee, TMS: Energy Committee
Program Organizers: Walter Luscher, Pacific Northwest National Laboratory; Xian-Ming Bai, Virginia Polytechnic Institute and State University; Lingfeng He, North Carolina State University; Sudipta Biswas, Idaho National Laboratory; Simon Middleburgh, Bangor University
Tuesday 5:30 PM
March 21, 2023
Room: Exhibit Hall G
Location: SDCC
O-31: Radiation Damage in Lithium Oxide, a Surrogate for Beryllium Carbide: David Magee1; Diego Muzquiz2; Stephen Raiman3; David Holcomb4; 1Lancaster University; 2University of Michigan; 3Texas A&M University; 4Oak Ridge National Laboratory
Be2C is a proposed moderator material for molten salt reactors due to its neutron moderating properties, high melting point, and corrosion resistance in FLiBe salt. Radiation damage in this material at high temperature is unknown. The only known study of irradiation damage in Be2C was at 90°C to a fluence of less than 0.01 dpa. The toxicity of beryllium makes studying its properties especially difficult. Li2O is a useful surrogate because it shares a common antifluorite crystal structure with Be2C.For this work, radiation damage in Li2O was investigated. Solid Li2O was fabricated from powder with spark plasma sintering, and irradiated with oxygen ions at 700ᴼC. Samples were examined with TEM and data is presented showing how radiation affected the microstructure of the materials. This work presents new data on the behaviour of an antifluorite structured ceramic under irradiation, for use in understanding the potential behaviour of Be2C.
O-1: Uranium Silicide Processing for Advanced Reactors: Zach Huber1; Elise Conte1; 1Pacific Northwest National Laboratory
This poster will show advances in key elements of uranium silicide (U3Si2) processing science. Pacific Northwest National Laboratory under the United States High Performance Research Reactor Project has been investigating multiple attributes of high quality U3Si2 dispersion fuel. Investigations include oxidation, mechanical behavior, particle size distribution effects on homogeneity, thermal conductivity as well as many other aspects regarding the processing and fabrication of dispersion fuels. This poster will highlight some of the areas of research interest with the U3Si2-Al dispersion fuel.