| Abstract Scope |
Beryllium carbide (Be2C) is an attractive alternative to graphite moderators because of its high melting point, moderating efficiency, and theoretical environmental compatibility as it does not decompose in HTGRs, MSRs, or fusion reactors. However, its behavior under neutron irradiation is not yet known. Research on otherwise promising beryllium compounds is restricted because of the toxic nature of the material. For this work, a novel experiment was designed to safely irradiate beryllium-containing samples at The Michigan Ion Beam Laboratory (MIBL). Using this new capability, Be2C samples were irradiated with 9 MeV C3+ ions from 2 dpa to 30 dpa, at temperatures up to 500 ˚ C. Samples were characterized to investigate changes in chemical composition and dislocation nucleation using TEM. No phase precipitation, dislocation loops, or amorphization was observed up to 30 dpa, indicating good radiation tolerance. Paired with these irradiations, samples were exposed to molten FLiBe salt for 100 hours exhibiting little to no signs of mass loss or salt intrusion.
This work advances our understanding of Be2C as a candidate moderator, and suggests that further research on the material may be warranted to fully determine its suitability for use in high temperature nuclear reactors, such as investigations into its compatibility with molten salts and its susceptibility to helium embrittlement. |