Beryllium Carbide Moderators

Project Lead: Diego Muzquiz

Graphite is a commonly proposed moderator for advanced reactors, including thermal spectrum molten salt reactors, despite its low moderating cross section and dimensional instability under irradiation. Beryllium carbide (Be2C) is an attractive alternative to graphite moderators because of its high melting point, moderating efficiency, and environmental compatibility. However, its behavior under neutron irradiation is not yet known.

For this work, a new experiment was designed to safely irradiate beryllium-containing samples at The Michigan Ion Beam Laboratory. Using this new capability, bulk Be2C samples are irradiated with carbon ions at varying doses and temperatures up to 700℃. Fallowing this, characterized with SEM, FIB, and TEM is used to measure effects of irradiation on sample microstructures. Material compatibility in molten salt is also explored for the project including FLiBe static and flowing exposure test.