On the Use of Be2C Compacts for Reactor Moderator Application

Beryllium carbide, a ceramic historically considered but not utilized as nuclear reactor moderator, has been fabricated to near-full density. Fabrication has been conducted using direct current sintering, by reaction sintering of beryllium and carbon powders, and by sintering of Be 2 C powder made by reaction of carbon with gas-phase beryllium. Processing optimization, microstructure, and thermal properties of produced ceramics are discussed. Density approaching the theoretical value of 2.44 g/cc was achieved with measured levels of <2% BeO in raw Be 2 C powder and <8% in ceramics. Additionally, two-phase materials of up to 40 volume percent Be 2 C entrained within matrices of either MgO, graphite, or SiC have been produced. Similarly, surrogate TRISO fuel has been entrained within a (low density) matrix of graphite and Be 2 C. All materials underwent neutron irradiation in the temperature range of 130-720°C at a neutron fluence of approximately 4 × 10 24 n/m 2 (E>0.1 MeV). For this limited irradiation campaign swelling of Be 2 C was found to be quite low, especially in comparison with other nuclear ceramics such as SiC. However, an unexpected and gross disintegration of Be 2 C ceramics occurred for both the irradiated and non-irradiated materials, due to ambient exposure to lab-atmosphere. This reaction was not seen in the two-phase materials, attributed to matrix phase protection from moist air.