An irradiation program to evaluate BeO as a moderator or fuel matrix material in nuclear power reactors has been completed at the Oak Ridge National Laboratory. The primary objectives were: to establish the limits of stability of sintered BeO of various grain sizes and densities toward fast-neutrons and to determine mechanisms of damage. Specimens were irradiated at 110, 650, and 1100 °C; fast neutron doses up to 4.7 × 1021 n/cm2 (> 1 MeV) were achieved. Both long-term and short-term irradiations were carried out in the same flux profile to determine the effect, if any, of flux intensity or dose rate on neutron damage. Specimen fracturing, volume expansion, and lattice parameter expansion all increased with increasing neutron dose and decreased with increasing temperature. However, even at the maximum temperature, some specimens disintegrated completely to powder when irradiated to doses greater than 2 × 1021 n/cm2 (> 1 MeV). The principal mode of damage under all conditions was grain boundary separation, which was caused by anisotropic crystal expansion resulting from fast-neutron produced defects. No flux intensity effect on gross damage was detected in any of the experiments. Volume expansion, a reliable criterion of neutron damage to BeO, was found to be greater in long-term than in short-term irradiations at 1100 °C at equivalent neutron doses. In-pile annealing of neutron damage at this temperature may be expected to produce a flux intensity effect on damage, but it is masked by another effect that is a function of the number of thermal cycles incurred during reactor operation. Thermal cycling appears to promote the separation of grain boundaries stressed by anisotropic crystal expansion, thereby increasing the volume expansion. To minimize neutron damage in reactor application, BeO of small grain size and low density should be used at as high a temperature and with as few thermal cycles as practicable.