Reactor performance and safety characteristics of two-phase composite moderator concepts for modular high temperature gas cooled reactors

Graphite moderators have an extensive historical performance record, but also feature inherent challenges for modular High Temperature Gas-Cooled Reactors (mHTGRs). Challenges with graphite include non-uniform expansion and contraction under irradiation and build-up of potential energy during the bombardment of high energy neutrons that results in a large energy release under annealing. These challenges have led to the investigation and development of alternative moderators to be utilized in mHTGRs, including beryllium- and hydride-based concepts with compositions selected for favorable moderating power and the potential for improved in-service lifetime as compared to graphite. The proposed moderators are fabricated as two-phase composites with magnesium oxide, MgO, as the radiation-stable host matrix and beryllium metal, Be, beryllium oxide, BeO, or zirconium hydride, ZrH_x=1 (to account for hydrogen loss from the hydride phase during processing), as the entrained moderating phase. Here, we evaluate the reactor performance and safety characteristics of these moderator concepts relative to a graphite reference using a Ft. Saint Vrain-style fuel block. We assessed the cycle length, discharge burnup, natural resource utilization, neutron flux spectra, moderating power, moderating ratio, critical size, moderator and fuel temperature feedback, fuel cycle cost, spent nuclear fuel and high level waste radioactivity per unit energy generated, and environmental impact per unit energy generated. The results demonstrate that the advanced moderators have the potential for comparable or enhanced cycle performance to that of the graphite reference case with significantly improved performance for an optimized moderator-to-fuel ratio design. These advanced moderators are also assessed from a reactor safety standpoint for Design Basis Accidents (DBAs) including Pressurized Loss of Forced Cooling and Depressurized Loss of Forced Cooling accidents for a 350 megawatt thermal prismatic-type mHTGR. The full core thermohydraulic analysis of DBAs show that the high volumetric heat capacity of the beryllium-based moderator grants them a greater margin to fuel failure in these analyses than a conventional graphite moderated system, but the lower thermal conductivity of the beryllium-based moderators leads to longer times at elevated temperatures.