A Heterogeneous Model for Burnup Calculation in High Temperature Gas-Cooled Reactors

摘要

A high resolution MCNPX model is developed to simulate nuclear design characteristics and fuel cycle features of High Temperature Gas-Cooled Reactors. Contrary to the conventional approach in the MCNPX model, fuel regions containing TRISO particles are not homogenized. A cube corner distribution approximation is used to directly model randomly dispersed TRISO fuel particles in a graphite matrix. The universe filling technique is used cover the entire range of fuel particles in the core. The heterogeneous MCNPX model is applied to simulate and analyze the complete fuel cycle of the General Atomics Plutonium-Consumption Modular Helium Reactor (PC-MHR). The PC-MHR reactor design is a variation of the General Atomic MHR design and is designed for the consumption or burning of excess Russian weapons plutonium. The MCNPX burnup calculation of the PC-MHR includes the simulation of a 260 effective full-power day fuel cycle at 600 MWt. Results of the MCNPX calculations suggest that during 260 effective full-power day cycle, 40% reduction in the whole core Pu-239 inventory could be achieved.Results of heterogeneous MCNPX burnup calculations in PC-MHR are compared with results of deterministically calculated values obtained from DIF3D codes. For the 260 effective full-power day cycle, the difference in mass Pu-239 mass reduction calculation using heterogeneous MCNPX and homogeneous DIF3D models is 6%. The difference in MCNPX and DIF3D calculated results for higher actinides are mostly higher than 6%.