DEVELOPMENT OF AN EFFICIENT PENTRAN MODEL FOR NEUTRON FLUX SIMULATION OF THE UFTR SHIELD TANK

摘要

The motivation of this work is to support development of a burnup reconstruction device which will provide 3-D characterization of spent nuclear fuel at the University of Florida Training Reactor (UFTR) within the shield tank experimental unit. In order to facilitate the design of such a device, a procedure is required to quickly and accurately determine radiation fields within the shield tank region. Utilizing both a 3-D Monte Carlo (MCNP5) and a 3-D deterministic particle transport code (PENTRAN), multi-group neutron flux distributions are calculated. The accuracy and efficiency of the PENTRAN code based on flux distributions throughout the reactor core, graphite reflector, and shield tank regions are assessed and further compared with MCNP5 results. It is demonstrated that the deterministic PENTRAN code package achieves accurate solutions at significantly reduced computational time compared to the analog Monte Carlo calculations.