Development of a Micro-Nuclear Reactor Design to Perform Criticality and Radiation Safety Analyses of Its Used Fuel in Storage

Abstract

The transportation and disposition of a High Temperature Gas Cooled (HTGR) micro-reactor’s spent fuel was the focus of this study. This was started by creating a simulation of the microreactor in MCNP6.2, which was used for its versatility in creating models, criticality safety analysis, and radiation shielding analysis. The model was based on the High Temperature Test Reactor from General Atomics made in 2004. From here, the fuel type and prismatic graphite block dimensions along with other aspects were taken and modified to create a new open-sourced model. This model was then used in a MCNP6.2 burnup simulation to get the material composition of the spent fuel. This spent fuel was then used to create source terms for both neutron and gamma production, which were used in the next model of the dry storage cask. Using the dry storage cask, scoping simulations were run to verify the criticality safety and radiation safety of the dry storage cask for transportation and disposition.