Burning fuel for cheap! Transport-independent depletion in OpenMC
OpenMC is an open source, community-developed, Monte Carlo tool for neutron transport simulations, featuring a depletion module for fuel burnup calculations in nuclear reactors and a Python API. Depletion calculations can be expensive as they require solving the neutron transport and bateman equations in each timestep to update the neutron flux and material composition, respectively. Material properties such as temperature and density govern material cross sections, which in turn govern reaction rates. The reaction rates can effect the neutron population. In a scenario where there is no significant change in the material properties or composition, the transport simulation may only need to be run once; the same cross sections are used for the entire depletion calculation. We recently extended the depletion module in OpenMC to enable transport-independent depletion using multigroup cross sections and fluxes. This talk will focus on the technical details of this feature, its validation, and briefly touch on areas where the feature has been used. Two recent use cases will be highlighted. The first use case calculates shutdown dose rates for fusion power applications, and the second performs depletion for fission reactor fuel cycle modeling.
