SMR Core Depletion and Spent Fuel Characterization with Nodal Code Ants

To complement the burnup capability of Serpent Monte Carlo transport code in the Kraken reactor simulator framework, a reduced-order approach in nodal code Ants is developed for fuel inventory calculations. The method is based on a microscopic depletion model, which is capable of tracking nuclide concentrations on the node level in 3D coupled core simulations. This enables explicit modeling of local irradiation conditions, which is not practical for Monte Carlo transport in regular engineering applications. This study demonstrates the Ants micro-depletion method in the simulation of three fuel cycles of a boron-free pressurized water reactor. Comparison to a Serpent 3D Monte Carlo simulation shows that Ants can calculate best-estimate nuclide inventories usable for spent fuel transport, storage, and final disposal. Based on the results, the micro-depletion method is an attractive option for spent fuel analysis.