Application of a Custom Depletion Framework to the Prediction of Neutron Flux Distribution Through Depletion

Previous works by the authors have introduced the spatial flux variation method (SFV) for predicting the changes in neutron flux due to a change in material compositions. In order to remove a full transport solution at the end-of-step, this work presents a framework responsible for computing macroscopic cross sections after a depletion event. These end of-step cross sections are estimators of changes in neutron loss and production, and enable the prediction of neutron flux using only information obtained from a single beginning of-step transport solution. The framework reads in all relevant data needed to model the depletion system, including one-group cross sections and effective fission yields to reproduce the problem using an external solver. The framework also supports extrapolating microscopic cross sections in order to rebuild the end-of-step macroscopic cross sections needed for the flux prediction. Results indicate that the SFV method is not adversely effected by the external depletion solution, and can be implemented alongside an existing transport-depletion framework.