On the fly prediction of th-dependent spatial macroscopic cross-sections using FFT

Monte Carlo (MC) codes can accurately model neutron transport in nuclear reactors. However, the efficient inclusion of thermal-hydraulic (TH) feedback within the MC calculation sequence is still an open problem, particularly when burnup's time-evolution must be included in the analysis. For this reason, deterministic codes, leveraging the use of macroscopic cross-sections generated with higher order methods from 2D lattice calculations, are still widely used to perform reduced-order multiphysics analyses. However, traditional cross-sections generation procedures typically decompose the large core problem into multiple assembly-level problems; thus not having the ability to capture inter-nodal effects. Moreover, the pre-generation procedure requires additional pre-computational time to perturb/branch the problem for various operational conditions (e.g. fuel temperature), which, again, is decoupled from the core. In this paper, we propose a new method leveraging the use of Fourier transfer functions to predict the cross-sections distribution due to a variation in TH conditions. The method was tested against a 3D BWR unit-cell problem with realistic density profile and axial fuel heterogeneity. The method was able to compute the mono-energetic cross-sections distribution with maximum error lower than 2%. Insights on the influence of the statistics used to generate the cross-sections on the accuracy of the results is also provided.