Solving the neutron transport equation on unstructured mesh and its application in high-fdelity multi-physics analysis of microreactors

It contains output files from two cases: the Godiva prompt critical transient and the 1D multi‑physics analytical benchmark. The manuscript presents a high‑fidelity neutron transport solver based on the self‑adjoint angular flux (SAAF) formulation, implemented within the MOOSE framework.

The solver uses unstructured meshes to accurately model complex geometries and explicit mesh deformation, which is essential for micro‑reactor analysis where neutron leakage, thermal‑hydraulics, and thermal‑expansion feedback are strongly coupled. The dataset supports the validation of the solver’s capabilities in handling nonlinearities arising from source‑term coupling, material properties, and geometric deformation.