This paper presents preliminary neutronics analyses for nuclear thermal propulsion (NTP) reactors using nodal diffusion codes. NTP core neutronics analyses typically rely on Monte Carlo (MC) codes, which are computationally expensive and can be prohibitive to use in transient simulations. The primary objective of this study is to demonstrate a framework for NTP analysis that is better aligned with conventional methods widely used in modeling of light water reactors.

This study is only the first step to using an adapted homogenization procedure. Here, the few-group macroscopic cross sections are generated using a single MC simulation and fed into a diffusion code to obtain the multiplication factor and spatial power distribution.

This paper investigates various homogenization techniques through several one-dimensional and two-dimensional sensitivity analyses. To ensure that the MC solution is reproduced by the diffusion solver, reference surface discontinuity factors (DFs) are generated using a Jacobian-Free Newton Krylov (JFNK) iterative scheme. These DFs are then applied as equivalence correction parameters in a NERVA-styled NTP core. The solution obtained by the nodal diffusion code DYN3D yields near perfect agreement with the reference power profile only when complemented by the JFNK-generated DFs.