Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.

Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis

Tools to work with MCNP models and results

A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak

A tool for flattening MCNP models

Tools for the manipulation and anlysis of Unstructured Meshes phantoms and integration in MCNP6

Monte Carlo Particle Lists

FISPACT-II GitHub pages

ENDF files of the FENDL library

PyNE: The Nuclear Engineering Toolkit

Tool to rename cells, surfaces, materials and universes in MCNP input files.

Tool to rename cells, surfaces, materials and universes in MCNP input files.

Some Monte Carlo tools