MCNP topic
The Monte Carlo N-Particle (MCNP) radiation transport code is a Monte Carlo transport code developed by Las Alamos National Laboratory (LANL). It supports over 37 different types of particles, and is widely used by nuclear engineers, and nuclear physicists.
openmc_mcnp_adapter
Tool for converting MCNP input files to OpenMC classes/XML
McCAD-Library
a CAD to MC geometry conversion tool
neutronics_material_maker
A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak
watts
Workflow and Template Toolkit for Simulation (WATTS)
MontePy
MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
neutronics_material_maker
A tool for making reproducible materials and standardizing use across several neutronics codes