There are 2 repositories under neutronics topic.
OpenMC Monte Carlo Code
A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutronics tools
Create DAGMC geometry from CAD
Stochastic Calculator Of Neutron transport Equation
Convert CAD geometry (STP files) or Cadquery assemblies to DAGMC h5m files
MC/DC: Monte Carlo Dynamic Code
List of open source projects related to OpenMC
Combines open source packages to produce an automated fusion specific neutronics workflow
Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis
A collection of neutronics models for comparing neutronics simulations in both CAD and CSG formats.
A Python package that extends OpenMC base classes to provide convenience features and standardized tallies when simulating DAGMC geometry with OpenMC.
A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.
DIF3D plugin to the ARMI nuclear reactor analysis framework
The package for reading mcnp input in a pythonic way
A minimal example implementation of an open source method of making DAGMC geometry with Paramak and simulating tritium production with OpenMC
A Python package for creating publication quality plots for neutron / photon / particle spectrum
Convert non overlapping STL files into a DAGMC h5m file complete with material tags and ready for use in neutronics simulations.
Converts mesh vertices and connectivity to h5m geometry files compatible with DAGMC simulations
XSPlot - Neutron cross section plotter for materials
A pretty viewer for XSM files generated by DRAGON/DONJON or APOLLO neutronic codes
A high-fidelity, free user input cylinder meshing tool for MCNP.
Openmc-FEnicsx for muLtiphysics tutorIAl
An open source utility to convert various publicly available macroscopic nuclear cross section formats
Energy-dependent neutron transport Monte Carlo implemented in Rust.
Isotropic Monte Carlo simulations examining thermal neutron statistics in water, lead, and graphite, including animated neutron history, delta tracking and spherical geometry adaptations.
A modular toolkit of fast and reliable libraries for neutronics analysis. Several command line tools are built with this core collection of crates.
Simple code to simulate Neutron Scattering in Non-Radiative matter using Monte Carlo simulations
Command line tool to convert MCNP mesh tallies to Visual ToolKit (VTK) formats. Supports all MCNPv6.2 legacy meshtal output formats, for both for rectangular and cylindrical meshes.
A Python package to identify the part ID number in Brep format CAD files
The same neutronics geometry made using Constructive Solid Geometry (CSG) and DAGMC faceteted surface mesh at different resolutions to compare simulation results