OpenMC Monte Carlo Code

Create parametric 3D fusion reactor CAD and neutronics models

Native plotting GUI for model design and verification

Meshing library for nuclear workflows

Tool for converting MCNP input files to OpenMC classes/XML

Creates a plasma source as an openmc.source object from input parameters that describe the plasma

Workflow and Template Toolkit for Simulation (WATTS)

Convert CAD geometry (STP files) or Cadquery assemblies to DAGMC h5m files

A tool for making reproducible materials and standardizing use across several neutronics codes

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

Combines open source packages to produce an automated fusion specific neutronics workflow

A Python package for downloading h5 cross section files for use in OpenMC.

A Python package for extracting and plotting the locations, directions, energy distributions of OpenMC source particles

A Python package that extends OpenMC base classes to provide convenience features and standardized tallies when simulating DAGMC geometry with OpenMC.

A minimal example implementation of an open source method of making DAGMC geometry with Paramak and simulating tritium production with OpenMC

A Python package that finds and converts OpenMC tally units.

Convert non overlapping STL files into a DAGMC h5m file complete with material tags and ready for use in neutronics simulations.

Segments cells into smaller cells. Useful for redefining geometry for cell based shutdown dose rate simulations.

Run OpenMC on Android phone

Plots slices of OpenMC native GSC surface geometry with hovertext surface identification