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

Meshing library for nuclear workflows

Stochastic Calculator Of Neutron transport Equation

Create parametric 3D fusion reactor CAD models

MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.

MC/DC: Monte Carlo Dynamic Code

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

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 Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.

THOR is a radiation transport code for unstructured meshes.

Openmc-FEnicsx for muLtiphysics tutorIAl

The package for reading mcnp input in a pythonic way

A collection of neutronics models for comparing neutronics simulations in both CAD and CSG formats.

DIF3D plugin to the ARMI nuclear reactor analysis framework

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