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

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

Stochastic Calculator Of Neutron transport Equation

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

List of open source projects related to OpenMC

Create parametric 3D fusion reactor CAD models

Openmc-FEnicsx for muLtiphysics tutorIAl

THOR is a radiation transport code for unstructured meshes.

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

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

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

Energy-dependent neutron transport Monte Carlo implemented in Rust.

The package for reading mcnp input in a pythonic way

An open-source MOOSE-based radiation transport and fluid activation solver.

MCNP SDEF to OpenMC conversion tool