OpenMC Monte Carlo Code

Create parametric 3D fusion reactor CAD and neutronics models

Meshing library for nuclear workflows

Native plotting GUI for model design and verification

Tool for converting MCNP input files to OpenMC classes/XML

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

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

Workflow and Template Toolkit for Simulation (WATTS)

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 Genetic Algorithm (GA) / Discrete Particle Swarm Optimization/ Hybrid (GA-PSO) for nuclear fuel optimization using ML surrogates (DNN, KNN, Random Forest, Ridge) and OpenMC. Optimizes fuel loading patterns for a target k-eff and minimal Power Peaking Factor (PPF).

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

Project repository for "An Open Source Nuclear Modeling Ecosystem to Support Fusion Pilot Plant Design" collaboration between MIT and ANL

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

MCNP SDEF to OpenMC conversion tool

A Python package that finds and converts OpenMC tally units.

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

Create an arbitrary parametric tokamak neutron source