TORAX: Tokamak transport simulation in JAX

Tools to work with MCNP models and results

SANDY + OpenMC + UncertaintyQuantification.jl

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

Repository for FES project focused on OpenMC performance improvements

Direct Accelerated Geometry Monte Carlo Toolkit

Vim-fork focused on extensibility and usability

Create parametric 3D fusion reactor CAD and neutronics models

Example Jupyter notebooks for OpenMC

FUsion Synthesis Engine

Convert OpenMC Geometry Components to CAD

Repository for FES Shutdown Dose Rate Workflows project

Native plotting GUI for model design and verification

Package Manager for Nuclear Engineering Development

A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutronics tools

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

A tool for flattening MCNP models

code to target the conversion from a step-file to a h5m-geometry for neutronics

Create DAGMC geometry from CAD

Tools to translate between different CSG geometry types

Parametric 3-D CAD modeling toolset for stellarator fusion devices

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

List of open source projects related to OpenMC

code to target the conversion from a step-file to a h5m-geometry for neutronics

A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes

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

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

Tool for converting MCNP input files to OpenMC classes/XML

OpenMC Monte Carlo Code