MC-kit

parametrized-tokamak-source

Create OpenMC source definition from tabulated plasma parameters

mckit

Tools to work with MCNP models and results

map-stp

Transfer meta information on cells from an STP file to MCNP model. Set materials and densities according to tags specified in STP model tree.

xpypact

Python workflow framework for FISPACT

iWW-GVR

A tool to manipulate MCNP weight window (WW) and to generate Global Variance Reduction (GVR) parameters

MontePy

Make objects not regexes. A python library to read, edit, and write MCNP input files.

networkx

Network Analysis in Python

openmc

OpenMC Monte Carlo Code

openmc_mcnp_adapter

Tool for converting MCNP input files to OpenMC classes/XML

parametric-plasma-source

Source and build files for parametric plasma source for use in fusion neutron transport calculations.

PyTables

A Python package to manage extremely large amounts of data

SpaceClaim_API_NeutronicsTools

Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis

SpaceClaim_scripts

Assitant script for accelerating the modeling

aioduckdb

asyncio bridge to the duckdb library

csg2csg

Tools to translate between different CSG geometry types

DAGMC

Direct Accelerated Geometry Monte Carlo Toolkit

ERSN-OpenMC

ERSN-OpenMC is a Graphical User Interface for OpenMC Monte Carlo particle transport simulation code, originally developed by Jaafar EL Bakkali & Tarek EL Bardouni, members of Radiation and Nuclear Systems Group ERSN at University Abdelmalek Essaady in Tetouan (Morocco).

GEOUNED

Current version of the code

JADE

JADE, a novel nuclear data libraries V&V tool

mc-tools

Some Monte Carlo tools

mckit-meshes

A Python package to work with MCNP meshtallies and weight meshes

mckit-nuclides

Python code to work with elements and their isotopes

mcnptools

neutronics_material_maker

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

ONIX

ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers coupling with the open-source transport code OpenMC.

openmc-plasma-source

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

openmc-plotter

Native plotting GUI for model design and verification