neutronics

OpenMC Monte Carlo Code

Meshing library for nuclear workflows

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

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

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.

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

The package for reading mcnp input in a pythonic way

MCNP SDEF to OpenMC conversion tool

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

DIF3D plugin to the ARMI nuclear reactor analysis framework

Python reference implementation for finite-volume neutron-diffusion calculations on regular hex-z lattices

A pretty viewer for XSM_FILE/SEQ_ASCII output files generated by DRAGON/DONJON/APOLLO neutronic codes

a tool for creating axially symetric CSG geometry

A Python package for creating publication quality plots for neutron / photon / particle spectrum

Materials that are dependent on conditions

Convert non overlapping STL files into a DAGMC h5m file complete with material tags and ready for use in neutronics simulations.

Converts mesh vertices and connectivity to h5m geometry files compatible with DAGMC simulations