OpenMC Monte Carlo Code
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Updated
Aug 23, 2024 - Python
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
Create parametric 3D fusion reactor CAD models
Stochastic Calculator Of Neutron transport Equation
MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
MC/DC: Monte Carlo Dynamic Code
List of open source projects related to OpenMC
Combines open source packages to produce an automated fusion specific neutronics workflow
Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis
THOR is a radiation transport code for unstructured meshes.
Openmc-FEnicsx for muLtiphysics tutorIAl
A collection of neutronics models for comparing neutronics simulations in both CAD and CSG formats.
DIF3D plugin to the ARMI nuclear reactor analysis framework
A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.
The package for reading mcnp input in a pythonic way
A Python package that extends OpenMC base classes to provide convenience features and standardized tallies when simulating DAGMC geometry with OpenMC.
A minimal example implementation of an open source method of making DAGMC geometry with Paramak and simulating tritium production with OpenMC
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