OpenMC Monte Carlo Code
-
Updated
Jun 13, 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
Create parametric 3D fusion reactor CAD models
Create DAGMC geometry from CAD
Stochastic Calculator Of Neutron transport Equation
MontePy is a Python library (API) to read, edit, and write MCNP input files.
List of open source projects related to OpenMC
MC/DC: Monte Carlo Dynamic Code
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
A collection of neutronics models for comparing neutronics simulations in both CAD and CSG formats.
THOR is a radiation transport code for unstructured meshes.
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.
A Python package that extends OpenMC base classes to provide convenience features and standardized tallies when simulating DAGMC geometry with OpenMC.
The package for reading mcnp input in a pythonic way
A minimal example implementation of an open source method of making DAGMC geometry with Paramak and simulating tritium production with OpenMC
Add a description, image, and links to the neutronics topic page so that developers can more easily learn about it.
To associate your repository with the neutronics topic, visit your repo's landing page and select "manage topics."