OpenMC Monte Carlo Code
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Updated
Jul 9, 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
Stochastic Calculator Of Neutron transport Equation
MC/DC: Monte Carlo Dynamic Code
Create parametric 3D fusion reactor CAD models
List of open source projects related to OpenMC
Create DAGMC geometry from CAD
Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis
MontePy is a Python library (API) to read, edit, and write MCNP input files.
THOR is a radiation transport code for unstructured meshes.
The package for reading mcnp input in a pythonic way
DIF3D plugin to the ARMI nuclear reactor analysis framework
Combines open source packages to produce an automated fusion specific neutronics workflow
A collection of neutronics models for comparing neutronics simulations in both CAD and CSG formats.
A pretty viewer for XSM files generated by DRAGON/DONJON or APOLLO neutronic codes
Converts mesh vertices and connectivity to h5m geometry files compatible with DAGMC simulations
An open source utility to convert various publicly available macroscopic nuclear cross section formats
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