neutronics
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A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutronics tools
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Mar 18, 2025 - Jupyter Notebook
Meshing library for nuclear workflows
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Nov 22, 2024 - Python
Stochastic Calculator Of Neutron transport Equation
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Jan 25, 2025 - Fortran
MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
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Mar 18, 2025 - Python
Create parametric 3D fusion reactor CAD models
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Sep 23, 2021
MC/DC: Monte Carlo Dynamic Code
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Mar 18, 2025 - Python
List of open source projects related to OpenMC
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Aug 28, 2024
Openmc-FEnicsx for muLtiphysics tutorIAl
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Jul 29, 2024 - Jupyter Notebook
Combines open source packages to produce an automated fusion specific neutronics workflow
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May 31, 2022 - Python
Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis
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Mar 19, 2021 - C#
THOR is a radiation transport code for unstructured meshes.
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Sep 17, 2024 - Fortran
A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.
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May 9, 2024 - Python
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Feb 6, 2024 - C++
The package for reading mcnp input in a pythonic way
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Jul 25, 2022 - Python
A collection of neutronics models for comparing neutronics simulations in both CAD and CSG formats.
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Apr 14, 2024 - Python
DIF3D plugin to the ARMI nuclear reactor analysis framework
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Jun 16, 2022 - Python
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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Dec 9, 2022 - Python
A Python package that extends OpenMC base classes to provide convenience features and standardized tallies when simulating DAGMC geometry with OpenMC.
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Aug 15, 2023 - Python
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