A high-fidelity, free user input cylinder meshing tool for MCNP.
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Updated
Jun 6, 2021 - C
A high-fidelity, free user input cylinder meshing tool for MCNP.
Intermediate project about Point Kintetics and Dynamics Equations simulation of AP1000 reactor
Command line tool to convert MCNP mesh tallies to Visual ToolKit (VTK) formats. Supports all MCNPv6.2 legacy meshtal output formats, for both for rectangular and cylindrical meshes.
Isotropic Monte Carlo simulations examining thermal neutron statistics in water, lead, and graphite, including animated neutron history, delta tracking and spherical geometry adaptations.
An open source slide deck on fusion neutronics
Energy-dependent neutron transport Monte Carlo implemented in Rust.
The same neutronics geometry made using Constructive Solid Geometry (CSG) and DAGMC faceteted surface mesh at different resolutions to compare simulation results
Simple code to simulate Neutron Scattering in Non-Radiative matter using Monte Carlo simulations
A modular toolkit of fast and reliable libraries for neutronics analysis. Several command line tools are built with this core collection of crates.
A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.
Converts mesh vertices and connectivity to h5m geometry files compatible with DAGMC simulations
A minimal example implementation of an open source method of making DAGMC geometry with Paramak and simulating tritium production with OpenMC
A Python package that extends OpenMC base classes to provide convenience features and standardized tallies when simulating DAGMC geometry with OpenMC.
An open source utility to convert various publicly available macroscopic nuclear cross section formats
DIF3D plugin to the ARMI nuclear reactor analysis framework
Openmc-FEnicsx for muLtiphysics tutorIAl
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