Pinned Repositories
ERSN-OpenMC
ERSN-OpenMC is a Graphical User Interface for OpenMC Monte Carlo particle transport simulation code, originally developed by Jaafar EL Bakkali & Tarek EL Bardouni, members of Radiation and Nuclear Systems Group ERSN at University Abdelmalek Essaady in Tetouan (Morocco).
iWW-GVR
A tool to manipulate MCNP weight window (WW) and to generate Global Variance Reduction (GVR) parameters
map-stp
Transfer meta information on cells from an STP file to MCNP model. Set materials and densities according to tags specified in STP model tree.
mckit
Tools to work with MCNP models and results
neutronics_material_maker
A tool for making reproducible materials and standardizing use across several neutronics codes
openmc
OpenMC Monte Carlo Code
parametrized-tokamak-source
Create OpenMC source definition from tabulated plasma parameters
SpaceClaim_API_NeutronicsTools
Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis
SpaceClaim_scripts
Assitant script for accelerating the modeling
xpypact
Python workflow framework for FISPACT
MC-kit's Repositories
MC-kit/mckit
Tools to work with MCNP models and results
MC-kit/parametrized-tokamak-source
Create OpenMC source definition from tabulated plasma parameters
MC-kit/map-stp
Transfer meta information on cells from an STP file to MCNP model. Set materials and densities according to tags specified in STP model tree.
MC-kit/xpypact
Python workflow framework for FISPACT
MC-kit/ERSN-OpenMC
ERSN-OpenMC is a Graphical User Interface for OpenMC Monte Carlo particle transport simulation code, originally developed by Jaafar EL Bakkali & Tarek EL Bardouni, members of Radiation and Nuclear Systems Group ERSN at University Abdelmalek Essaady in Tetouan (Morocco).
MC-kit/iWW-GVR
A tool to manipulate MCNP weight window (WW) and to generate Global Variance Reduction (GVR) parameters
MC-kit/MontePy
Make objects not regexes. A python library to read, edit, and write MCNP input files.
MC-kit/networkx
Network Analysis in Python
MC-kit/neutronics_material_maker
A tool for making reproducible materials and standardizing use across several neutronics codes
MC-kit/openmc
OpenMC Monte Carlo Code
MC-kit/openmc_mcnp_adapter
Tool for converting MCNP input files to OpenMC classes/XML
MC-kit/parametric-plasma-source
Source and build files for parametric plasma source for use in fusion neutron transport calculations.
MC-kit/PyTables
A Python package to manage extremely large amounts of data
MC-kit/SpaceClaim_API_NeutronicsTools
Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis
MC-kit/SpaceClaim_scripts
Assitant script for accelerating the modeling
MC-kit/aioduckdb
asyncio bridge to the duckdb library
MC-kit/csg2csg
Tools to translate between different CSG geometry types
MC-kit/DAGMC
Direct Accelerated Geometry Monte Carlo Toolkit
MC-kit/GEOUNED
Current version of the code
MC-kit/JADE
JADE, a novel nuclear data libraries V&V tool
MC-kit/mc-tools
Some Monte Carlo tools
MC-kit/mckit-meshes
A Python package to work with MCNP meshtallies and weight meshes
MC-kit/mckit-nuclides
Python code to work with elements and their isotopes
MC-kit/mcnptools
MC-kit/ONIX
ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers coupling with the open-source transport code OpenMC.
MC-kit/openmc-plasma-source
Creates a plasma source as an openmc.source object from input parameters that describe the plasma
MC-kit/openmc-plotter
Native plotting GUI for model design and verification