There are 1 repository under openmc topic.
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
ENRICO: Exascale Nuclear Reactor Investigative COde
Create parametric 3D fusion reactor CAD and neutronics models
Native plotting GUI for model design and verification
Meshing library for nuclear workflows
Tool for converting MCNP input files to OpenMC classes/XML
Creates a plasma source as an openmc.source object from input parameters that describe the plasma
Convert CAD geometry (STP files) or Cadquery assemblies to DAGMC h5m files
A tool for making reproducible materials and standardizing use across several neutronics codes
List of open source projects related to OpenMC
A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak
Combines open source packages to produce an automated fusion specific neutronics workflow
A Python package for extracting and plotting the locations, directions, energy distributions of OpenMC source particles
A Python package for downloading h5 cross section files for use in OpenMC.
Openmc-FEnicsx for muLtiphysics tutorIAl
Package Manager for Nuclear Engineering Development
A Python package that extends OpenMC base classes to provide convenience features and standardized tallies when simulating DAGMC geometry with OpenMC.
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 finds and converts OpenMC tally units.
Convert non overlapping STL files into a DAGMC h5m file complete with material tags and ready for use in neutronics simulations.
The jupyter notebook contains python code which creates a BWR square assembly with a 3-by-3 fuel-less center.
Segments cells into smaller cells. Useful for redefining geometry for cell based shutdown dose rate simulations.
Code to simulate the flux through cylinders with the MUTR Neutron Imager
This repository contains a beginner-friendly tutorial for OpenMC, a Monte Carlo particle transport simulation code widely used in nuclear engineering.
USNA Engineering Design Capstone. OpenMC simulations for the High Flux Production Reactor (HYPR) design concepts.
Student research repository for independent research of the material attractiveness of prospective fuel to be used in microreactors
Radiation transport simulations with a human phantoms in OpenMC. Meshes are created in Cubit Coreform. The phantom used in this project is obtained from: https://www.icrp.org/publication.asp?id=ICRP%20Publication%20145
OpenMC model of the EVOL reference MSFR. Includes code used to simulate and plot results in DTU student project "Impact of temperature feedback on reactivity parameters in the Molten Salt Fast Reactor" by Morten Nygaard (ongoing per June 2024). See "README.md" for detailed description of code and project.
Nuclear data processing with ACEMAKER, FRENDY and NJOY
Plots slices of OpenMC native GSC surface geometry with hovertext surface identification