openmc

There are 41 repositories under openmc topic.

  • parametric-plasma-source

    Language:C++9

  • neutronics-workshop

    A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutronics tools

    Language:Jupyter Notebook127

  • paramak

    Create parametric 3D fusion reactor CAD and neutronics models

    Language:Python71

  • enrico

    ENRICO: Exascale Nuclear Reactor Investigative COde

    Language:C++64

  • plotter

    Native plotting GUI for model design and verification

    Language:Python52

  • stellarmesh

    Meshing library for nuclear workflows

    Language:Python45

  • openmc_mcnp_adapter

    Tool for converting MCNP input files to OpenMC classes/XML

    Language:Python37

  • openmc-plasma-source

    Creates a plasma source as an openmc.source object from input parameters that describe the plasma

    Language:Python31

  • watts

    Workflow and Template Toolkit for Simulation (WATTS)

    Language:Python25

  • cad_to_dagmc

    Convert CAD geometry (STP files) or Cadquery assemblies to DAGMC h5m files

    Language:Python24

  • openmc-ecosystem

    List of open source projects related to OpenMC

    21

  • neutronics_material_maker

    A tool for making reproducible materials and standardizing use across several neutronics codes

    Language:Python21

  • neutronics_material_maker

    A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak

    Language:Python19

  • fusion_neutronics_workflow

    Combines open source packages to produce an automated fusion specific neutronics workflow

    Language:Python14

  • MP-OFELIA

    Openmc-FEnicsx for muLtiphysics tutorIAl

    Language:Jupyter Notebook11

  • openmc_data_downloader

    A Python package for downloading h5 cross section files for use in OpenMC.

    Language:Python11

  • openmc_source_plotter

    A Python package for extracting and plotting the locations, directions, energy distributions of OpenMC source particles

    Language:Python10

  • fes-project-ii

    Project repository for "An Open Source Nuclear Modeling Ecosystem to Support Fusion Pilot Plant Design" collaboration between MIT and ANL

    9

  • nukebox

    Package Manager for Nuclear Engineering Development

    Language:Shell9

  • openmc-dagmc-wrapper

    A Python package that extends OpenMC base classes to provide convenience features and standardized tallies when simulating DAGMC geometry with OpenMC.

    Language:Python7

  • inertial_fusion_openmc_dagmc_paramak_example

    A minimal example implementation of an open source method of making DAGMC geometry with Paramak and simulating tritium production with OpenMC

    Language:Python6

  • openmc_tally_unit_converter

    A Python package that finds and converts OpenMC tally units.

    Language:Python5

  • pincell_msr

    openmc msr depletion capabilities example

    Language:Jupyter Notebook4

  • stl_to_h5m

    Convert non overlapping STL files into a DAGMC h5m file complete with material tags and ready for use in neutronics simulations.

    Language:Python3

  • openmc_sdef_parser

    MCNP SDEF to OpenMC conversion tool

    Language:Python2

  • openmc_first_tutorial

    This repository contains a beginner-friendly tutorial for OpenMC, a Monte Carlo particle transport simulation code widely used in nuclear engineering.

    Language:Jupyter Notebook2

  • openmc_cell_segmenter

    Segments cells into smaller cells. Useful for redefining geometry for cell based shutdown dose rate simulations.

    Language:Python2
  • BWR_SqauredArraySubChannel-OpenMC

    BWR_SqauredArraySubChannel-OpenMC

    The jupyter notebook contains python code which creates a BWR square assembly with a 3-by-3 fuel-less center.

    Language:Jupyter Notebook2

  • Neutron-Imaging-Simulation

    Code to simulate the flux through cylinders with the MUTR Neutron Imager

    Language:Jupyter Notebook1

  • HYPR-Capstone-2023

    USNA Engineering Design Capstone. OpenMC simulations for the High Flux Production Reactor (HYPR) design concepts.

    Language:Jupyter Notebook1

  • ESFR_Model

    European SFR core sub-assembly model

    Language:Jupyter Notebook1

  • Proliferation_Research

    Student research repository for independent research of the material attractiveness of prospective fuel to be used in microreactors

    Language:Jupyter Notebook1

  • OpenMC-termux

    Run OpenMC on Android phone

    Language:Python1

  • neutronic-openmc-stuff

    Neutronics using OpenMC Monte Carlo Code

    Language:Jupyter Notebook

  • msfr

    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.

    Language:Python

  • radiationtransport

    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

    Language:Jupyter Notebook