openmc

There are 1 repository under openmc topic.

• openmc-dev / openmc

  OpenMC Monte Carlo Code

  monte-carlo-simulationparticle-transportnuclear-engineeringnuclear-energynuclear-fusionopenmcneutronicsphoton-transportneutron-transportradiation-transportcomputational-physicshigh-performance-computingnuclear-data
  Language:Python 710

• fusion-energy / neutronics-workshop

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

  neutronicssimulationsopenmcdagmcneutronsphotonscontainerisedtrainingcourselearningfusionenergyfluxnuclearspectrumtritiumtutorial
  Language:Jupyter Notebook 95

• enrico-dev / enrico

  ENRICO: Exascale Nuclear Reactor Investigative COde

  computational-fluid-dynamicsexascale-computinghigh-performance-computingmonte-carlo-simulationmultiphysicsnek5000nekrsnuclear-energyopenmcparticle-transport
  Language:C++ 58

• fusion-energy / paramak

  Create parametric 3D fusion reactor CAD and neutronics models

  fusionreactorgeometryparametriccadpythonmodeldagmcopenmc
  Language:Python 58

• openmc-dev / plotter

  Native plotting GUI for model design and verification

  openmcparticle-transportmonte-carlo-simulationvisualizationgeometry
  Language:Python 43

• Thea-Energy / stellarmesh

  Create DAGMC geometry from CAD

  cadquerydagmcgmshmoabmonte-carlo-simulationneutronicsopenmcbuild123dopencascade
  Language:Python 33

• openmc-dev / openmc_mcnp_adapter

  Tool for converting MCNP input files to OpenMC classes/XML

  mcnpmonte-carlo-simulationnuclear-engineeringopenmc
  Language:Python 30

• fusion-energy / openmc-plasma-source

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

  plasmatokamakfusionopenmcsourceneutron
  Language:Python 23

• watts-dev / watts

  Workflow and Template Toolkit for Simulation (WATTS)

  pythonsimulationtemplatesnuclear-energymcnpworkflow-automationopenmc
  Language:Python 22

• ukaea / neutronics_material_maker

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

  fispactlibrarymaterialmcnpopenmcserpent
  Language:Python 18

• fusion-energy / cad_to_dagmc

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

  cadconverterdagmcmeshneutronicsopenmc
  Language:Python 17

• fusion-energy / neutronics_material_maker

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

  neutronsmateriallibrarydatabaseopenmcmcnpserpentfispact
  Language:Python 17

• openmc-dev / openmc-ecosystem

  List of open source projects related to OpenMC

  monte-carlo-simulationmultiphysics-simulationneutronicsnuclear-engineeringopenmc
  15

• fusion-energy / fusion_neutronics_workflow

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

  neutronicsautomaticsimulationfusionneutronphotoncadopenmcdagmcparamak
  Language:Python 12

• fusion-energy / openmc_source_plotter

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

  openmcsourceparticleplotcoordinatesenergydirectiongamma
  Language:Python 11

• open-radiation-sources / parametric-plasma-source

  neutronicsparticle-physicsneutronphotonopenmcparametric-plasma-source
  Language:C++ 9

• openmc-data-storage / openmc_data_downloader

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

  openmcdatah5crosssectionxml
  Language:Python 9

• fusion-energy / openmc-dagmc-wrapper

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

  neutronicssimulationsopenmcparametricdagmc
  Language:Python 7

• nukehub-dev / nukebox

  Package Manager for Nuclear Engineering Development

  dagmcgeant4nuclear-engineeringnuclear-simulationopenmcpynepackage-manager
  Language:Shell 6

• fusion-energy / openmc_tally_unit_converter

  A Python package that finds and converts OpenMC tally units.

  openmcunitsconvertconversionnormalization
  Language:Python 5

• fusion-energy / 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

  fusionneutronicsopen-sourcetritiumopenmcdagmcparamak
  Language:Python 4

• fusion-energy / stl_to_h5m

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

  dagmcopenmch5mcadneutronicsstl
  Language:Python 3

• openmsr / pincell_msr

  openmc msr depletion capabilities example

  depletionmsropenmcremoval-rate
  Language:Jupyter Notebook 3

• ERMETE-Lab / MP-OFELIA

  Openmc-FEnicsx for muLtiphysics tutorIAl

  dolfinxmulti-physicsneutronicsopenmcpythonthermal-hydraulic
  2
• BWR_SqauredArraySubChannel-OpenMC

  Fuad-HH / BWR_SqauredArraySubChannel-OpenMC

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

  openmc
  Language:Jupyter Notebook 2

• fusion-energy / openmc_cell_segmenter

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

  cellopenmcsegment
  Language:Python 2

• AI-Pranto / ESFR_Model

  European SFR core sub-assembly model

  openmcserpent
  Language:Python 1

• AI-Pranto / OpenMC-termux

  Run OpenMC on Android phone

  openmc
  Language:Python 1

• CJD-1113 / Neutron-Imaging-Simulation

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

  monte-carlo-simulationneutron-imagingopenmcpython
  Language:Jupyter Notebook 1

• dvp2015 / TokamakNeutronSource.jl

  Methods to compute neutron source for a given plasma equilibrium state presented with EQDSK file

  mcnpneutronopenmcplasma
  Language:Julia 1

• eealanoca / openmc_first_tutorial

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

  monte-carlo-simulationnuclear-engineeringopenmcparticle-transport
  Language:Jupyter Notebook 1

• Toshi-23 / HYPR-Capstone-2023

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

  openmccapstone
  Language:Jupyter Notebook 1

• Toshi-23 / Proliferation_Research

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

  openmc
  Language:Jupyter Notebook 1

• krislins / 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

  dagmcnuclear-physicsopenmcphysics-simulationpyhtonradiation-physicscubit-coreform
  Language:Jupyter Notebook 0

• AI-Pranto / triga-dataLibrary

  Nuclear data processing with ACEMAKER, FRENDY and NJOY

  frendynuclear-datanjoy2016openmcacemaker
  Language:Jupyter Notebook

• fusion-energy / openmc_csg_slice_plotter

  Plots slices of OpenMC native GSC surface geometry with hovertext surface identification

  csgopenmcplotslice
  Language:Python