Alternatives and similar repositories for openmc_fusion_benchmarks:

Users that are interested in openmc_fusion_benchmarks are comparing it to the libraries listed below

• svalinn / pydagmc
  A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB
  ☆16Updated last month
• openmc-dev / openmc-ecosystem
  List of open source projects related to OpenMC
  ☆24Updated 6 months ago
• fusion-energy / openmc_regular_mesh_plotter
  A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.
  ☆11Updated 10 months ago
• paulromano / endf-python
  Python ENDF Parser
  ☆31Updated 7 months ago
• ukaea / openmc_workshop
  A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy
  ☆39Updated 3 years ago
• pshriwise / xdg
  Accelerated Discretized Geometry
  ☆13Updated this week
• ERMETE-Lab / MP-OFELIA
  Openmc-FEnicsx for muLtiphysics tutorIAl
  ☆14Updated 7 months ago
• jlball / arc-nonproliferation
  An OpenMC-based repo for studying proliferation issues in ARC-class fusion reactors.
  ☆11Updated 6 months ago
• openmc-dev / openmc_cad_adapter
  Convert OpenMC Geometry Components to CAD
  ☆17Updated 4 months ago
• RemDelaporteMathurin / h-transport-materials
  Easily access hydrogen transport properties
  ☆15Updated 3 months ago
• openmc-dev / openmc_mcnp_adapter
  Tool for converting MCNP input files to OpenMC classes/XML
  ☆39Updated last month
• lanl / TOGA
  Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections
  ☆10Updated 2 years ago
• JADE-V-V / JADE
  JADE, a novel nuclear data libraries V&V tool
  ☆25Updated last week
• fispact / pypact
  A Python package for parsing FISPACT-II output
  ☆28Updated last year
• Open-Sn / opensn
  open-source Sn software
  ☆27Updated this week
• openmc-data-storage / isotope-xs-plotter
  XSPlot - Neutron cross section plotter for isotopes
  ☆14Updated last year
• exasmr / openmc
  OpenMC Monte Carlo Code
  ☆17Updated last month
• openmc-data-storage / openmc_data_downloader
  A Python package for downloading h5 cross section files for use in OpenMC.
  ☆12Updated 2 months ago
• fusion-energy / neutronics_material_maker
  A tool for making reproducible materials and standardizing use across several neutronics codes
  ☆23Updated 5 months ago
• openmc-dev / plotter
  Native plotting GUI for model design and verification
  ☆56Updated 2 months ago
• toshihikokawano / DeCE
  ENDF-6 data interface and nuclear data evaluation assist code
  ☆15Updated last week
• CambridgeNuclear / SCONE
  Stochastic Calculator Of Neutron transport Equation
  ☆42Updated 2 months ago
• ezsolti / RFP
  ☆30Updated last month
• LLNL / fudge
  For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data
  ☆29Updated 2 months ago
• fusion-energy / openmc-plasma-source
  Creates a plasma source as an openmc.source object from input parameters that describe the plasma
  ☆33Updated 3 months ago
• ukaea / neutronics_material_maker
  A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak
  ☆19Updated 3 years ago
• luca-fiorito-11 / sandy
  Sampling nuclear data and uncertainty
  ☆51Updated last week
• njoy / ENDFtk
  Toolkit for reading and interacting with ENDF-6 formatted files
  ☆38Updated 2 weeks ago
• makeclean / csg2csg
  Tools to translate between different CSG geometry types
  ☆36Updated last year
• aurora-multiphysics / aurora
  Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.
  ☆40Updated 9 months ago