Alternatives and similar repositories for openmc-notebooks

Users that are interested in openmc-notebooks are comparing it to the libraries listed below

• ukaea / openmc_workshop
  A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy
  ☆39Updated 3 years ago
• mit-crpg / openmc-workshop
  Contains slides and example problems for OpenMC workshops and presentations
  ☆23Updated 3 years ago
• paulromano / endf-python
  Python ENDF Parser
  ☆33Updated 10 months ago
• openmc-dev / plotter
  Native plotting GUI for model design and verification
  ☆59Updated 5 months ago
• openmc-dev / openmc-ecosystem
  List of open source projects related to OpenMC
  ☆26Updated 3 weeks ago
• mit-crpg / BEAVRS
  A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant
  ☆36Updated 3 months ago
• GEOUNED-org / GEOUNED
  A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes
  ☆70Updated last month
• fusion-energy / neutronics-workshop
  A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutroni…
  ☆153Updated this week
• openmc-dev / openmc_mcnp_adapter
  Tool for converting MCNP input files to OpenMC classes/XML
  ☆42Updated last month
• fusion-energy / openmc-plasma-source
  Creates a plasma source as an openmc.source object from input parameters that describe the plasma
  ☆36Updated this week
• openmc-dev / openmc_cad_adapter
  Convert OpenMC Geometry Components to CAD
  ☆18Updated 7 months ago
• eepeterson / openmc_fusion_benchmarks
  Automated V&V for fusion relevant CAD-based benchmarks
  ☆23Updated 2 weeks ago
• CambridgeNuclear / SCONE
  Stochastic Calculator Of Neutron transport Equation
  ☆44Updated 2 weeks ago
• fusion-energy / neutronics_material_maker
  A tool for making reproducible materials and standardizing use across several neutronics codes
  ☆24Updated 8 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 last month
• openmc-dev / data
  Collection of scripts for managing data for OpenMC
  ☆33Updated 5 months ago
• JADE-V-V / JADE
  JADE, a novel nuclear data libraries V&V tool
  ☆29Updated 3 weeks ago
• aurora-multiphysics / aurora
  Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.
  ☆41Updated last year
• luca-fiorito-11 / sandy
  Sampling nuclear data and uncertainty
  ☆51Updated last week
• mit-crpg / openmc-reactor-examples
  Catalogs a plethora of reactor inputs for OpenMC
  ☆25Updated 3 months ago
• svalinn / DAGMC
  Direct Accelerated Geometry Monte Carlo Toolkit
  ☆111Updated 4 months ago
• svalinn / parastell
  Parametric 3-D CAD modeling toolset for stellarator fusion devices
  ☆32Updated last week
• Fusion4Energy / F4Enix
  API for parsing and post-processing many MC simulations related files
  ☆17Updated last week
• njoy / ENDFtk
  Toolkit for reading and interacting with ENDF-6 formatted files
  ☆41Updated 2 weeks ago
• Open-Sn / opensn
  open-source Sn software
  ☆35Updated this week
• svalinn / pydagmc
  A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB
  ☆16Updated last month
• exasmr / openmc
  OpenMC Monte Carlo Code
  ☆18Updated 4 months ago
• mit-crpg / benchmarks
  Collection of models for reactor physics benchmarks
  ☆53Updated 8 months ago
• IAEA-NDS / endf-parserpy
  Python package for reading, writing, verifying and translating ENDF-6 formatted files
  ☆16Updated 2 weeks ago
• lanl / mcnptools
  ☆79Updated last year