Alternatives and similar repositories for openmc-notebooks

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

• 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…
  ☆175Updated this week
• openmc-dev / openmc-ecosystem
  List of open source projects related to OpenMC
  ☆31Updated 8 months ago
• openmc-dev / openmc_mcnp_adapter
  Tool for converting MCNP input files to OpenMC classes/XML
  ☆53Updated last month
• openmc-dev / plotter
  Native plotting GUI for model design and verification
  ☆64Updated 2 weeks ago
• ukaea / openmc_workshop
  A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy
  ☆41Updated 4 years ago
• openmc-dev / data
  Collection of scripts for managing data for OpenMC
  ☆38Updated last month
• njoy / ENDFtk
  Toolkit for reading and interacting with ENDF-6 formatted files
  ☆41Updated 2 weeks ago
• svalinn / DAGMC
  Direct Accelerated Geometry Monte Carlo Toolkit
  ☆120Updated 11 months ago
• fusion-energy / openmc_regular_mesh_plotter
  A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.
  ☆14Updated 9 months ago
• openmsr / CAD_to_OpenMC
  code to target the conversion from a step-file to a h5m-geometry for neutronics
  ☆38Updated 7 months ago
• lanl / mcnptools
  ☆87Updated last year
• paulromano / endf-python
  Python ENDF Parser
  ☆39Updated last week
• luca-fiorito-11 / sandy
  Sampling nuclear data and uncertainty
  ☆55Updated 4 months ago
• GEOUNED-org / GEOUNED
  A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes
  ☆85Updated 2 months ago
• idaholab / MontePy
  MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
  ☆52Updated this week
• CambridgeNuclear / SCONE
  Stochastic Calculator Of Neutron transport Equation
  ☆55Updated 2 weeks ago
• mit-crpg / openmc-workshop
  Contains slides and example problems for OpenMC workshops and presentations
  ☆26Updated 3 years ago
• openmc-dev / openmc_cad_adapter
  Convert OpenMC Geometry Components to CAD
  ☆20Updated last year
• svalinn / pydagmc
  A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB
  ☆18Updated last month
• fusion-energy / neutronics_material_maker
  A tool for making reproducible materials and standardizing use across several neutronics codes
  ☆25Updated last year
• aurora-multiphysics / aurora
  Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.
  ☆43Updated last year
• Open-Sn / opensn
  Open-source Sn software
  ☆58Updated this week
• Fusion4Energy / F4Enix
  API for parsing and post-processing many MC simulations related files
  ☆17Updated last week
• eepeterson / openmc_fusion_benchmarks
  Automated V&V for fusion relevant CAD-based benchmarks
  ☆23Updated 3 weeks ago
• IAEA-NDS / PREPRO
  ENDF/B pre-processing codes (PREPRO)
  ☆22Updated 2 years ago
• JADE-V-V / JADE
  JADE, a novel nuclear data libraries V&V tool
  ☆34Updated 3 weeks ago
• njoy / NJOY21
  NJOY for the 21st Century
  ☆84Updated 2 years ago
• jlanversin / ONIX
  ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers c…
  ☆42Updated last year
• exasmr / openmc
  OpenMC Monte Carlo Code
  ☆24Updated last year
• makeclean / csg2csg
  Tools to translate between different CSG geometry types
  ☆40Updated 8 months ago