Alternatives and similar repositories for neutronics_material_maker

Users that are interested in neutronics_material_maker 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
• openmc-dev / openmc-ecosystem
  List of open source projects related to OpenMC
  ☆24Updated 8 months ago
• eepeterson / openmc_fusion_benchmarks
  ☆22Updated this week
• svalinn / pydagmc
  A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB
  ☆16Updated 3 weeks ago
• paulromano / endf-python
  Python ENDF Parser
  ☆33Updated 9 months ago
• lanl / TOGA
  Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections
  ☆10Updated 2 years ago
• IAEA-NDS / PREPRO
  ENDF/B pre-processing codes (PREPRO)
  ☆19Updated last year
• LLNL / fudge
  For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data
  ☆29Updated 3 weeks ago
• Fusion4Energy / F4Enix
  API for parsing and post-processing many MC simulations related files
  ☆15Updated 3 weeks 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 week
• luca-fiorito-11 / sandy
  Sampling nuclear data and uncertainty
  ☆51Updated last week
• njoy / ENDFtk
  Toolkit for reading and interacting with ENDF-6 formatted files
  ☆40Updated last week
• toshihikokawano / DeCE
  ENDF-6 data interface and nuclear data evaluation assist code
  ☆16Updated 2 weeks ago
• JADE-V-V / JADE
  JADE, a novel nuclear data libraries V&V tool
  ☆27Updated this week
• openmc-dev / data
  Collection of scripts for managing data for OpenMC
  ☆31Updated 3 months ago
• openmc-data-storage / openmc_data_downloader
  A Python package for downloading h5 cross section files for use in OpenMC.
  ☆13Updated 4 months ago
• openmc-dev / openmc_mcnp_adapter
  Tool for converting MCNP input files to OpenMC classes/XML
  ☆41Updated 3 months ago
• makeclean / csg2csg
  Tools to translate between different CSG geometry types
  ☆37Updated last year
• fusion-energy / magnetic_fusion_openmc_dagmc_paramak_example
  ☆15Updated last year
• fispact / pypact
  A Python package for parsing FISPACT-II output
  ☆28Updated 3 weeks ago
• openmc-dev / openmc_cad_adapter
  Convert OpenMC Geometry Components to CAD
  ☆17Updated 5 months ago
• openmc-data-storage / isotope-xs-plotter
  XSPlot - Neutron cross section plotter for isotopes
  ☆14Updated last year
• ukaea / SpaceClaim_API_NeutronicsTools
  Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis
  ☆14Updated 4 years ago
• jlball / arc-nonproliferation
  An OpenMC-based repo for studying proliferation issues in ARC-class fusion reactors.
  ☆11Updated 8 months ago
• idaholab / MontePy
  MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
  ☆43Updated this week
• openmsr / msr-archive
  curated documents pertaining to early molten salt reactor research.
  ☆22Updated last year
• Open-Sn / opensn
  open-source Sn software
  ☆33Updated this week
• watts-dev / watts
  Workflow and Template Toolkit for Simulation (WATTS)
  ☆30Updated last week
• fusion-energy / neutronics_material_maker
  A tool for making reproducible materials and standardizing use across several neutronics codes
  ☆24Updated 7 months ago
• fusion-energy / openmc-plasma-source
  Creates a plasma source as an openmc.source object from input parameters that describe the plasma
  ☆33Updated 5 months ago