openmc-dev / validation
Collection of validation scripts, notebooks, results
☆11Updated last year
Alternatives and similar repositories for validation:
Users that are interested in validation are comparing it to the libraries listed below
- Collection of scripts for managing data for OpenMC☆29Updated last month
- OpenMC Monte Carlo Code☆17Updated 2 weeks ago
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆10Updated 2 years ago
- ☆18Updated this week
- Workflow and Template Toolkit for Simulation (WATTS)☆25Updated 5 months ago
- List of open source projects related to OpenMC☆24Updated 5 months ago
- OpenMC wrapped as a Moose App☆16Updated last year
- XSPlot - Neutron cross section plotter for isotopes☆14Updated last year
- ENDF/B pre-processing codes (PREPRO)☆17Updated last year
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆40Updated 7 months ago
- Stochastic Calculator Of Neutron transport Equation☆39Updated 3 weeks ago
- A depletion framework for OpenMC☆14Updated 7 years ago
- Plugins and command extensions for Coreform Cubit☆18Updated 8 months ago
- open-source Sn software☆25Updated this week
- Framework for REsearch in Nuclear ScIence and Engineering☆14Updated 3 years ago
- Sampling nuclear data and uncertainty☆50Updated last month
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆14Updated 2 weeks ago
- Native plotting GUI for model design and verification☆55Updated 3 weeks ago
- Collection of models for reactor physics benchmarks☆51Updated 4 months ago
- Format description for ACE nuclear data files☆20Updated 2 years ago
- ☆13Updated 10 months ago
- An OpenMC-based repo for studying proliferation issues in ARC-class fusion reactors.☆11Updated 5 months ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated 9 months ago
- A tool for making reproducible materials and standardizing use across several neutronics codes☆23Updated 4 months ago
- A deterministic finite-element-based solver for the neutron transport equation designed for the evaluation of acceleration methods.☆27Updated 3 years ago
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆28Updated last month
- Converts CAD file(s) such as STP and SAT to a h5m file compatible with DAGMC based simulations using the Cubit Python API☆12Updated last year
- Tool for converting MCNP input files to OpenMC classes/XML☆37Updated this week