t2015q / OpenMC-Users-GuideLinks
OpenMC中文教程 ( 如何编译、安装、使用OpenMC教程 )
☆42Updated 3 years ago
Alternatives and similar repositories for OpenMC-Users-Guide
Users that are interested in OpenMC-Users-Guide are comparing it to the libraries listed below
Sorting:
- In this project we develop an OpenNTP package to solve the multi-group neutron transport equation.☆13Updated 4 years ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- OpenMC wrapped as a Moose App☆16Updated 2 years ago
- Neutronics, Thermal-hydraulics and Structure Multi-physics Coupled Simulation of Heat Pipe Reactor☆10Updated 5 years ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆36Updated 3 months ago
- Collection of scripts for managing data for OpenMC☆33Updated 5 months ago
- Collection of models for reactor physics benchmarks☆53Updated 8 months ago
- open-source Sn software☆35Updated this week
- Native plotting GUI for model design and verification☆59Updated 5 months ago
- Stochastic Calculator Of Neutron transport Equation☆44Updated 2 weeks ago
- ☆31Updated last week
- A deterministic finite-element-based solver for the neutron transport equation designed for the evaluation of acceleration methods.☆27Updated 4 years ago
- Openmc-FEnicsx for muLtiphysics tutorIAl☆17Updated last month
- The National Reactor Innovation Center's (NRIC) Virtual Test Bed Repository☆55Updated this week
- List of open source projects related to OpenMC☆26Updated 3 weeks ago
- Tool for converting MCNP input files to OpenMC classes/XML☆42Updated last month
- Ducted Assembly Steady State Heat Transfer Software (DASSH)☆17Updated 2 years ago
- API for parsing and post-processing many MC simulations related files☆17Updated last week
- Convert OpenMC Geometry Components to CAD☆18Updated 7 months ago
- ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers c…☆40Updated 6 months ago
- ENRICO: Exascale Nuclear Reactor Investigative COde☆65Updated last year
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated 2 weeks ago
- Nuclear data processing with legacy NJOY☆110Updated last month
- Tools to translate between different CSG geometry types☆38Updated 3 weeks ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆58Updated 2 months ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆70Updated this week
- An unstructured mesh library for automated method of characteristic mesh generation☆10Updated 9 months ago
- NJOY for the 21st Century☆79Updated last year
- Catalogs a plethora of reactor inputs for OpenMC☆25Updated 3 months ago
- OpenMC Monte Carlo Code☆18Updated 4 months ago