mohamedlahdour / OpenNTPLinks
In this project we develop an OpenNTP package to solve the multi-group neutron transport equation.
☆11Updated 4 years ago
Alternatives and similar repositories for OpenNTP
Users that are interested in OpenNTP are comparing it to the libraries listed below
Sorting:
- Neutronics, Thermal-hydraulics and Structure Multi-physics Coupled Simulation of Heat Pipe Reactor☆10Updated 5 years ago
- Ducted Assembly Steady State Heat Transfer Software (DASSH)☆17Updated 2 years ago
- A deterministic finite-element-based solver for the neutron transport equation designed for the evaluation of acceleration methods.☆26Updated 4 years ago
- open-source Sn software☆34Updated this week
- Openmc-FEnicsx for muLtiphysics tutorIAl☆16Updated last week
- OpenMC wrapped as a Moose App☆16Updated 2 years ago
- Stochastic Calculator Of Neutron transport Equation☆43Updated 3 weeks ago
- ☆30Updated 3 months ago
- The National Reactor Innovation Center's (NRIC) Virtual Test Bed Repository☆53Updated last week
- A nuclear reactor lattice physics code.☆12Updated 3 weeks ago
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated this week
- OpenMC中文教程 ( 如何编译、安装、使用OpenMC教程 )☆41Updated 3 years ago
- List of open source projects related to OpenMC☆26Updated last week
- Convert OpenMC Geometry Components to CAD☆18Updated 6 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- Python ENDF Parser☆33Updated 10 months ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆34Updated 2 months ago
- Accelerated Discretized Geometry☆17Updated this week
- 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
- API for parsing and post-processing many MC simulations related files☆17Updated last week
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆11Updated 2 years ago
- ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers c…☆39Updated 5 months ago
- ENRICO: Exascale Nuclear Reactor Investigative COde☆65Updated last year
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆40Updated 11 months ago
- Vim plugin for MOOSE Framework input file format☆12Updated last year
- Collection of models for reactor physics benchmarks☆52Updated 8 months ago
- Native plotting GUI for model design and verification☆58Updated 4 months ago
- curated documents pertaining to early molten salt reactor research.☆22Updated last year
- JADE, a novel nuclear data libraries V&V tool☆29Updated this week
- Collection of scripts for managing data for OpenMC☆33Updated 4 months ago