rachelslaybaugh / NE250Links
Nuclear Reactor Theory
☆19Updated 7 years ago
Alternatives and similar repositories for NE250
Users that are interested in NE250 are comparing it to the libraries listed below
Sorting:
- Collection of models for reactor physics benchmarks☆52Updated 8 months ago
- Nuclear data processing with legacy NJOY☆109Updated 2 weeks ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆57Updated last month
- ENRICO: Exascale Nuclear Reactor Investigative COde☆65Updated last year
- Collection of scripts for managing data for OpenMC☆33Updated 4 months ago
- Direct Accelerated Geometry Monte Carlo Toolkit☆112Updated 3 months ago
- In this project we develop an OpenNTP package to solve the multi-group neutron transport equation.☆11Updated 4 years ago
- NJOY for the 21st Century☆79Updated last year
- ☆30Updated 3 months ago
- Repository for Moltres, a code for simulating Molten Salt Reactors☆68Updated last week
- Native plotting GUI for model design and verification☆58Updated 4 months ago
- curated documents pertaining to early molten salt reactor research.☆22Updated last year
- Stochastic Calculator Of Neutron transport Equation☆43Updated 3 weeks ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆34Updated 2 months ago
- List of open source projects related to OpenMC☆26Updated last week
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- Tools to translate between different CSG geometry types☆37Updated last week
- ERSN-OpenMC is a Graphical User Interface for OpenMC Monte Carlo particle transport simulation code, originally developed by Jaafar EL Ba…☆13Updated 7 years ago
- The National Reactor Innovation Center's (NRIC) Virtual Test Bed Repository☆53Updated last week
- OpenMC wrapped as a Moose App☆16Updated 2 years ago
- Sampling nuclear data and uncertainty☆51Updated last month
- A method of characteristics code for nuclear reactor physics calculations.☆163Updated 10 months ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆41Updated 11 months ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆46Updated last week
- ERSN-OpenMC is a Graphical User Interface for OpenMC Monte Carlo particle transport simulation code, originally developed by Jaafar EL Ba…☆30Updated last year
- OpenMC Monte Carlo Code☆18Updated 4 months ago
- OpenMC Workshop for Canadian Nuclear Laboratories☆10Updated 8 years ago
- Contains slides and example problems for OpenMC workshops and presentations☆23Updated 2 years ago
- Python ENDF Parser☆33Updated 10 months ago
- open-source Sn software☆34Updated this week