arfc / moltresLinks
Repository for Moltres, a code for simulating Molten Salt Reactors
☆68Updated last week
Alternatives and similar repositories for moltres
Users that are interested in moltres are comparing it to the libraries listed below
Sorting:
- ENRICO: Exascale Nuclear Reactor Investigative COde☆65Updated last year
- The Cyclus Nuclear Fuel Cycle Simulator☆82Updated 3 weeks ago
- Computational library for chemical thermodynamics and phase equilibrium calculation. Multiphysics and standalone estimations of chemical …☆54Updated 4 months ago
- High-Fidelity Multiphysics☆109Updated last week
- MC/DC: Monte Carlo Dynamic Code☆41Updated last month
- Stochastic Calculator Of Neutron transport Equation☆43Updated 3 weeks ago
- Ducted Assembly Steady State Heat Transfer Software (DASSH)☆17Updated 2 years ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆57Updated last month
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆41Updated 11 months ago
- THOR is a radiation transport code for unstructured meshes.☆14Updated 8 months ago
- Direct Accelerated Geometry Monte Carlo Toolkit☆112Updated 3 months ago
- Workflow and Template Toolkit for Simulation (WATTS)☆30Updated last month
- The National Reactor Innovation Center's (NRIC) Virtual Test Bed Repository☆53Updated last week
- A tool for making reproducible materials and standardizing use across several neutronics codes☆24Updated 8 months ago
- open-source Sn software☆34Updated this week
- A set of radiation transport mini-applications used for performance optimization on HPC systems.☆29Updated 6 years ago
- curated documents pertaining to early molten salt reactor research.☆22Updated last year
- Collection of scripts for managing data for OpenMC☆33Updated 4 months ago
- Collection of models for reactor physics benchmarks☆52Updated 8 months ago
- continuous energy monte carlo neutron transport in general geometries on GPUs☆43Updated 8 years ago
- A deterministic finite-element-based solver for the neutron transport equation designed for the evaluation of acceleration methods.☆26Updated 4 years ago
- Simulation of transmutation and decay in nuclear systems☆28Updated 4 months ago
- OpenMC wrapped as a Moose App☆16Updated 2 years ago
- Native plotting GUI for model design and verification☆58Updated 4 months ago
- List of open source projects related to OpenMC☆26Updated last week
- ☆30Updated 3 months ago
- In this project we develop an OpenNTP package to solve the multi-group neutron transport equation.☆11Updated 4 years ago
- ☆19Updated last week
- detailed cad model of the msre☆36Updated 3 months ago
- Sampling nuclear data and uncertainty☆51Updated last month