MCNP
The Monte Carlo N-Particle (MCNP) radiation transport code is a Monte Carlo transport code developed by Las Alamos National Laboratory (LANL).
It supports over 37 different types of particles, and is widely used by nuclear engineers,
and nuclear physicists.
Here are 36 public repositories matching this topic...
A high-fidelity, free user input cylinder meshing tool for MCNP.
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Jun 6, 2021 - C
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May 26, 2023 - Assembly
A modular toolkit of fast and reliable libraries for neutronics analysis. Several command line tools are built with this core collection of crates.
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May 19, 2024 - Rust
MCNP6 Syntax highlighting and code snippets for VSCode. Written primarily for MCNP6.x input decks.
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Jan 5, 2024
Command line tool to convert MCNP mesh tallies to Visual ToolKit (VTK) formats. Supports all MCNPv6.2 legacy meshtal output formats, for both for rectangular and cylindrical meshes.
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Apr 14, 2024 - Rust
This repository automates the execution of MCNP simulations for a specific problem related to the sensitivity of a radiation detector
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May 20, 2024 - TeX
Tools to work with MCNP models and results
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May 22, 2024 - Jupyter Notebook
Command line tool to split meshtal results into individual files. This is extremely simple but nice to have when needed.
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Mar 30, 2024 - Rust
Thermal Hydraulic Sub-Channel Code for an Average Rod (Using MCNP for input values)
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May 8, 2018 - Python
Gists to aid in MCNP simulation and analysis
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Apr 17, 2020 - Jupyter Notebook
Tally table is a simple GUI program which extracts user defined tallies from a MCNP output.
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Feb 3, 2022 - Python
Created by Los Alamos National Laboratory
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- Website
- mcnp.lanl.gov
- Wikipedia
- Wikipedia