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 49 public repositories matching this topic...
a CAD to MC geometry conversion tool
-
Updated
Apr 4, 2024 - C++
MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
-
Updated
Oct 20, 2025 - Python
Tool for converting MCNP input files to OpenMC classes/XML
-
Updated
Sep 22, 2025 - Python
Workflow and Template Toolkit for Simulation (WATTS)
-
Updated
Jun 26, 2025 - Python
MCNP6 Syntax highlighting and code snippets for VSCode. Written primarily for MCNP6.x input decks as a placeholder for the full LSP is implemented.
-
Updated
Jul 7, 2024
a companion for writing MCNP input decks
-
Updated
Apr 2, 2021 - Python
A code package to produce ACE-formatted files for MCNP calculations.
-
Updated
Jun 14, 2025 - Fortran
The package for reading mcnp input in a pythonic way
-
Updated
Jul 25, 2022 - Python
MCNP SDEF to OpenMC conversion tool
-
Updated
Sep 22, 2025 - Python
Tools to work with MCNP models and results
-
Updated
Oct 22, 2025 - Python
Tool to rename cells, surfaces, materials and universes in MCNP input files.
-
Updated
Sep 11, 2025 - Python
Created by Los Alamos National Laboratory
- Followers
- 7 followers
- Website
- github.com/topics/mcnp
- Wikipedia
- Wikipedia