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
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Apr 4, 2024 - C++
MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
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Oct 22, 2025 - Python
Tool for converting MCNP input files to OpenMC classes/XML
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Sep 22, 2025 - Python
Workflow and Template Toolkit for Simulation (WATTS)
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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.
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Jul 7, 2024
a companion for writing MCNP input decks
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Apr 2, 2021 - Python
A code package to produce ACE-formatted files for MCNP calculations.
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Jun 14, 2025 - Fortran
The package for reading mcnp input in a pythonic way
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Jul 25, 2022 - Python
MCNP SDEF to OpenMC conversion tool
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Sep 22, 2025 - Python
Tools to work with MCNP models and results
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Oct 22, 2025 - Python
Tool to rename cells, surfaces, materials and universes in MCNP input files.
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Sep 11, 2025 - Python
Created by Los Alamos National Laboratory
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