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 41 public repositories matching this topic...
a CAD to MC geometry conversion tool
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Apr 4, 2024 - C++
Tool for converting MCNP input files to OpenMC classes/XML
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Oct 23, 2024 - Python
MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.
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Nov 25, 2024 - Python
Workflow and Template Toolkit for Simulation (WATTS)
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Aug 27, 2024 - Python
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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Dec 2, 2022 - Fortran
MCNP6 Syntax highlighting and code snippets for VSCode. Written primarily for MCNP6.x input decks.
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Jul 7, 2024
The package for reading mcnp input in a pythonic way
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Jul 25, 2022 - Python
Tools to work with MCNP models and results
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Nov 25, 2024 - Jupyter Notebook
Tool to rename cells, surfaces, materials and universes in MCNP input files.
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Dec 5, 2022 - Python
Tools used for MCNP input deck syntax highlighting
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May 24, 2024 - Python
A high-fidelity, free user input cylinder meshing tool for MCNP.
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Jun 6, 2021 - C
Created by Los Alamos National Laboratory
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- Wikipedia
- Wikipedia