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 35 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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Mar 25, 2024 - Python
MontePy is a Python library (API) to read, edit, and write MCNP input files.
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Apr 15, 2024 - Python
Workflow and Template Toolkit for Simulation (WATTS)
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Apr 18, 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
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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May 14, 2024 - Jupyter Notebook
Tool to rename cells, surfaces, materials and universes in MCNP input files.
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Dec 5, 2022 - Python
MCNP6 Syntax highlighting and code snippets for VSCode. Written primarily for MCNP6.x input decks.
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Jan 5, 2024
A high-fidelity, free user input cylinder meshing tool for MCNP.
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Jun 6, 2021 - C
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