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...
Workflow and Template Toolkit for Simulation (WATTS)
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Apr 18, 2024 - Python
Tool for converting MCNP input files to OpenMC classes/XML
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Mar 25, 2024 - Python
a CAD to MC geometry conversion tool
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Apr 4, 2024 - C++
a companion for writing MCNP input decks
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Apr 2, 2021 - Python
MontePy is a Python library (API) to read, edit, and write MCNP input files.
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Apr 15, 2024 - 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 used for MCNP input deck syntax highlighting
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May 24, 2024 - Python
Tally table is a simple GUI program which extracts user defined tallies from a MCNP output.
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Feb 3, 2022 - Python
Tool to rename cells, surfaces, materials and universes in MCNP input files.
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Dec 5, 2022 - Python
A python library to allow ease of data reduction and data viewing for MCNP output file
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Jul 8, 2015 - Python
Created by Los Alamos National Laboratory
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- Website
- mcnp.lanl.gov
- Wikipedia
- Wikipedia