Workflow and Template Toolkit for Simulation (WATTS)
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Updated
May 30, 2024 - Python
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.
Workflow and Template Toolkit for Simulation (WATTS)
Tool for converting MCNP input files to OpenMC classes/XML
a companion for writing MCNP input decks
a CAD to MC geometry conversion tool
MontePy is a Python library (API) to read, edit, and write MCNP input files.
A code package to produce ACE-formatted files for MCNP calculations.
The package for reading mcnp input in a pythonic way
Tools used for MCNP input deck syntax highlighting
Tally table is a simple GUI program which extracts user defined tallies from a MCNP output.
Tool to rename cells, surfaces, materials and universes in MCNP input files.
A python library to allow ease of data reduction and data viewing for MCNP output file
Created by Los Alamos National Laboratory