MCNP

MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.

Tool for converting MCNP input files to OpenMC classes/XML

Some Monte Carlo tools

Workflow and Template Toolkit for Simulation (WATTS)

A tool for making reproducible materials and standardizing use across several neutronics codes

Python tools for MCNP

A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak

a companion for writing MCNP input decks

Tools to work with MCNP models and results

Tool to convert MCNP geometries into TRIPOLI-4 geometries.

The package for reading mcnp input in a pythonic way

MCNP SDEF to OpenMC conversion tool

Create an arbitrary parametric tokamak neutron source

Tool to rename cells, surfaces, materials and universes in MCNP input files.

notepad++ plugin for MCNP deck development. shows informative popups for selected cell/surface/physics cards. Inbuilt error checking.

Tools used for MCNP input deck syntax highlighting

Materials that are dependent on conditions

a tool for creating axially symetric CSG geometry

Transfer meta information on cells from an STP file to MCNP model. Set materials and densities according to tags specified in STP model tree.

Tally table is a simple GUI program which extracts user defined tallies from a MCNP output.