Some Monte Carlo tools

Tool for converting MCNP input files to OpenMC classes/XML

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

Workflow and Template Toolkit for Simulation (WATTS)

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

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

Python tools for MCNP

a companion for writing MCNP input decks

Tool to convert MCNP geometries into TRIPOLI-4 geometries.

MCNP SDEF to OpenMC conversion tool

Tools to work with MCNP models and results

The package for reading mcnp input in a pythonic way

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

Tools used for MCNP input deck syntax highlighting

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

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

Create an arbitrary parametric tokamak neutron source

a tool for creating axially symetric CSG geometry

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

Materials that are dependent on conditions