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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...

lanl / mcnptools

Updated Sep 9, 2022
C++

kbat / mc-tools

Some Monte Carlo tools

monte-carlo povray root-cern mcnp fluka phits mcnpx

Updated Feb 21, 2024
Python

inr-kit / McCAD-Library

a CAD to MC geometry conversion tool

cmake opencascade mcnp cpluplus boostcpp

Updated Apr 4, 2024
C++

openmc-dev / openmc_mcnp_adapter

Tool for converting MCNP input files to OpenMC classes/XML

monte-carlo-simulation nuclear-engineering mcnp openmc

Updated Mar 25, 2024
Python

MontePy

idaholab / MontePy

MontePy is a Python library (API) to read, edit, and write MCNP input files.

api monte-carlo mcnp radiation-transport neutronics mcnp6

Updated Apr 15, 2024
Python

watts-dev / watts

Workflow and Template Toolkit for Simulation (WATTS)

python simulation templates workflow-automation mcnp nuclear-energy openmc

Updated Apr 18, 2024
Python

ukaea / neutronics_material_maker

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

library material serpent mcnp fispact openmc

Updated Sep 13, 2021
Python

fusion-energy / neutronics_material_maker

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

library database material serpent mcnp neutrons fispact openmc

Updated Apr 21, 2024
Python

alexhagen / wig

a companion for writing MCNP input decks

python simulation monte-carlo python-2 python-2-7 mcnp

Updated Apr 2, 2021
Python

IAEA-NDS / ACEMAKER

A code package to produce ACE-formatted files for MCNP calculations.

ace nuclear-data mcnp

Updated Dec 2, 2022
Fortran

danyalturkoglu / MCNP-syntax-highlighting

Syntax highlighting for MCNP input files

Updated Apr 6, 2018

arekfu / t4_geom_convert

Tool to convert MCNP geometries into TRIPOLI-4 geometries.

converter geometry mcnp mcnpx tripoli-4

Updated Apr 23, 2024
Python

ENEA-Fusion-Neutronics / MCNP-Input-Reader

The package for reading mcnp input in a pythonic way

radiation nuclear-engineering mcnp neutronics

Updated Jul 25, 2022
Python

MC-kit / mckit

Tools to work with MCNP models and results

tally nuclear-engineering mcnp radiation-physics meshtally

Updated May 6, 2024
Jupyter Notebook

travleev / numjuggler

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

python iter mcnp renumbering

Updated Dec 5, 2022
Python

repositony / vscode_mcnp

MCNP6 Syntax highlighting and code snippets for VSCode. Written primarily for MCNP6.x input decks.

syntax-highlighting snippets vscode-extension mcnp radiation-transport mcnp6

Updated Jan 5, 2024

repositony / meshtal

A small suite of tools for interacting with mcnp meshtally files. Supports all common output mesh formats, along with a few uncommon.

rust mesh monte-carlo-simulation vtk mcnp meshtally

Updated Mar 30, 2024
Rust

ZXG818 / CylinderMeshingTool

A high-fidelity, free user input cylinder meshing tool for MCNP.

udf coupling mcnp neutronics thermal-hydraulic

Updated Jun 6, 2021
C

mark-fortuna / MCNP-tally-table

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

nuclear-engineering mcnp nuclear-fusion nuclear-reactor

Updated Feb 3, 2022
Python

patrickpark910 / critloadexp

Python-automated MCNP simulation of the critical loading experiment for a TRIGA Mk.I reactor

python fortran mcnp

Updated Feb 12, 2021
Assembly

Created by Los Alamos National Laboratory

Followers: 5 followers
Website: mcnp.lanl.gov
Wikipedia: Wikipedia