neutronics

There are 2 repositories under neutronics topic.

• openmc-dev / openmc

  OpenMC Monte Carlo Code

  computational-physicshigh-performance-computingmonte-carlo-simulationneutron-transportneutronicsnuclear-datanuclear-energynuclear-engineeringnuclear-fusionopenmcparticle-transportphoton-transportradiation-transport
  Language:Python 904

• fusion-energy / neutronics-workshop

  A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutronics tools

  containerisedcoursedagmcenergyfluxfusionlearningneutronicsneutronsnuclearopenmcphotonssimulationsspectrumtrainingtritiumtutorial
  Language:Jupyter Notebook 167

• stellarmesh / stellarmesh

  Meshing library for nuclear workflows

  cadquerydagmcgmshmoabmonte-carlo-simulationneutronicsopenmcbuild123dopencascade
  Language:Python 54
• SCONE

  CambridgeNuclear / SCONE

  Stochastic Calculator Of Neutron transport Equation

  monte-carloneutronicsradiation-transportscone
  Language:Fortran 51
• MontePy

  idaholab / MontePy

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

  mcnp6monte-carloneutronicsapiradiation-transportmcnphacktoberfest
  Language:Python 48

• CEMeNT-PSAAP / MCDC

  MC/DC: Monte Carlo Dynamic Code

  high-performance-computingmonte-carlo-simulationneutron-transportneutronicsnuclear-engineeringparticle-transportradiation-transportparallelpythonsimulation
  Language:Python 45

• fusion-energy / cad_to_dagmc

  Convert CAD geometry (STP files) or Cadquery assemblies to DAGMC h5m files

  cadconverterdagmcmeshneutronicsopenmc
  Language:Python 37

• ukaea / paramak

  Create parametric 3D fusion reactor CAD models

  caddesignfusiongeometryneutronicsparametric
  37

• openmc-dev / openmc-ecosystem

  List of open source projects related to OpenMC

  monte-carlo-simulationmultiphysics-simulationneutronicsnuclear-engineeringopenmc
  29

• ERMETE-Lab / MP-OFELIA

  Openmc-FEnicsx for muLtiphysics tutorIAl

  dolfinxmulti-physicsneutronicsopenmcpythonthermal-hydraulic
  Language:Jupyter Notebook 16

• NCSU-NCSG / THOR

  THOR is a radiation transport code for unstructured meshes.

  nuclearnuclear-engineeringnuclear-physicsneutronicsdiscrete-ordinates-methodneutron-transportnuclear-fissionparticle-transportphoton-transportradiation-transportunstructured-meshesdeterministic-transportcomputational-physics
  Language:Fortran 16

• fusion-energy / fusion_neutronics_workflow

  Combines open source packages to produce an automated fusion specific neutronics workflow

  neutronicsautomaticsimulationfusionneutronphotoncadopenmcdagmcparamak
  Language:Python 15

• fusion-energy / openmc_regular_mesh_plotter

  A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.

  neutronicsmeshtallydosemapregularstructuredplot
  Language:Python 14

• ukaea / SpaceClaim_API_NeutronicsTools

  Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis

  modellingcadgeometryneutronics
  Language:C# 14

• fusion-energy / model_benchmark_zoo

  A collection of neutronics models for comparing neutronics simulations in both CAD and CSG formats.

  benchmarkcadcomparisonconstructive-solid-geometriesgeometryneutronicssimulationvalidation
  Language:Python 10

• open-radiation-sources / parametric-plasma-source

  neutronicsparticle-physicsneutronphotonopenmcparametric-plasma-source
  Language:C++ 9

• egor1abs / openmc_sdef_parser

  MCNP SDEF to OpenMC conversion tool

  fusionmcnpneutronicsopenmcplasma-physics
  Language:Python 8

• ENEA-Fusion-Neutronics / MCNP-Input-Reader

  The package for reading mcnp input in a pythonic way

  mcnpneutronicsradiationnuclear-engineering
  Language:Python 8

• fusion-energy / inertial_fusion_openmc_dagmc_paramak_example

  A minimal example implementation of an open source method of making DAGMC geometry with Paramak and simulating tritium production with OpenMC

  fusionneutronicsopen-sourcetritiumopenmcdagmcparamak
  Language:Python 7

• fusion-energy / openmc-dagmc-wrapper

  A Python package that extends OpenMC base classes to provide convenience features and standardized tallies when simulating DAGMC geometry with OpenMC.

  neutronicssimulationsopenmcparametricdagmc
  Language:Python 7

• terrapower / armicontrib-dif3d

  DIF3D plugin to the ARMI nuclear reactor analysis framework

  nuclear-engineeringneutronicsenergyclean-energy
  Language:Python 7

• fusion-energy / spectrum_plotter

  A Python package for creating publication quality plots for neutron / photon / particle spectrum

  neutronphotonspectraspectrumdoseneutronicsgraphplot
  Language:Python 4

• ijs-f8 / polygontorus

  a tool for creating axially symetric CSG geometry

  fusionmcnpneutronicsopenmctokamak
  Language:Python 4

• NielsBongers / neutron-transport-monte-carlo

  Energy-dependent neutron transport Monte Carlo implemented in Rust.

  monte-carloneutron-diffusionneutron-transportneutronicsnuclear-engineeringnuclear-physicsnuclear-energyparticle-transport
  Language:Rust 4

• tumregels / asciiviewer

  A pretty viewer for XSM files generated by DRAGON/DONJON or APOLLO neutronic codes

  dragondonjonappoloneutronicsnuclear-engineeringnuclear-reactorxsm
  Language:Python 4

• fusion-energy / stl_to_h5m

  Convert non overlapping STL files into a DAGMC h5m file complete with material tags and ready for use in neutronics simulations.

  dagmcopenmch5mcadneutronicsstl
  Language:Python 3

• fusion-energy / vertices_to_h5m

  Converts mesh vertices and connectivity to h5m geometry files compatible with DAGMC simulations

  coordinatesdagmch5mmeshneutronicstrianglesvertices
  Language:Python 3

• nfherrin / OpenXSConverter

  An open source utility to convert various publicly available macroscopic nuclear cross section formats

  nuclearnuclear-engineeringnuclear-physicsnuclear-dataneutronics
  Language:Fortran 3

• openmc-data-storage / material_xs_plotter

  XSPlot - Neutron cross section plotter for materials

  neutroncrosssectionmaterialneutronicsnuclearbarnsgraphplot
  Language:Python 3

• ZXG818 / CylinderMeshingTool

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

  couplingmcnpneutronicsthermal-hydraulicudf
  Language:C 3

• patrikasvanagas / MC-Thermal-Neutron-Transport

  Isotropic Monte Carlo simulations examining thermal neutron statistics in water, lead, and graphite, including animated neutron history, delta tracking and spherical geometry adaptations.

  computational-physicsmonte-carlo-simulationneutron-transportneutronicsnuclearnuclear-physicsphysics-3dphysics-based-animationphysics-simulationscientific-computing
  Language:Jupyter Notebook 2

• quentinducasse / decima

  DECIMA (Data Extraction & Contextual Inference for MCNP Analysis) — MCNP PTRAC parsing, advanced analysis, and contextual inference tool combining a nuclear physics knowledge graph and an LLM-powered assistant to deliver accurate, no-code answers for neutron and photon transport simulations, built on the official MCNPTools library from LANL.

  fissionfusion-energyknowledge-graphlarge-language-modelsllmmcnpneutronicsnuclear-simulationopen-source-softwareparticle-transportptracptrackmctal
  Language:C++ 1

• repositony / mesh2vtk

  Command line tool to convert MCNP mesh tallies to Visual ToolKit (VTK) formats. Supports all MCNPv6.2 legacy meshtal output formats, for both for rectangular and cylindrical meshes.

  mcnpmeshmeshtallyneutronicsvtk
  Language:Rust 1

• repositony / ntools_doc

  Public documentation and API for Tony's neutronics toolbox. The core library will remain private until I have time to finalise the API.

  mcnpmonte-carloneutronicsrustscientific
  Language:HTML 1

• sg2405 / Neutronics_MonteCarlo

  Simple code to simulate Neutron Scattering in Non-Radiative matter using Monte Carlo simulations

  monte-carlo-methodsmonte-carlo-simulationneutronicsnuclear-physicsscattering
  Language:R 1

• etare8 / fission-reactor-physics

  Fission Reactor Physics Homeworks

  fissionneutronicspythonreactor-physics
  Language:Jupyter Notebook