Neutronics in Fusion: Difference between revisions

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# '''fusion-energy''' – This is a massive and highly recommended GitHub organization for any neutron physicist working in fusion. It hosts a wide range of projects for fusion neutronics and contains tools like fusion_neutron_utils, fusion_neutronics_workflow, and OpenMC plasma source utilities. These projects enable neutron source generation, Monte Carlo transport simulations, activation and decay analysis, and 3D modeling of fusion device neutronics, making it an indispensable resource for both research and development (GitHub: https://github.com/fusion-energy).

# '''aurora-multiphysics / aurora''' – Integrates neutron transport with multiphysics solvers, allowing simulation of neutron energy deposition, radiation effects, and thermal feedback in complex reactor geometries. (GitHub: https://github.com/aurora-multiphysics/aurora)

# '''Fusion-Power-Plant-Framework / tokamak-neutron-source''' – Provides Python tools to generate parametric tokamak neutron sources for Monte Carlo transport codes such as OpenMC and MCNP, based on plasma profiles and fusion reaction distributions. (GitHub: https://github.com/Fusion-Power-Plant-Framework/tokamak-neutron-source)

# '''Fusion-Power-Plant-Framework / tokamak-neutron-source''' – A Python package to create an arbitrary parametric tokamak neutron source for use with Monte Carlo radiation transport codes such as OpenMC and MCNP. It allows specification of plasma profiles (ion density, temperature, equilibrium) and exports source definitions for neutronics simulations. (GitHub: https://github.com/Fusion-Power-Plant-Framework/tokamak-neutron-source)

= See Also =

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 # '''fusion-energy''' – This is a massive and highly recommended GitHub organization for any neutron physicist working in fusion. It hosts a wide range of projects for fusion neutronics and contains tools like fusion_neutron_utils, fusion_neutronics_workflow, and OpenMC plasma source utilities. These projects enable neutron source generation, Monte Carlo transport simulations, activation and decay analysis, and 3D modeling of fusion device neutronics, making it an indispensable resource for both research and development (GitHub: https://github.com/fusion-energy).
 # '''aurora-multiphysics / aurora''' – Integrates neutron transport with multiphysics solvers, allowing simulation of neutron energy deposition, radiation effects, and thermal feedback in complex reactor geometries. (GitHub: https://github.com/aurora-multiphysics/aurora)
-# '''Fusion-Power-Plant-Framework / tokamak-neutron-source''' – Provides Python tools to generate parametric tokamak neutron sources for Monte Carlo transport codes such as OpenMC and MCNP, based on plasma profiles and fusion reaction distributions. (GitHub: https://github.com/Fusion-Power-Plant-Framework/tokamak-neutron-source)
 # '''Fusion-Power-Plant-Framework / tokamak-neutron-source''' – A Python package to create an arbitrary parametric tokamak neutron source for use with Monte Carlo radiation transport codes such as OpenMC and MCNP. It allows specification of plasma profiles (ion density, temperature, equilibrium) and exports source definitions for neutronics simulations. (GitHub: https://github.com/Fusion-Power-Plant-Framework/tokamak-neutron-source)
 = See Also =