tokamak-neutron-source 0.1.4

Create an arbitrary parametric tokamak neutron source for OpenMC and MCNP

Introduction

tokamak-neutron-source is a package that provides a flexible and high-fidelity fusion neutron source for tokamaks in OpenMC and other Monte Carlo radiation transport codes.

Installation

We don't try to manage the installation of your neutronics codes. We recommend you install your neutronics code first. If you are using tokamak-neutron-source to create an OpenMC source you can create a simple install of OpenMC using conda with:

conda install -c conda-forge 'openmc>=0.15.0'

To install the latest release of tokamak-neutron-source

pip install tokamak-neutron-source

Inputs

A tokamak neutron source can be created by specifing the plasma ion density and temperature profiles, and a description of the plasma magneto-hydrodynamic equilibrium.

Profiles can be specified in terms of arrays or as typical parameterisations, such as a parabolic-pedestal parameterisation.

Equilibrium information can be specified via an EQDSK file or as a parameterisation, such as the one found in Fausser et al., 2012.

Outputs

A source object can be used to create an idiomatic source for use in OpenMC or exported as an sdef or h5 file for use in OpenMC and MCNP6.

A neutron source from some typical parameterised profiles and a Fausser flux surface parameterisation:

A neutron source from some arbitrary profiles and a free-boundary equilibrium: