openmc-plasma-source 0.3.1

Creates tokamak and ICF plasma sources for OpenMC

OpenMC-plasma-source

This python-based package offers a collection of pre-built OpenMC neutron sources for fusion applications.

Installation

OpenMC is required to use this package.

To install openmc-plasma-source, simply run:

pip install openmc-plasma-source

Usage

Tokamak Source

Create a source with a spatial and temperature distribution of a tokamak plasma. The OpenMC sources are ring sources which reduces the computational cost and the settings.xml file size. Each source has its own strength (or probability that a neutron spawns in this location).

The equations implemented here are taken from this paper.

from openmc_plasma_source import TokamakSource

my_source = TokamakSource(
    elongation=1.557,
    ion_density_centre=1.09e20,
    ion_density_peaking_factor=1,
    ion_density_pedestal=1.09e20,
    ion_density_separatrix=3e19,
    ion_temperature_centre=45.9,
    ion_temperature_peaking_factor=8.06,
    ion_temperature_pedestal=6.09,
    ion_temperature_separatrix=0.1,
    major_radius=9.06,
    minor_radius=2.92258,
    pedestal_radius=0.8 * 2.92258,
    mode="H",
    shafranov_factor=0.44789,
    triangularity=0.270,
    ion_temperature_beta=6
  ).make_openmc_sources()

For a more complete example check out the example script.

Ring Source

Create a ring source with temperature distribution of a 2000 eV plasma.

from openmc_plasma_source import FusionRingSource

my_plasma = FusionRingSource(
    angles = (0., 6.28318530718),  # input is in radians
    radius = 400,  # units in cm
    temperature = 20000.,  # ion temperature in eV
    fuel='DT'  # or 'DD'
)

Point Source

Create a point source with temperature distribution of a 2000 eV plasma.

from openmc_plasma_source import FusionPointSource

my_plasma = FusionPointSource(
    coordinate = (0, 0, 0),
    temperature = 20000.,  # ion temperature in eV
    fuel = 'DT'  # or 'DD'
)

Testing

To run the tests, simply run

pytest tests/