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 function returns an openmc.MeshSource, plasma volume, and fusion power. The source is backed by a CylindricalMesh. The plasma cross-section is discretised onto the mesh and each mesh voxel is assigned its own neutron source strength (its neutron emission rate in neutrons per second) and energy distribution based on the local ion temperature. The strength of the returned source, my_source.strength, is the total neutron emission rate of the plasma, which can be used to normalise tallies (OpenMC tally results are per source neutron).

The equations implemented here are taken from this paper.

from openmc_plasma_source import tokamak_source

make_source = tokamak_source(
    elongation=1.557,
    ion_density_centre=1.09e20,
    ion_density_pedestal=1.09e20,
    ion_density_peaking_factor=1,
    ion_density_separatrix=3e19,
    ion_temperature_centre=45.9e3,
    ion_temperature_pedestal=6.09e3,
    ion_temperature_separatrix=0.1e3,
    ion_temperature_peaking_factor=8.06,
    ion_temperature_beta=6,
    major_radius=906,
    minor_radius=292.258,
    pedestal_radius=0.8 * 292.258,
    mode="H",
    shafranov_factor=0.44789,
    triangularity=0.270,
    fuel={"D": 0.5, "T": 0.5},
)
my_source = make_source.source

The mesh resolution and the density of the internal sampling grid can be controlled with optional arguments, along with the toroidal extent of the source:

make_source = tokamak_source(
    # ... plasma parameters as above ...
    start_angle=0,                 # toroidal start angle in radians
    rotation_angle=2 * 3.14159,    # toroidal extent in radians
    mesh_resolution=(100, 100), # number of mesh bins in (r, z)
    grid_density=500,              # points per dimension in the (a, alpha) grid
)
my_source = make_source.source

The plasma volume and total fusion power are returned separately:

print(make_source.plasma_volume)  # plasma volume in m^3
print(make_source.fusion_power)   # total fusion power in MW

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 fusion_ring_source

my_source = fusion_ring_source(
    radius=700,
    start_angle=0.0,            # toroidal start angle in radians
    rotation_angle=2 * math.pi, # toroidal extent in radians (360deg source)
    temperature=20000.0,
    fuel={"D": 0.5, "T": 0.5},
)

Point Source

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

from openmc_plasma_source import fusion_point_source

my_source = fusion_point_source(
    coordinate=(0, 0, 0),
    temperature=20000.0,
    fuel={"D": 0.09, "T": 0.91},  # note this is mainly tritium fuel so that TT reactions are more likely
)

Testing

To run the tests, simply run

pytest tests