Metadata-Version: 2.4
Name: openmc_plasma_source
Version: 0.9.0
Summary: Creates tokamak and fusion point sources for OpenMC
Author: Rémi Delaporte-Mathurin
Project-URL: Homepage, https://github.com/fusion-energy/openmc-plasma-source
Project-URL: Bug Tracker, https://github.com/fusion-energy/openmc-plasma-source/issues
Keywords: python,neutron,fusion,source,openmc,energy,tokamak
Classifier: Programming Language :: Python :: 3
Classifier: License :: OSI Approved :: MIT License
Classifier: Operating System :: OS Independent
Requires-Python: >=3.9
Description-Content-Type: text/markdown
License-File: LICENSE.txt
Requires-Dist: numpy>=1.9
Requires-Dist: matplotlib>=3.2.2
Requires-Dist: NeSST>=1.1.0
Provides-Extra: tests
Requires-Dist: pytest>=5.4.3; extra == "tests"
Requires-Dist: pytest-cov; extra == "tests"
Requires-Dist: hypothesis; extra == "tests"
Requires-Dist: NeSST>=1.1.0; extra == "tests"
Requires-Dist: openmc_source_plotter; extra == "tests"
Dynamic: license-file

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# OpenMC-plasma-source

This python-based package offers a collection of pre-built [OpenMC](https://github.com/openmc-dev/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 a single `openmc.MeshSource` 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](https://doi.org/10.1016/j.fusengdes.2012.02.025).

```python
from openmc_plasma_source import tokamak_source

my_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},
)
```

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:

```python
my_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
)
```

For a more complete example check out the [example script](https://github.com/fusion-energy/openmc-plasma-source/blob/main/examples/tokamak_source_example.py).

![out](https://user-images.githubusercontent.com/40028739/135100022-330aa51c-e2a2-401c-9738-90f3e99c84d4.png)
 ![out](https://user-images.githubusercontent.com/40028739/135098576-a94709ef-96b4-4b8d-8fa0-76a201b6c5d2.png)

### Ring Source


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

```python
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.


```python
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
```
