Part 3 - plasma source plotting#
As show in Part 1, OpenMC can be used to create point sources with different energy distributions. However, there are other ways to create neutron sources for use in neutronics simulations.
This python notebook allows users to plot the energy, position and initial directions of a parametric plasma source.
The plasma source used is from the This notebook makes use of the openmc-plasma-source package.
First import OpenMC and configure the nuclear data path
import openmc
from pathlib import Path
# Setting the cross section path to the correct location in the docker image.
# If you are running this outside the docker image you will have to change this path to your local cross section path.
openmc.config['cross_sections'] = Path.home() / 'nuclear_data' / 'cross_sections.xml'
# provides simple source plotting functions
# for more details here is a link to the package repository
# https://github.com/fusion-energy/openmc_source_plotter
from openmc_source_plotter import plot_source_direction, plot_source_position
This first code block creates a neutron source using the PlasmaSource class from the parametric_plasma_source package. The properties of the source are controlled by the input parameters.
from openmc_plasma_source import TokamakSource
my_sources = 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,
sample_size=50, # the number of individual sources to use to make a combined source
angles=(0, 2 * 3.14) # angle in radians
).make_openmc_sources() # returns a list of openmc sources
/home/jon/.neutronicsworkshop2/lib/python3.11/site-packages/openmc_plasma_source/tokamak_source.py:305: RuntimeWarning: invalid value encountered in power
* (1 - (r / self.pedestal_radius) ** self.ion_temperature_beta)
Makes a simple model for the sources.
settings = openmc.Settings()
settings.particles = 1
settings.batches = 1
settings.source = my_sources
materials = openmc.Materials()
sph = openmc.Sphere(r=1000000, boundary_type="vacuum")
cell = openmc.Cell(region=-sph)
geometry = openmc.Geometry([cell])
model = openmc.Model(geometry, materials, settings)
This code block then plots the birth location of each neutron, coloured by neutron birth energy.
plot = plot_source_position(this=model, n_samples=2000)
plot.show()
We can also plot the birth direction of each neutron.
plot = plot_source_direction(this=model, n_samples=500)
plot.show()