Part 1 - Point source plotting

To perform a neutronics simulation a neutron source must also be defined.

This python notebook allows users to make a simple OpenMC point source and plot its energy, position and initial directions.

from IPython.display import IFrame
IFrame(src="https://www.youtube.com/embed/j9dT1Viqcu4", width=560, height=340)

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_energy, plot_source_position, plot_source_direction

This first code block creates an isotropic point source with 14 MeV monoenergetic neutrons.

# initialises a new source object
my_source = openmc.IndependentSource()

# sets the location of the source to x=0 y=0 z=0
my_source.space = openmc.stats.Point((0, 0, 0))

# sets the direction to isotropic
my_source.angle = openmc.stats.Isotropic()

# sets the energy distribution to 100% 14MeV neutrons
my_source.energy = openmc.stats.Discrete([14e6], [1])

this next code block runs openmc and creates a file containing information on the initial particles. The energy information is the extracted and plotted

plot_source_energy(my_source)

The next code block creates an isotropic point source with a fission energy distribution.

my_source_2 = openmc.IndependentSource()
my_source_2.space = openmc.stats.Point((0, 0, 0))
my_source_2.angle = openmc.stats.Isotropic()

# Documentation on the Watt distribution is here
# https://docs.openmc.org/en/stable/pythonapi/generated/openmc.data.WattEnergy.html
my_source_2.energy = openmc.stats.Watt(a=988000.0, b=2.249e-06)

plot_source_energy(my_source_2)

This code block creates an isotropic point source with a fusion energy distribution.

my_source_3 = openmc.IndependentSource()
my_source_3.space = openmc.stats.Point((0, 0, 0))
my_source_3.angle = openmc.stats.Isotropic()

# Documentation on the Muir distribution is here
# https://docs.openmc.org/en/stable/pythonapi/generated/openmc.stats.muir.html
my_source_3.energy = openmc.stats.muir(e0=14080000.0, m_rat=5.0, kt=20000.0)

plot_source_energy(my_source_3)

The following code block plots the birth location of the neutrons from a 14 MeV monoenergetic point source.

# Creates an isotropic point source with monoenergetic 14MeV neutrons
my_source_4 = openmc.IndependentSource()
my_source_4.space = openmc.stats.Point((0, 0, 0))
my_source_4.angle = openmc.stats.Isotropic()
my_source_4.energy = openmc.stats.Discrete([14e6], [1])

plot_source_position(my_source_4)

Finally, the following code block plots the birth direction of the neutrons from the same source.

# Creates an isotropic point source with monoenergetic 14MeV neutrons
my_source_5 = openmc.IndependentSource()
my_source_5.space = openmc.stats.Point((0, 0, 0))
my_source_5.angle = openmc.stats.Isotropic()
my_source_5.energy = openmc.stats.Discrete([14e6], [1])

plot_source_direction(my_source_5)

For convenience a plasma source package has been made that wraps openmc.IndependentSource and provides easy creation of point (and other) fusion sources. fusion-energy/openmc-plasma-source you could try plotting the energy this source

Learning Outcomes for Part 1:

• OpenMC can be used to create neutron point sources with different energy distributions.