Different distribution of Neutron density in openmc_plasma_source

Shailja · April 1, 2024, 10:19am

Hii
I was trying to model a tokamak source using openmc_plasma_source utility in openmc and I added following lines to source definition as given in the test code (similar to tokamak_source.py in example folder of utility) to print neutron_source_density attributes as it is shown in graphs of the same.

data=my_plasma.neutron_source_density
print(data)

I got 1000 diff points denoting neutron density(/src/cm3/s), till this no issue.

but as soon as I re-run I got those 1000 points different. I calculated average of neutron source density in each case which differs by 1-3 percent. I don’ t understand that why this distribution nos. are changing, does’nt it should be same at every run when no change in any other input?

I am sorry if I asked something very trivial but I really have no idea.

I am putting here input code which I tried (changed tokamak_source.py file with my parameters)

import openmc
from openmc_plasma_source import TokamakSource
import openmc_plasma_source.plotting as plt2
import matplotlib.pyplot as plt
from matplotlib.colors import Normalize
import matplotlib as mpl
from matplotlib import cm
import numpy as np

minimal geometry

sphere_surface = openmc.Sphere(r=1000.0, boundary_type=“vacuum”)
cell = openmc.Cell(region=-sphere_surface)
geometry = openmc.Geometry([cell])

create a plasma source

my_plasma = TokamakSource(
elongation=2,
ion_density_centre=1.78e20,
ion_density_peaking_factor=1,
ion_density_pedestal=1.78e20,
ion_density_separatrix=3e19,

ion_temperature_centre=45.9,

ion_temperature_centre=4.71,

ion_temperature_centre=45.9,
ion_temperature_peaking_factor=8.06,

ion_temperature_pedestal=6.09,

ion_temperature_pedestal=0.427,

ion_temperature_pedestal=6.09,
ion_temperature_separatrix=0.1,

major_radius=9.06,

minor_radius=2.92258,

major_radius=150,
minor_radius=90,
pedestal_radius=0.8 * 90,
mode="H",
shafranov_factor=0.44789,

triangularity=0.270,

 triangularity=0.5,
ion_temperature_beta=4,

ion_temperature_beta=3.86,

)

Tell OpenMC we’re going to use our custom source

settings = openmc.Settings()
settings.run_mode = “fixed source”
settings.batches = 10
settings.particles = 1000
settings.source = my_plasma.sources

model = openmc.model.Model(
materials=None, geometry=geometry, settings=settings
)

model.run()

data=my_plasma.neutron_source_density
print(data)

create a figure and axis

fig, ax = plt.subplots()

plot the tokamak source with a color scale

norm = Normalize()
scatter = plt2.scatter_tokamak_source(my_plasma, quantity=“neutron_source_density”, ax=ax, cmap=‘viridis’, norm=norm)

create a ScalarMappable object for the colorbar

sm = plt.cm.ScalarMappable(cmap=‘viridis’, norm=norm )
#sm = plt.cm.ScalarMappable(cmap=‘viridis’)
sm.set_array()

add a colorbar

plt.colorbar(sm, ax=ax, label=‘Neutron Source Density [per cm^3/s]’)

show the plot

plt.show()

plt2.plot_tokamak_source_3D(my_plasma, quantity=“neutron_source_density”)

show the plot

plt.show()

data=my_plasma.neutron_source_density
print(data)

Shimwell · April 1, 2024, 7:42pm

Currently the RZ location of the coordiates are randomly selected each time a plasma source is created.

So they will be slightly different each time. This can be reduced by sampling more that 1000 points. Or alternatively you can change the sampling method used one these two lines of code openmc-plasma-source/src/openmc_plasma_source/tokamak_source.py at 64e4aa13f1e81efb8d0e16d7b76acb40bacce468 · fusion-energy/openmc-plasma-source · GitHub

Im actually working on that package at the moment and trying to improve it. This can be one of the improvements to look into. I’m hoping to first improve the energy distribution then move it to using mesh sources (which will fix the reproducibility bug you noticed). Next openmc release supports mesh sources so hopefully a release of openmc-plasma-source will get it fixed just after the next openmc release.

Shimwell · April 8, 2024, 10:14am

I’ve just put in a pull request that should fix this issue. It would be great if you could let me know if this works for you

added grid for sampling a, alpha by shimwell · Pull Request #91 · fusion-energy/openmc-plasma-source (github.com)

Shailja · April 15, 2024, 6:51am

First I am sorry for delayed reply.
I tried same changes you made in added grid for sampling a, alpha by shimwell · Pull Request #91 · fusion-energy/openmc-plasma-source · GitHub in my tokamak_source.py file of package and then I ran my input file to print neutron density values. But again values seem changing.

Am I not doing it properly because I understood that you did some changes in sampling by adding 4 lines for a and alpha sampling.

Shimwell · April 15, 2024, 7:42am

The new version of openmc-plasma-source is now on pypi. So it should be possible to install with

pip install==0.3.1

Shailja · March 28, 2025, 5:05am

Hii

github.com/fusion-energy/openmc-plasma-source

src/openmc_plasma_source/tokamak_source.py

main

from typing import Tuple, Dict, List, Union

import numpy as np
from numpy.typing import NDArray
import openmc
from openmc import IndependentSource
import openmc.checkvalue as cv
from NeSST.spectral_model import reac_DD, reac_DT, reac_TT

from .fuel_types import get_neutron_energy_distribution, get_reactions_from_fuel


def tokamak_source(
    major_radius: float,
    minor_radius: float,
    elongation: float,
    triangularity: float,
    mode: str,
    ion_density_centre: float,
    ion_density_peaking_factor: float,

This file has been truncated. show original

This is changed from version which pip command makes download in the local system.
So when i tried to run examples section (using python)

github.com/fusion-energy/openmc-plasma-source

examples/tokamak_source_example.py

main

from pathlib import Path

import openmc

from openmc_plasma_source import tokamak_source

# just making use of a local cross section xml file, replace with your own cross sections or comment out
openmc.config["cross_sections"] = Path(__file__).parent.resolve() / "cross_sections.xml"


# minimal geometry
sphere_surface = openmc.Sphere(r=100000.0, boundary_type="vacuum")
cell = openmc.Cell(region=-sphere_surface)
geometry = openmc.Geometry([cell])

# create a plasma source
my_sources = tokamak_source(
    elongation=1.557,
    ion_density_centre=1.09e20,
    ion_density_pedestal=1.09e20,

This file has been truncated. show original

It is giving following error

$ python3 tokamak_source_example.py

Traceback (most recent call last):
** File “/home/shailja/openmc/openmc-plasma-source-main/examples/tokamak_source_example .py”, line 24, in **
** my_sources = tokamak_source(**
TypeError: ‘module’ object is not callable

$ python3 ring_source_example.py

Traceback (most recent call last):
File “/home/shailja/openmc/openmc-plasma-source-main/examples/ring_source_example.py”, line 6, in
from openmc_plasma_source import fusion_ring_source
ImportError: cannot import name ‘fusion_ring_source’ from ‘openmc_plasma_source’ (/home/shailja/.local/lib/python3.10/site-packages/openmc_plasma_source/init.py)

I observe that this is coming due to online installation of verion 3.1 of openmc_plasma_source which is not same vesion as given on platform openmc-plasma-source/src/openmc_plasma_source/tokamak_source.py at main · fusion-energy/openmc-plasma-source · GitHub

Can you plz make source files update so that pip can install the recent version.
I tried to install without pip also but to no avail.

regards
Shailja

Shimwell · March 31, 2025, 9:49pm

you could try installing directly from the repo main branch if you want the newest code that is not yet released on pypi

pip install git+https://github.com/fusion-energy/openmc-plasma-source

Let me know if this fixes the issue

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Different distribution of Neutron density in openmc_plasma_source

minimal geometry

create a plasma source

ion_temperature_centre=45.9,

ion_temperature_centre=4.71,

ion_temperature_pedestal=6.09,

ion_temperature_pedestal=0.427,

major_radius=9.06,

minor_radius=2.92258,

triangularity=0.270,

ion_temperature_beta=3.86,

Tell OpenMC we’re going to use our custom source

create a figure and axis

plot the tokamak source with a color scale

create a ScalarMappable object for the colorbar

add a colorbar

show the plot

show the plot

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