About how to use nuclear data library downloaded from the official website

User Support
1

I would like to ask how to import the nuclear data library downloaded from the official website into OpenMC. I have downloaded three databases from the official website, but I don’t know how to import them into OpenMC on my Linux system. Any reply would be greatly appreciated.

2

I am a newcomer to OpenMC and have some questions about how to start and run OpenMC. I would appreciate any help.

3

Hi @Utopia and welcome to the community. You can find usage instructions here:

For a given library, download and unpack the archive in a desired location. Inside the directory that is extracted is a file called cross_sections.xml . Set your OPENMC_CROSS_SECTIONS environment variable to the absolute path of this file and then it will be recognized when you run OpenMC.

4

Thank you for your patient reply,I wonder what operating system (Windows or Linux) will the environment be configured on?

5

Additionally, I think I need a simple example to verify that my OpenMC is running properly.

6

Hi @Utopia there are some examples you will find in your open clone ( I am guessing you installed it from the source code). Please go to your openmc directory and open it in the terminal. Then
run this commands:

 cd examples/jezebel
 python jezebel.py

if this works then OK other wise you can just copy this input file and run it as a python program.

import openmc

# Create plutonium metal material
pu = openmc.Material()
pu.set_density('sum')
pu.add_nuclide('Pu239', 3.7047e-02)
pu.add_nuclide('Pu240', 1.7512e-03)
pu.add_nuclide('Pu241', 1.1674e-04)
pu.add_element('Ga', 1.3752e-03)
mats = openmc.Materials([pu])
mats.export_to_xml()

# Create a single cell filled with the Pu metal
sphere = openmc.Sphere(r=6.3849, boundary_type='vacuum')
cell = openmc.Cell(fill=pu, region=-sphere)
geom = openmc.Geometry([cell])
geom.export_to_xml()

# Finally, define some run settings
settings = openmc.Settings()
settings.batches = 200
settings.inactive = 10
settings.particles = 10000
settings.export_to_xml()

# Run the simulation
openmc.run()

# Get the resulting k-effective value
n = settings.batches
with openmc.StatePoint(f'statepoint.{n}.h5') as sp:
    keff = sp.keff
    print(f'Final k-effective = {keff}')

also make sure that your OPENMC_CROSS_SECTIONS is set. To check try to run this command in the terminal,

echo $OPENMC_CROSS_SECTIONS

and it should print out something like this: /home/ebny_walid/OpenMC/openmc-setup/endfb-viii.0-hdf5/cross_sections.xml

if you get a blank then copy the path of the cross_sections.xml file [ for me it was /home/ebny_walid/OpenMC/openmc-setup/endfb-viii.0-hdf5/cross_sections.xml this] and try to set

export OPENMC_CROSS_SECTIONS ="your_copied_path"

to get more examples you can visit this repository: GitHub - mit-crpg/openmc-reactor-examples: Catalogs a plethora of reactor inputs for OpenMC

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