OpenMC data downloader
A Python package for downloading preprocessed cross section data in the h5 file
format for use with OpenMC.
This package allows you to download a fully reproducible composite nuclear data
library with one command.
There are several methods of obtaining complete data libraries for use with
OpenMC, for example:
History
The package was originally conceived by Jonathan Shimwell as a means of
downloading a minimal selection of data for use on continuous integration
platforms.
The package can also used to produce traceable and reproducible
nuclear data distributions.
System Installation
To install the openmc_data_downloader you need to have Python3 installed,
OpenMC is also advisable if you want to run simulations using the h5 data files
but not actually mandatory if you just want to download the cross sections.
pip install openmc_data_downloader
Features
The OpenMC data downloader is able to download cross section files for isotopes
from nuclear data libraries.The user specifies the nuclear data libraries in
order of their preference. When an isotope is found in multiple libraries it
will be downloaded from the highest preference library. This avoid duplication
of isotopes and provides a reproducible nuclear data environment.
The nuclear data h5 file are downloaded from the OpenMC-data-storage
repository. Prior to being added to the repository they have been automatically
processed using scripts from OpenMC data repository, these scripts convert ACE
and ENDF file to h5 files.
The resulting h5 files are then used in and automated test suite of simulations
to help ensure they are suitable for their intended purpose.
Isotopes for downloading can be found in a variety of ways as demonstrated below.
When downloading a cross_section.xml file is automatically created and h5 files
are named with their nuclear data library and the isotope. This helps avoid
downloading files that already exist locally and the overwrite
argument
can be used to control if these files are downloaded again.
Usage - command line usage
Getting a description of the input options
openmc_data_downloader --help
Downloading a single isotope from the FENDL 3.1d nuclear library
openmc_data_downloader -l FENDL-3.1d -i Li6
Downloading a multiple isotopes from the TENDL 2019 nuclear library
openmc_data_downloader -l TENDL-2019 -i Li6 Li7
Downloading a single element from the TENDL 2019 nuclear library
openmc_data_downloader -l TENDL-2019 -e Li
Downloading a multiple element from the TENDL 2019 nuclear library
openmc_data_downloader -l TENDL-2019 -e Li Si Na
Downloading h5 files from the ENDF/B 7.1 NNDC library to a specific folder (destination)
openmc_data_downloader -l ENDFB-7.1-NNDC -i Be9 -d my_h5_files
Downloading a combination of isotopes and element from the TENDL 2019 nuclear library
openmc_data_downloader -l TENDL-2019 -e Li Si Na -i Fe56 U235
Downloading all the isotopes from the TENDL 2019 nuclear library
openmc_data_downloader -l TENDL-2019 -i all
Downloading all the stable isotopes from the TENDL 2019 nuclear library
openmc_data_downloader -l TENDL-2019 -i stable
Downloading all the isotopes in a materials.xml file from the TENDL 2019 nuclear library
openmc_data_downloader -l TENDL-2019 -m materials.xml
Downloading 3 isotopes from ENDF/B 7.1 NNDC (first choice) and TENDL 2019 (second choice) nuclear library
openmc_data_downloader -l ENDFB-7.1-NNDC TENDL-2019 -i Li6 Li7 Be9
Downloading the photon only cross section for an element ENDF/B 7.1 NNDC
openmc_data_downloader -l ENDFB-7.1-NNDC -e Li -p photon
Downloading the neutron and photon cross section for an element ENDF/B 7.1 NNDC
openmc_data_downloader -l ENDFB-7.1-NNDC -e Li -p neutron photon
Downloading the neutron cross section for elements and an SaB cross sections
openmc_data_downloader -l ENDFB-7.1-NNDC -e Be O -s c_Be_in_BeO
Usage - within a Python environment
When using the Python API the just_in_time_library_generator()
function
provides similar capabilities to the openmc_data_downloader
terminal
command. With one key difference being that just_in_time_library_generator()
sets the OPENMC_CROSS_SECTIONS
environmental variable to point to the
newly created cross_sections.xml by default.
Downloading the isotopes present in an OpenMC material
import openmc
import openmc_data_downloader as odd
mat1 = openmc.Material()
mat1.add_element('Fe', 0.95)
mat1.add_element('C', 0.05)
odd.just_in_time_library_generator(
libraries='FENDL-3.1d',
materials=mat1
)
Downloading the isotopes present in an OpenMC material from two libraries but with a preference for ENDF/B 7.1 NNDC library over TENDL 2019
import openmc
import openmc_data_downloader as odd
mat1 = openmc.Material()
mat1.add_element('Fe', 0.95)
mat1.add_element('C', 0.05)
odd.just_in_time_library_generator(
libraries=['ENDFB-7.1-NNDC', 'TENDL-2019'],
materials=mat1
)
Downloading the isotopes in several OpenMC materials
import openmc
import openmc_data_downloader as odd
mat1 = openmc.Material()
mat1.add_element('Fe', 0.95)
mat1.add_element('C', 0.05)
mat2 = openmc.Material()
mat2.add_element('H', 0.66)
mat2.add_element('0', 0.33)
odd.just_in_time_library_generator(
libraries='ENDFB-7.1-NNDC',
materials=[mat1, mat2]
)
mats = openmc.Materials([mat1, mat2])
odd.just_in_time_library_generator(
libraries='ENDFB-7.1-NNDC',
materials=mats
)
Downloading neutron cross sections for a material with an SaB
import openmc
import openmc_data_downloader as odd
my_mat = openmc.Material()
my_mat.add_element('Be', 0.5)
my_mat.add_element('O', 0.5)
my_mat.add_s_alpha_beta('Be_in_BeO')
odd.just_in_time_library_generator(
libraries='ENDFB-7.1-NNDC',
materials= my_mat
particles = ['neutron'],
)
Downloading photon and neutron cross sections for isotopes and elements from the TENDL 2019 library
import openmc
import openmc_data_downloader as odd
odd.just_in_time_library_generator(
libraries='TENDL-2019',
elements=['Li', 'Be'],
particles = ['photon', 'neutron'],
isotopes=['Fe56', 'U235'],
)