openmc-data-storage / openmc_data_downloader Goto Github PK

When using the just in time library generator the and with WMP as the preferred choice the downloader should also download additional libraries for the unresolved resonance region as explained by Paul in this issue https://openmc.discourse.group/t/minimal-wmp-simulation-with-minimal-cross-section-xml/1100/9

The utils python file has several functions but is missing docstrings, these are on the todo list

Neutrons for the JEFF 3.2 distribution have been processed in this repo https://github.com/openmc-data-storage/JEFF-3.2

However they need adding to the cross_sections_directory.py file like the other h5 file available.

This can be done with a large pull request or several smaller ones to the develop branch:

• the introduction of a new variable called jeff_32_neutron_isotopes in the cross_sections_directory.py file . The varible should contains all the isotope names in this folder

• then a small section that builds up the dictionary of information on the cross section like I've done for other neutron libraries for example

• then the add the new collection to the existing collections on this line

• then change the README.md file so that it has a few JEFF-3.2 examples

• Add some tests to here and here

As discussed briefly in #14 it would be great to have some of the h5 files that are currently found on openmc.org

To do this then there are a few stages, another base repository to template that would take a different approach to the existing template repository.

The new template would download from https://openmc.org/official-data-libraries/ then extract the contents and self commit to the repo using a GitHub action.

Currently the just_in_time_library_generator accepts openmc.Material objects

It would be great if these were acceptabed as well

• openmc.Materials
• neutronics_material_maker.Material()
• neutronics_material_maker.MultMaterial()

error caused when expanding a element without a cross section file set

import openmc
import openmc_data_downloader

openmc.config['cross_sections']=''

mat1 = openmc.Material()
mat1.add_element('Li', 1)
mat1.set_density('g/cm3',1)
materials=openmc.Materials([mat1])
materials.download_cross_section_data(
        libraries=["FENDL-3.1d"],
        set_OPENMC_CROSS_SECTIONS=True,
        particles=["neutron"],
    )

surf1 = openmc.Sphere(r=10, boundary_type='vacuum')
region1 = -surf1
cell1=openmc.Cell(region=region1,fill=mat1)
geometry=openmc.Geometry([cell1])

point = openmc.stats.Point((0, 0, 0))
energy = openmc.stats.Discrete([14.1e6], [1.0])
source = openmc.Source(space=point, energy=energy)

settings = openmc.Settings()
settings.source = source
settings.run_mode = 'fixed source'
settings.particles = 10000
settings.batches = 10

model = openmc.Model(geometry, materials, settings)
model.run()

ile ~/venv/openmc_env/lib/python3.8/site-packages/openmc/material.py:744, in Material.add_element(self, element, percent, percent_type, enrichment, enrichment_target, enrichment_type)
    742 # Add naturally-occuring isotopes
    743 element = openmc.Element(element)
--> 744 for nuclide in element.expand(percent,
    745                               percent_type,
    746                               enrichment,
    747                               enrichment_target,
    748                               enrichment_type):
    749     self.add_nuclide(*nuclide)

File ~/venv/openmc_env/lib/python3.8/site-packages/openmc/element.py:138, in Element.expand(self, percent, percent_type, enrichment, enrichment_target, enrichment_type, cross_sections)
    136 if cross_sections is not None:
    137     library_nuclides = set()
--> 138     tree = ET.parse(cross_sections)
    139     root = tree.getroot()
    140     for child in root.findall('library'):

File /usr/lib/python3.8/xml/etree/ElementTree.py:1202, in parse(source, parser)
   1193 """Parse XML document into element tree.

@shimwell It's really amazing I ❤️ this.

One question, How to install thermal scattering data?