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Description
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.
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