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Updating openmc.Materials object used in _generate_materials_xml.

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This commit is contained in:
Patrick Shriwise 2021-10-11 22:04:05 -05:00
parent f1234a38b0
commit e6cd4449e1
1 changed files with 2 additions and 4 deletions
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6
openmc/deplete/operator.py
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@ -618,8 +618,6 @@ class Operator(TransportOperator):
through direct memory writing.
"""
materials = openmc.Materials(self.materials)
# Sort nuclides according to order in AtomNumber object
nuclides = list(self.number.nuclides)
for mat in self.materials:
@ -631,9 +629,9 @@ class Operator(TransportOperator):
tree = ET.parse(str(mfile))
xs = tree.find('cross_sections')
if xs is not None:
materials.cross_sections = xs.text
self.materials.cross_sections = xs.text
materials.export_to_xml()
self.materials.export_to_xml()
def _get_tally_nuclides(self):
"""Determine nuclides that should be tallied for reaction rates.