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updated tests and updated setting of nuclides and elements in example files

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This commit is contained in:
samuelshaner 2016-12-15 21:06:12 +00:00
parent ff3d034275
commit 1a7630083b
18 changed files with 114 additions and 201 deletions
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14
examples/python/boxes/build-xml.py
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@ -15,25 +15,19 @@ particles = 10000
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
u238 = openmc.Nuclide('U238')
# Instantiate some Materials and register the appropriate Nuclides
fuel1 = openmc.Material(material_id=1, name='fuel')
fuel1.set_density('g/cc', 4.5)
fuel1.add_nuclide(u235, 1.)
fuel1.add_nuclide('U235', 1.)
fuel2 = openmc.Material(material_id=2, name='depleted fuel')
fuel2.set_density('g/cc', 4.5)
fuel2.add_nuclide(u238, 1.)
fuel2.add_nuclide('U238', 1.)
moderator = openmc.Material(material_id=3, name='moderator')
moderator.set_density('g/cc', 1.0)
moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_element('H', 2.)
moderator.add_element('O', 1.)
moderator.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML