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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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11
examples/python/basic/build-xml.py
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@ -15,21 +15,16 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
# Instantiate some Materials and register the appropriate Nuclides
moderator = openmc.Material(material_id=41, 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')
fuel = openmc.Material(material_id=40, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel])

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

14
examples/python/lattice/hexagonal/build-xml.py
View file

@ -14,26 +14,20 @@ particles = 10000
# Exporting to OpenMC materials.xml File
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
fe56 = openmc.Nuclide('Fe56')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
moderator = openmc.Material(material_id=2, 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')
iron = openmc.Material(material_id=3, name='iron')
iron.set_density('g/cc', 7.9)
iron.add_nuclide(fe56, 1.)
iron.add_element('Fe', 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel, iron])

11
examples/python/lattice/nested/build-xml.py
View file

@ -14,20 +14,15 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
moderator = openmc.Material(material_id=2, 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

11
examples/python/lattice/simple/build-xml.py
View file

@ -14,20 +14,15 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
o16 = openmc.Nuclide('O16')
u235 = openmc.Nuclide('U235')
# Instantiate some Materials and register the appropriate Nuclides
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
moderator = openmc.Material(material_id=2, 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

79
examples/python/pincell/build-xml.py
View file

@ -14,86 +14,29 @@ particles = 1000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H1')
h2 = openmc.Nuclide('H2')
he4 = openmc.Nuclide('He4')
b10 = openmc.Nuclide('B10')
b11 = openmc.Nuclide('B11')
o16 = openmc.Nuclide('O16')
o17 = openmc.Nuclide('O17')
cr50 = openmc.Nuclide('Cr50')
cr52 = openmc.Nuclide('Cr52')
cr53 = openmc.Nuclide('Cr53')
cr54 = openmc.Nuclide('Cr54')
fe54 = openmc.Nuclide('Fe54')
fe56 = openmc.Nuclide('Fe56')
fe57 = openmc.Nuclide('Fe57')
fe58 = openmc.Nuclide('Fe58')
zr90 = openmc.Nuclide('Zr90')
zr91 = openmc.Nuclide('Zr91')
zr92 = openmc.Nuclide('Zr92')
zr94 = openmc.Nuclide('Zr94')
zr96 = openmc.Nuclide('Zr96')
sn112 = openmc.Nuclide('Sn112')
sn114 = openmc.Nuclide('Sn114')
sn115 = openmc.Nuclide('Sn115')
sn116 = openmc.Nuclide('Sn116')
sn117 = openmc.Nuclide('Sn117')
sn118 = openmc.Nuclide('Sn118')
sn119 = openmc.Nuclide('Sn119')
sn120 = openmc.Nuclide('Sn120')
sn122 = openmc.Nuclide('Sn122')
sn124 = openmc.Nuclide('Sn124')
u = openmc.Element('U')
o = openmc.Element('O')
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
uo2.set_density('g/cm3', 10.29769)
uo2.add_element(u, 1., enrichment=2.4)
uo2.add_element(o, 2.)
uo2.add_element('U', 1., enrichment=2.4)
uo2.add_element('O', 2.)
helium = openmc.Material(material_id=2, name='Helium for gap')
helium.set_density('g/cm3', 0.001598)
helium.add_nuclide(he4, 2.4044e-4)
helium.add_element('He', 2.4044e-4)
zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
zircaloy.set_density('g/cm3', 6.55)
zircaloy.add_nuclide(o16, 3.0743e-4)
zircaloy.add_nuclide(o17, 7.4887e-7)
zircaloy.add_nuclide(cr50, 3.2962e-6)
zircaloy.add_nuclide(cr52, 6.3564e-5)
zircaloy.add_nuclide(cr53, 7.2076e-6)
zircaloy.add_nuclide(cr54, 1.7941e-6)
zircaloy.add_nuclide(fe54, 8.6699e-6)
zircaloy.add_nuclide(fe56, 1.3610e-4)
zircaloy.add_nuclide(fe57, 3.1431e-6)
zircaloy.add_nuclide(fe58, 4.1829e-7)
zircaloy.add_nuclide(zr90, 2.1827e-2)
zircaloy.add_nuclide(zr91, 4.7600e-3)
zircaloy.add_nuclide(zr92, 7.2758e-3)
zircaloy.add_nuclide(zr94, 7.3734e-3)
zircaloy.add_nuclide(zr96, 1.1879e-3)
zircaloy.add_nuclide(sn112, 4.6735e-6)
zircaloy.add_nuclide(sn114, 3.1799e-6)
zircaloy.add_nuclide(sn115, 1.6381e-6)
zircaloy.add_nuclide(sn116, 7.0055e-5)
zircaloy.add_nuclide(sn117, 3.7003e-5)
zircaloy.add_nuclide(sn118, 1.1669e-4)
zircaloy.add_nuclide(sn119, 4.1387e-5)
zircaloy.add_nuclide(sn120, 1.5697e-4)
zircaloy.add_nuclide(sn122, 2.2308e-5)
zircaloy.add_nuclide(sn124, 2.7897e-5)
zircaloy.add_element('Sn', 0.014 , 'wo')
zircaloy.add_element('Fe', 0.00165, 'wo')
zircaloy.add_element('Cr', 0.001 , 'wo')
zircaloy.add_element('Zr', 0.98335, 'wo')
borated_water = openmc.Material(material_id=4, name='Borated water at 975 ppm')
borated_water = openmc.Material(material_id=4, name='Borated water')
borated_water.set_density('g/cm3', 0.740582)
borated_water.add_nuclide(b10, 8.0042e-6)
borated_water.add_nuclide(b11, 3.2218e-5)
borated_water.add_nuclide(h1, 4.9457e-2)
borated_water.add_nuclide(h2, 7.4196e-6)
borated_water.add_nuclide(o16, 2.4672e-2)
borated_water.add_nuclide(o17, 6.0099e-5)
borated_water.add_element('B', 4.0e-5)
borated_water.add_element('H', 5.0e-2)
borated_water.add_element('O', 2.4e-2)
borated_water.add_s_alpha_beta('c_H_in_H2O')
# Instantiate a Materials collection and export to XML

5
examples/python/reflective/build-xml.py
View file

@ -15,13 +15,10 @@ particles = 10000
# Exporting to OpenMC materials.xml file
###############################################################################
# Instantiate a Nuclides
u235 = openmc.Nuclide('U235')
# Instantiate a Material and register the Nuclide
fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
fuel.add_nuclide('U235', 1.)
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([fuel])