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Added full openmc.data package and hooks in code

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This commit is contained in:
Paul Romano 2016-05-13 15:57:39 -05:00
parent bc2fab296b
commit 68e5a093aa
158 changed files with 10186 additions and 4878 deletions
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8
examples/python/basic/build-xml.py
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@ -16,16 +16,16 @@ particles = 10000
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
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_s_alpha_beta('HH2O', '71t')
moderator.add_s_alpha_beta('c_H_in_H2O', '71t')
fuel = openmc.Material(material_id=40, name='fuel')
fuel.set_density('g/cc', 4.5)

10
examples/python/boxes/build-xml.py
View file

@ -16,10 +16,10 @@ particles = 10000
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
u238 = openmc.Nuclide('U-238')
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')
@ -34,7 +34,7 @@ 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_s_alpha_beta('HH2O', '71t')
moderator.add_s_alpha_beta('c_H_in_H2O', '71t')
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([fuel1, fuel2, moderator])

10
examples/python/lattice/hexagonal/build-xml.py
View file

@ -15,10 +15,10 @@ particles = 10000
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
fe56 = openmc.Nuclide('Fe-56')
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')
@ -29,7 +29,7 @@ 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_s_alpha_beta('HH2O', '71t')
moderator.add_s_alpha_beta('c_H_in_H2O', '71t')
iron = openmc.Material(material_id=3, name='iron')
iron.set_density('g/cc', 7.9)

8
examples/python/lattice/nested/build-xml.py
View file

@ -15,9 +15,9 @@ particles = 10000
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
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')
@ -28,7 +28,7 @@ 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_s_alpha_beta('HH2O', '71t')
moderator.add_s_alpha_beta('c_H_in_H2O', '71t')
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials((moderator, fuel))

8
examples/python/lattice/simple/build-xml.py
View file

@ -15,9 +15,9 @@ particles = 10000
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
o16 = openmc.Nuclide('O-16')
u235 = openmc.Nuclide('U-235')
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')
@ -28,7 +28,7 @@ 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_s_alpha_beta('HH2O', '71t')
moderator.add_s_alpha_beta('c_H_in_H2O', '71t')
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([moderator, fuel])

68
examples/python/pincell/build-xml.py
View file

@ -15,39 +15,39 @@ particles = 1000
###############################################################################
# Instantiate some Nuclides
h1 = openmc.Nuclide('H-1')
h2 = openmc.Nuclide('H-2')
he4 = openmc.Nuclide('He-4')
b10 = openmc.Nuclide('B-10')
b11 = openmc.Nuclide('B-11')
o16 = openmc.Nuclide('O-16')
o17 = openmc.Nuclide('O-17')
cr50 = openmc.Nuclide('Cr-50')
cr52 = openmc.Nuclide('Cr-52')
cr53 = openmc.Nuclide('Cr-53')
cr54 = openmc.Nuclide('Cr-54')
fe54 = openmc.Nuclide('Fe-54')
fe56 = openmc.Nuclide('Fe-56')
fe57 = openmc.Nuclide('Fe-57')
fe58 = openmc.Nuclide('Fe-58')
zr90 = openmc.Nuclide('Zr-90')
zr91 = openmc.Nuclide('Zr-91')
zr92 = openmc.Nuclide('Zr-92')
zr94 = openmc.Nuclide('Zr-94')
zr96 = openmc.Nuclide('Zr-96')
sn112 = openmc.Nuclide('Sn-112')
sn114 = openmc.Nuclide('Sn-114')
sn115 = openmc.Nuclide('Sn-115')
sn116 = openmc.Nuclide('Sn-116')
sn117 = openmc.Nuclide('Sn-117')
sn118 = openmc.Nuclide('Sn-118')
sn119 = openmc.Nuclide('Sn-119')
sn120 = openmc.Nuclide('Sn-120')
sn122 = openmc.Nuclide('Sn-122')
sn124 = openmc.Nuclide('Sn-124')
u234 = openmc.Nuclide('U-234')
u235 = openmc.Nuclide('U-235')
u238 = openmc.Nuclide('U-238')
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')
u234 = openmc.Nuclide('U234')
u235 = openmc.Nuclide('U235')
u238 = openmc.Nuclide('U238')
# Instantiate some Materials and register the appropriate Nuclides
uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
@ -98,7 +98,7 @@ 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_s_alpha_beta('HH2O', '71t')
borated_water.add_s_alpha_beta('c_H_in_H2O', '71t')
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uo2, helium, zircaloy, borated_water])

2
examples/python/reflective/build-xml.py
View file

@ -16,7 +16,7 @@ particles = 10000
###############################################################################
# Instantiate a Nuclides
u235 = openmc.Nuclide('U-235')
u235 = openmc.Nuclide('U235')
# Instantiate a Material and register the Nuclide
fuel = openmc.Material(material_id=1, name='fuel')