crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-28 14:15:42 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

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

Browse source
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
Download patch file Download diff file Expand all files Collapse all files

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