crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-29 06:35:48 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

Make Materials, Plots, and Tallies list-like

Browse source
This commit is contained in:
Paul Romano 2016-04-29 16:14:02 -05:00
parent ae083cf5d4
commit d9b097dbaf
31 changed files with 373 additions and 297 deletions
Download patch file Download diff file Expand all files Collapse all files

16
examples/python/pincell_multigroup/build-xml.py
View file

@ -81,10 +81,9 @@ water = openmc.Material(material_id=2, name='Water')
water.set_density('macro', 1.0)
water.add_macroscopic(h2o_data)
# Instantiate a Materials collection, register all Materials, and export to XML
materials_file = openmc.Materials()
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([uo2, water])
materials_file.default_xs = '300K'
materials_file.add_materials([uo2, water])
materials_file.export_to_xml()
@ -167,14 +166,9 @@ mesh_filter.mesh = mesh
# Instantiate the Tally
tally = openmc.Tally(tally_id=1, name='tally 1')
tally.add_filter(energy_filter)
tally.add_filter(mesh_filter)
tally.add_score('flux')
tally.add_score('fission')
tally.add_score('nu-fission')
tally.filters = [energy_filter, mesh_filter]
tally.scores = ['flux', 'fission', 'nu-fission']
# Instantiate a Tallies collection, register all Tallies, and export to XML
tallies_file = openmc.Tallies()
tallies_file.add_mesh(mesh)
tallies_file.add_tally(tally)
tallies_file = openmc.Tallies([tally])
tallies_file.export_to_xml()