crp/OpenMC
1
0
Fork 0
mirror of https://github.com/openmc-dev/openmc.git synced 2026-07-27 21:55:41 -04:00
Code Issues Projects Releases Packages Wiki Activity Actions 5

Updating with develop

Browse source
This commit is contained in:
Adam Nelson 2016-10-04 19:55:52 -04:00
commit a52921e40a
50 changed files with 2799 additions and 2088 deletions
Download patch file Download diff file Expand all files Collapse all files

6
examples/python/basic/build-xml.py
View file

@ -100,9 +100,9 @@ settings_file.export_to_xml()
###############################################################################
# Instantiate some tally Filters
cell_filter = openmc.Filter(type='cell', bins=100)
energy_filter = openmc.Filter(type='energy', bins=[0., 20.])
energyout_filter = openmc.Filter(type='energyout', bins=[0., 20.])
cell_filter = openmc.CellFilter(100)
energy_filter = openmc.EnergyFilter([0., 20.])
energyout_filter = openmc.EnergyoutFilter([0., 20.])
# Instantiate the first Tally
first_tally = openmc.Tally(tally_id=1, name='first tally')

2
examples/python/lattice/hexagonal/build-xml.py
View file

@ -159,7 +159,7 @@ plot_file.export_to_xml()
# Instantiate a distribcell Tally
tally = openmc.Tally(tally_id=1)
tally.filters = [openmc.Filter(type='distribcell', bins=[cell2.id])]
tally.filters = [openmc.DistribcellFilter(cell2.id)]
tally.scores = ['total']
# Instantiate a Tallies collection and export to XML

3
examples/python/lattice/nested/build-xml.py
View file

@ -162,8 +162,7 @@ mesh.lower_left = [-2, -2]
mesh.width = [1, 1]
# Instantiate tally Filter
mesh_filter = openmc.Filter()
mesh_filter.mesh = mesh
mesh_filter = openmc.MeshFilter(mesh)
# Instantiate the Tally
tally = openmc.Tally(tally_id=1)

3
examples/python/lattice/simple/build-xml.py
View file

@ -155,8 +155,7 @@ mesh.lower_left = [-2, -2]
mesh.width = [1, 1]
# Instantiate tally Filter
mesh_filter = openmc.Filter()
mesh_filter.mesh = mesh
mesh_filter = openmc.MeshFilter(mesh)
# Instantiate tally Trigger
trigger = openmc.Trigger(trigger_type='rel_err', threshold=1E-2)

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

@ -185,9 +185,8 @@ mesh.lower_left = [-0.62992, -0.62992, -1.e50]
mesh.upper_right = [0.62992, 0.62992, 1.e50]
# Instantiate some tally Filters
energy_filter = openmc.Filter(type='energy', bins=[0., 4.e-6, 20.])
mesh_filter = openmc.Filter()
mesh_filter.mesh = mesh
energy_filter = openmc.EnergyFilter([0., 4.e-6, 20.])
mesh_filter = openmc.MeshFilter(mesh)
# Instantiate the Tally
tally = openmc.Tally(tally_id=1, name='tally 1')

6
examples/python/pincell_multigroup/build-xml.py
View file

@ -156,11 +156,9 @@ mesh.lower_left = [-0.63, -0.63, -1.e50]
mesh.upper_right = [0.63, 0.63, 1.e50]
# Instantiate some tally Filters
energy_filter = openmc.Filter(type='energy',
bins=[1E-11, 0.0635E-6, 10.0E-6, 1.0E-4, 1.0E-3,
energy_filter = openmc.EnergyFilter([1E-11, 0.0635E-6, 10.0E-6, 1.0E-4, 1.0E-3,
0.5, 1.0, 20.0])
mesh_filter = openmc.Filter()
mesh_filter.mesh = mesh
mesh_filter = openmc.MeshFilter(mesh)
# Instantiate the Tally
tally = openmc.Tally(tally_id=1, name='tally 1')