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Merge pull request #637 from paulromano/list-like-classes

Browse source 
Make Materials, Plots, and Tallies list-like
This commit is contained in:
Will Boyd 2016-04-30 10:21:45 -06:00
commit cac928cb45
31 changed files with 400 additions and 334 deletions
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16
docs/source/pythonapi/examples/mgxs-part-i.ipynb
View file

@ -212,10 +212,9 @@
},
"outputs": [],
"source": [
"# Instantiate a Materials object, register all Materials, and export to XML\n",
"materials_file = openmc.Materials()\n",
"# Instantiate a Materials collection and export to XML\n",
"materials_file = openmc.Materials([inf_medium])\n",
"materials_file.default_xs = '71c'\n",
"materials_file.add_material(inf_medium)\n",
"materials_file.export_to_xml()"
]
},
@ -466,17 +465,14 @@
"tallies_file = openmc.Tallies()\n",
"\n",
"# Add total tallies to the tallies file\n",
"for tally in total.tallies.values():\n",
" tallies_file.add_tally(tally)\n",
"tallies_file += total.tallies.values()\n",
"\n",
"# Add absorption tallies to the tallies file\n",
"for tally in absorption.tallies.values():\n",
" tallies_file.add_tally(tally)\n",
"tallies_file += absorption.tallies.values()\n",
"\n",
"# Add scattering tallies to the tallies file\n",
"for tally in scattering.tallies.values():\n",
" tallies_file.add_tally(tally)\n",
" \n",
"tallies_file += scattering.tallies.values()\n",
"\n",
"# Export to \"tallies.xml\"\n",
"tallies_file.export_to_xml()"
]

9
docs/source/pythonapi/examples/mgxs-part-ii.ipynb
View file

@ -133,11 +133,8 @@
},
"outputs": [],
"source": [
"# Instantiate a Materials object, add Materials\n",
"materials_file = openmc.Materials()\n",
"materials_file.add_material(fuel)\n",
"materials_file.add_material(water)\n",
"materials_file.add_material(zircaloy)\n",
"# Instantiate a Materials collection\n",
"materials_file = openmc.Materials((fuel, water, zircaloy))\n",
"materials_file.default_xs = '71c'\n",
"\n",
"# Export to \"materials.xml\"\n",
@ -408,7 +405,7 @@
" \n",
" # Add OpenMC tallies to the tallies file for XML generation\n",
" for tally in xs_library[cell.id][rxn_type].tallies.values():\n",
" tallies_file.add_tally(tally, merge=True)\n",
" tallies_file.append(tally, merge=True)\n",
"\n",
"# Export to \"tallies.xml\"\n",
"tallies_file.export_to_xml()"

15
docs/source/pythonapi/examples/mgxs-part-iii.ipynb
View file

@ -133,11 +133,8 @@
},
"outputs": [],
"source": [
"# Instantiate a Materials object, add Materials\n",
"materials_file = openmc.Materials()\n",
"materials_file.add_material(fuel)\n",
"materials_file.add_material(water)\n",
"materials_file.add_material(zircaloy)\n",
"# Instantiate a Materials object\n",
"materials_file = openmc.Materials((fuel, water, zircaloy))\n",
"materials_file.default_xs = '71c'\n",
"\n",
"# Export to \"materials.xml\"\n",
@ -419,8 +416,7 @@
"plot.color = 'mat'\n",
"\n",
"# Instantiate a Plots object, add Plot, and export to \"plots.xml\"\n",
"plot_file = openmc.Plots()\n",
"plot_file.add_plot(plot)\n",
"plot_file = openmc.Plots([plot])\n",
"plot_file.export_to_xml()"
]
},
@ -685,9 +681,8 @@
"tally.filters = [mesh_filter]\n",
"tally.scores = ['fission', 'nu-fission']\n",
"\n",
"# Add mesh and tally to Tallies\n",
"tallies_file.add_mesh(mesh)\n",
"tallies_file.add_tally(tally)"
"# Add tally to collection\n",
"tallies_file.append(tally)"
]
},
{

19
docs/source/pythonapi/examples/pandas-dataframes.ipynb
View file

@ -108,11 +108,8 @@
},
"outputs": [],
"source": [
"# Instantiate a Materials object, add Materials\n",
"materials_file = openmc.Materials()\n",
"materials_file.add_material(fuel)\n",
"materials_file.add_material(water)\n",
"materials_file.add_material(zircaloy)\n",
"# Instantiate a Materials collection\n",
"materials_file = openmc.Materials((fuel, water, zircaloy))\n",
"materials_file.default_xs = '71c'\n",
"\n",
"# Export to \"materials.xml\"\n",
@ -329,9 +326,8 @@
"plot.pixels = [250, 250]\n",
"plot.color = 'mat'\n",
"\n",
"# Instantiate a Plots object, add Plot, and export to \"plots.xml\"\n",
"plot_file = openmc.Plots()\n",
"plot_file.add_plot(plot)\n",
"# Instantiate a Plots collection and export to \"plots.xml\"\n",
"plot_file = openmc.Plots([plot])\n",
"plot_file.export_to_xml()"
]
},
@ -449,8 +445,7 @@
"tally.scores = ['fission', 'nu-fission']\n",
"\n",
"# Add mesh and Tally to Tallies\n",
"tallies_file.add_mesh(mesh)\n",
"tallies_file.add_tally(tally)"
"tallies_file.append(tally)"
]
},
{
@ -478,7 +473,7 @@
"tally.nuclides = [u235, u238]\n",
"\n",
"# Add mesh and tally to Tallies\n",
"tallies_file.add_tally(tally)"
"tallies_file.append(tally)"
]
},
{
@ -510,7 +505,7 @@
"tally.triggers = [trigger]\n",
"\n",
"# Add mesh and tally to Tallies\n",
"tallies_file.add_tally(tally)"
"tallies_file.append(tally)"
]
},
{

17
docs/source/pythonapi/examples/post-processing.ipynb
View file

@ -104,11 +104,8 @@
},
"outputs": [],
"source": [
"# Instantiate a Materials object, add Materials\n",
"materials_file = openmc.Materials()\n",
"materials_file.add_material(fuel)\n",
"materials_file.add_material(water)\n",
"materials_file.add_material(zircaloy)\n",
"# Instantiate a Materials collection\n",
"materials_file = openmc.Materials((fuel, water, zircaloy))\n",
"materials_file.default_xs = '71c'\n",
"\n",
"# Export to \"materials.xml\"\n",
@ -298,9 +295,8 @@
"plot.pixels = [250, 250]\n",
"plot.color = 'mat'\n",
"\n",
"# Instantiate a Plots object, add Plot, and export to \"plots.xml\"\n",
"plot_file = openmc.Plots()\n",
"plot_file.add_plot(plot)\n",
"# Instantiate a Plots collection and export to \"plots.xml\"\n",
"plot_file = openmc.Plots([plot])\n",
"plot_file.export_to_xml()"
]
},
@ -393,17 +389,16 @@
"mesh.dimension = [100, 100]\n",
"mesh.lower_left = [-0.63, -0.63]\n",
"mesh.upper_right = [0.63, 0.63]\n",
"tallies_file.add_mesh(mesh)\n",
"\n",
"# Create mesh filter for tally\n",
"mesh_filter = openmc.Filter(type='mesh', bins=[1])\n",
"mesh_filter = openmc.Filter(type='mesh')\n",
"mesh_filter.mesh = mesh\n",
"\n",
"# Create mesh tally to score flux and fission rate\n",
"tally = openmc.Tally(name='flux')\n",
"tally.filters = [mesh_filter]\n",
"tally.scores = ['flux', 'fission']\n",
"tallies_file.add_tally(tally)"
"tallies_file.append(tally)"
]
},
{

93
docs/source/pythonapi/examples/tally-arithmetic.ipynb
View file

@ -15,16 +15,7 @@
"metadata": {
"collapsed": false
},
"outputs": [
{
"name": "stdout",
"output_type": "stream",
"text": [
"The autoreload extension is already loaded. To reload it, use:\n",
" %reload_ext autoreload\n"
]
}
],
"outputs": [],
"source": [
"%load_ext autoreload\n",
"%autoreload 2"
@ -126,11 +117,8 @@
},
"outputs": [],
"source": [
"# Instantiate a Materials object, add Materials\n",
"materials_file = openmc.Materials()\n",
"materials_file.add_material(fuel)\n",
"materials_file.add_material(water)\n",
"materials_file.add_material(zircaloy)\n",
"# Instantiate a Materials collection\n",
"materials_file = openmc.Materials((fuel, water, zircaloy))\n",
"materials_file.default_xs = '71c'\n",
"\n",
"# Export to \"materials.xml\"\n",
@ -321,9 +309,8 @@
"plot.pixels = [250, 250]\n",
"plot.color = 'mat'\n",
"\n",
"# Instantiate a Plots object, add Plot, and export to \"plots.xml\"\n",
"plot_file = openmc.Plots()\n",
"plot_file.add_plot(plot)\n",
"# Instantiate a Plots collection and export to \"plots.xml\"\n",
"plot_file = openmc.Plots([plot])\n",
"plot_file.export_to_xml()"
]
},
@ -366,7 +353,7 @@
"outputs": [
{
"data": {
"image/png": "iVBORw0KGgoAAAANSUhEUgAAAPoAAAD6AgMAAAD1grKuAAAABGdBTUEAALGPC/xhBQAAAAFzUkdC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\nVDExOjM5OjE0LTA0OjAwemPqXwAAAABJRU5ErkJggg==\n",
"image/png": "iVBORw0KGgoAAAANSUhEUgAAAPoAAAD6AgMAAAD1grKuAAAABGdBTUEAALGPC/xhBQAAACBjSFJN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\nfTFU5QAAAABJRU5ErkJggg==\n",
"text/plain": [
"<IPython.core.display.Image object>"
]
@ -421,7 +408,7 @@
"tally.filters = [openmc.Filter(type='cell', bins=[fuel_cell.id, moderator_cell.id])]\n",
"tally.filters.append(energy_filter)\n",
"tally.scores = ['flux']\n",
"tallies_file.add_tally(tally)\n",
"tallies_file.append(tally)\n",
"\n",
"# Instantiate reaction rate Tally in fuel\n",
"tally = openmc.Tally(name='fuel rxn rates')\n",
@ -429,7 +416,7 @@
"tally.filters.append(energy_filter)\n",
"tally.scores = ['nu-fission', 'scatter']\n",
"tally.nuclides = [u238, u235]\n",
"tallies_file.add_tally(tally)\n",
"tallies_file.append(tally)\n",
"\n",
"# Instantiate reaction rate Tally in moderator\n",
"tally = openmc.Tally(name='moderator rxn rates')\n",
@ -437,7 +424,7 @@
"tally.filters.append(energy_filter)\n",
"tally.scores = ['absorption', 'total']\n",
"tally.nuclides = [o16, h1]\n",
"tallies_file.add_tally(tally)"
"tallies_file.append(tally)"
]
},
{
@ -453,8 +440,7 @@
"abs_rate = openmc.Tally(name='abs. rate')\n",
"fiss_rate.scores = ['nu-fission']\n",
"abs_rate.scores = ['absorption']\n",
"tallies_file.add_tally(fiss_rate)\n",
"tallies_file.add_tally(abs_rate)"
"tallies_file += (fiss_rate, abs_rate)"
]
},
{
@ -469,7 +455,7 @@
"therm_abs_rate = openmc.Tally(name='therm. abs. rate')\n",
"therm_abs_rate.scores = ['absorption']\n",
"therm_abs_rate.filters = [openmc.Filter(type='energy', bins=[0., 0.625e-6])]\n",
"tallies_file.add_tally(therm_abs_rate)"
"tallies_file.append(therm_abs_rate)"
]
},
{
@ -485,7 +471,7 @@
"fuel_therm_abs_rate.scores = ['absorption']\n",
"fuel_therm_abs_rate.filters = [openmc.Filter(type='energy', bins=[0., 0.625e-6]),\n",
" openmc.Filter(type='cell', bins=[fuel_cell.id])]\n",
"tallies_file.add_tally(fuel_therm_abs_rate)"
"tallies_file.append(fuel_therm_abs_rate)"
]
},
{
@ -500,7 +486,7 @@
"therm_fiss_rate = openmc.Tally(name='therm. fiss. rate')\n",
"therm_fiss_rate.scores = ['nu-fission']\n",
"therm_fiss_rate.filters = [openmc.Filter(type='energy', bins=[0., 0.625e-6])]\n",
"tallies_file.add_tally(therm_fiss_rate)"
"tallies_file.append(therm_fiss_rate)"
]
},
{
@ -520,7 +506,7 @@
"tally.filters.append(energy_filter)\n",
"tally.scores = ['nu-fission', 'scatter']\n",
"tally.nuclides = [h1, u238]\n",
"tallies_file.add_tally(tally)"
"tallies_file.append(tally)"
]
},
{
@ -567,12 +553,11 @@
" 888\n",
" 888\n",
"\n",
" Copyright: 2011-2015 Massachusetts Institute of Technology\n",
" License: http://mit-crpg.github.io/openmc/license.html\n",
" Copyright: 2011-2016 Massachusetts Institute of Technology\n",
" License: http://openmc.readthedocs.org/en/latest/license.html\n",
" Version: 0.7.1\n",
" Git SHA1: eeb5091ca3a34cc85df73a3318cae2b6c7097413\n",
" Date/Time: 2016-04-13 11:39:14\n",
" MPI Processes: 1\n",
" Git SHA1: ae083cf5d491e6a778d5b762dad19c8d5fe45238\n",
" Date/Time: 2016-04-30 06:37:41\n",
"\n",
" ===========================================================================\n",
" ========================> INITIALIZATION <=========================\n",
@ -628,20 +613,20 @@
"\n",
" =======================> TIMING STATISTICS <=======================\n",
"\n",
" Total time for initialization = 4.0300E-01 seconds\n",
" Reading cross sections = 8.6000E-02 seconds\n",
" Total time in simulation = 1.4439E+01 seconds\n",
" Time in transport only = 1.4430E+01 seconds\n",
" Time in inactive batches = 2.2790E+00 seconds\n",
" Time in active batches = 1.2160E+01 seconds\n",
" Time synchronizing fission bank = 2.0000E-03 seconds\n",
" Total time for initialization = 7.0900E-01 seconds\n",
" Reading cross sections = 4.0400E-01 seconds\n",
" Total time in simulation = 1.7108E+01 seconds\n",
" Time in transport only = 1.7093E+01 seconds\n",
" Time in inactive batches = 3.3970E+00 seconds\n",
" Time in active batches = 1.3711E+01 seconds\n",
" Time synchronizing fission bank = 1.0000E-03 seconds\n",
" Sampling source sites = 1.0000E-03 seconds\n",
" SEND/RECV source sites = 1.0000E-03 seconds\n",
" SEND/RECV source sites = 0.0000E+00 seconds\n",
" Time accumulating tallies = 0.0000E+00 seconds\n",
" Total time for finalization = 1.0000E-03 seconds\n",
" Total time elapsed = 1.4856E+01 seconds\n",
" Calculation Rate (inactive) = 5484.86 neutrons/second\n",
" Calculation Rate (active) = 3083.88 neutrons/second\n",
" Total time elapsed = 1.7835E+01 seconds\n",
" Calculation Rate (inactive) = 3679.72 neutrons/second\n",
" Calculation Rate (active) = 2735.03 neutrons/second\n",
"\n",
" ============================> RESULTS <============================\n",
"\n",
@ -815,7 +800,7 @@
" <tbody>\n",
" <tr>\n",
" <th>0</th>\n",
" <td>0</td>\n",
" <td>0.0</td>\n",
" <td>6.250000e-07</td>\n",
" <td>total</td>\n",
" <td>absorption</td>\n",
@ -877,7 +862,7 @@
" <tbody>\n",
" <tr>\n",
" <th>0</th>\n",
" <td>0</td>\n",
" <td>0.0</td>\n",
" <td>6.250000e-07</td>\n",
" <td>total</td>\n",
" <td>nu-fission</td>\n",
@ -941,7 +926,7 @@
" <tbody>\n",
" <tr>\n",
" <th>0</th>\n",
" <td>0</td>\n",
" <td>0.0</td>\n",
" <td>6.250000e-07</td>\n",
" <td>10000</td>\n",
" <td>total</td>\n",
@ -1007,7 +992,7 @@
" <tbody>\n",
" <tr>\n",
" <th>0</th>\n",
" <td>0</td>\n",
" <td>0.0</td>\n",
" <td>6.250000e-07</td>\n",
" <td>10000</td>\n",
" <td>total</td>\n",
@ -1072,7 +1057,7 @@
" <tbody>\n",
" <tr>\n",
" <th>0</th>\n",
" <td>0</td>\n",
" <td>0.0</td>\n",
" <td>6.250000e-07</td>\n",
" <td>10000</td>\n",
" <td>total</td>\n",
@ -1615,21 +1600,21 @@
],
"metadata": {
"kernelspec": {
"display_name": "Python 2",
"display_name": "Python 3",
"language": "python",
"name": "python2"
"name": "python3"
},
"language_info": {
"codemirror_mode": {
"name": "ipython",
"version": 2
"version": 3
},
"file_extension": ".py",
"mimetype": "text/x-python",
"name": "python",
"nbconvert_exporter": "python",
"pygments_lexer": "ipython2",
"version": "2.7.6"
"pygments_lexer": "ipython3",
"version": "3.5.1"
}
},
"nbformat": 4,

12
examples/python/basic/build-xml.py
View file

@ -31,10 +31,9 @@ fuel = openmc.Material(material_id=40, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
# 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([moderator, fuel])
materials_file.default_xs = '71c'
materials_file.add_materials([moderator, fuel])
materials_file.export_to_xml()
@ -124,9 +123,6 @@ third_tally = openmc.Tally(tally_id=3, name='third tally')
third_tally.filters = [cell_filter, energy_filter, energyout_filter]
third_tally.scores = ['scatter', 'nu-scatter', 'nu-fission']
# Instantiate a Tallies collection, register all Tallies, and export to XML
tallies_file = openmc.Tallies()
tallies_file.add_tally(first_tally)
tallies_file.add_tally(second_tally)
tallies_file.add_tally(third_tally)
# Instantiate a Tallies collection and export to XML
tallies_file = openmc.Tallies((first_tally, second_tally, third_tally))
tallies_file.export_to_xml()

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

@ -36,10 +36,9 @@ moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('HH2O', '71t')
# 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([fuel1, fuel2, moderator])
materials_file.default_xs = '71c'
materials_file.add_materials([fuel1, fuel2, moderator])
materials_file.export_to_xml()
@ -129,7 +128,6 @@ plot.width = [20, 20]
plot.pixels = [200, 200]
plot.color = 'cell'
# Instantiate a Plots collection, add Plot, and export to XML
plot_file = openmc.Plots()
plot_file.add_plot(plot)
# Instantiate a Plots collection and export to XML
plot_file = openmc.Plots([plot])
plot_file.export_to_xml()

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

@ -35,10 +35,9 @@ iron = openmc.Material(material_id=3, name='iron')
iron.set_density('g/cc', 7.9)
iron.add_nuclide(fe56, 1.)
# 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([moderator, fuel, iron])
materials_file.default_xs = '71c'
materials_file.add_materials([moderator, fuel, iron])
materials_file.export_to_xml()
@ -152,9 +151,7 @@ plot_yz.pixels = [400, 400]
plot_yz.color = 'mat'
# Instantiate a Plots collection, add plots, and export to XML
plot_file = openmc.Plots()
plot_file.add_plot(plot_xy)
plot_file.add_plot(plot_yz)
plot_file = openmc.Plots((plot_xy, plot_yz))
plot_file.export_to_xml()
@ -167,7 +164,6 @@ tally = openmc.Tally(tally_id=1)
tally.filters = [openmc.Filter(type='distribcell', bins=[cell2.id])]
tally.scores = ['total']
# Instantiate a Tallies collection, register Tally/Mesh, and export to XML
tallies_file = openmc.Tallies()
tallies_file.add_tally(tally)
# Instantiate a Tallies collection and export to XML
tallies_file = openmc.Tallies([tally])
tallies_file.export_to_xml()

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

@ -30,10 +30,9 @@ moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('HH2O', '71t')
# 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((moderator, fuel))
materials_file.default_xs = '71c'
materials_file.add_materials([moderator, fuel])
materials_file.export_to_xml()
@ -150,9 +149,8 @@ plot.width = [4, 4]
plot.pixels = [400, 400]
plot.color = 'mat'
# Instantiate a Plots object, add Plot, and export to XML
plot_file = openmc.Plots()
plot_file.add_plot(plot)
# Instantiate a Plots object and export to XML
plot_file = openmc.Plots([plot])
plot_file.export_to_xml()
@ -177,7 +175,5 @@ tally.filters = [mesh_filter]
tally.scores = ['total']
# Instantiate a Tallies collection, register Tally/Mesh, 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()

16
examples/python/lattice/simple/build-xml.py
View file

@ -30,10 +30,9 @@ moderator.add_nuclide(h1, 2.)
moderator.add_nuclide(o16, 1.)
moderator.add_s_alpha_beta('HH2O', '71t')
# 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([moderator, fuel])
materials_file.default_xs = '71c'
materials_file.add_materials([moderator, fuel])
materials_file.export_to_xml()
@ -142,9 +141,8 @@ plot.width = [4, 4]
plot.pixels = [400, 400]
plot.color = 'mat'
# Instantiate a Plots collection, add Plot, and export to XML
plot_file = openmc.Plots()
plot_file.add_plot(plot)
# Instantiate a Plots collection and export to XML
plot_file = openmc.Plots([plot])
plot_file.export_to_xml()
@ -173,8 +171,6 @@ tally.filters = [mesh_filter]
tally.scores = ['total']
tally.triggers = [trigger]
# Instantiate a Tallies collection, register Tally/Mesh, and export to XML
tallies_file = openmc.Tallies()
tallies_file.add_mesh(mesh)
tallies_file.add_tally(tally)
# Instantiate a Tallies collection and export to XML
tallies_file = openmc.Tallies([tally])
tallies_file.export_to_xml()

11
examples/python/pincell/build-xml.py
View file

@ -100,10 +100,9 @@ borated_water.add_nuclide(o16, 2.4672e-2)
borated_water.add_nuclide(o17, 6.0099e-5)
borated_water.add_s_alpha_beta('HH2O', '71t')
# 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, helium, zircaloy, borated_water])
materials_file.default_xs = '71c'
materials_file.add_materials([uo2, helium, zircaloy, borated_water])
materials_file.export_to_xml()
@ -197,8 +196,6 @@ tally = openmc.Tally(tally_id=1, name='tally 1')
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)
# Instantiate a Tallies collection and export to XML
tallies_file = openmc.Tallies([tally])
tallies_file.export_to_xml()

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()

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

@ -23,10 +23,9 @@ fuel = openmc.Material(material_id=1, name='fuel')
fuel.set_density('g/cc', 4.5)
fuel.add_nuclide(u235, 1.)
# Instantiate a Materials collection, register Material, and export to XML
materials_file = openmc.Materials()
# Instantiate a Materials collection and export to XML
materials_file = openmc.Materials([fuel])
materials_file.default_xs = '71c'
materials_file.add_material(fuel)
materials_file.export_to_xml()

26
openmc/checkvalue.py
View file

@ -1,3 +1,4 @@
import copy
from collections import Iterable
from numbers import Integral, Real
@ -57,7 +58,7 @@ def check_type(name, value, expected_type, expected_iter_type=None):
else:
msg = 'Unable to set "{0}" to "{1}" which is not of type "{2}"'.format(
name, value, expected_type.__name__)
raise ValueError(msg)
raise TypeError(msg)
if expected_iter_type:
for item in value:
@ -71,7 +72,7 @@ def check_type(name, value, expected_type, expected_iter_type=None):
msg = 'Unable to set "{0}" to "{1}" since each item must be ' \
'of type "{2}"'.format(name, value,
expected_iter_type.__name__)
raise ValueError(msg)
raise TypeError(msg)
def check_iterable_type(name, value, expected_type, min_depth=1, max_depth=1):
@ -122,7 +123,7 @@ def check_iterable_type(name, value, expected_type, min_depth=1, max_depth=1):
if len(tree) < min_depth:
msg = 'Error setting "{0}": The item at {1} does not meet the '\
'minimum depth of {2}'.format(name, ind_str, min_depth)
raise ValueError(msg)
raise TypeError(msg)
# This item is okay. Move on to the next item.
index[-1] += 1
@ -140,7 +141,7 @@ def check_iterable_type(name, value, expected_type, min_depth=1, max_depth=1):
msg = 'Error setting {0}: Found an iterable at {1}, items '\
'in that iterable exceed the maximum depth of {2}' \
.format(name, ind_str, max_depth)
raise ValueError(msg)
raise TypeError(msg)
else:
# This item is completely unexpected.
@ -148,7 +149,7 @@ def check_iterable_type(name, value, expected_type, min_depth=1, max_depth=1):
"item at {2} is of type '{3}'"\
.format(name, expected_type.__name__, ind_str,
type(current_item).__name__)
raise ValueError(msg)
raise TypeError(msg)
def check_length(name, value, length_min, length_max=None):
@ -278,6 +279,21 @@ class CheckedList(list):
for item in items:
self.append(item)
def __add__(self, other):
new_instance = copy.copy(self)
new_instance += other
return new_instance
def __radd__(self, other):
return self + other
def __iadd__(self, other):
check_type('CheckedList add operand', other, Iterable,
self.expected_type)
for item in other:
self.append(item)
return self
def append(self, item):
"""Append item to list

122
openmc/material.py
View file

@ -8,7 +8,7 @@ if sys.version_info[0] >= 3:
basestring = str
import openmc
from openmc.checkvalue import check_type, check_value, check_greater_than
import openmc.checkvalue as cv
from openmc.clean_xml import *
@ -202,15 +202,15 @@ class Material(object):
self._id = AUTO_MATERIAL_ID
AUTO_MATERIAL_ID += 1
else:
check_type('material ID', material_id, Integral)
check_greater_than('material ID', material_id, 0, equality=True)
cv.check_type('material ID', material_id, Integral)
cv.check_greater_than('material ID', material_id, 0, equality=True)
self._id = material_id
@name.setter
def name(self, name):
if name is not None:
check_type('name for Material ID="{0}"'.format(self._id),
name, basestring)
cv.check_type('name for Material ID="{0}"'.format(self._id),
name, basestring)
self._name = name
else:
self._name = ''
@ -228,9 +228,9 @@ class Material(object):
"""
check_type('the density for Material ID="{0}"'.format(self._id),
density, Real)
check_value('density units', units, DENSITY_UNITS)
cv.check_type('the density for Material ID="{0}"'.format(self._id),
density, Real)
cv.check_value('density units', units, DENSITY_UNITS)
if density is None and units is not 'sum':
msg = 'Unable to set the density for Material ID="{0}" ' \
@ -642,9 +642,25 @@ class Material(object):
return element
class Materials(object):
"""Collection of Materials used for an OpenMC simulation. Corresponds directly
to the materials.xml input file.
class Materials(cv.CheckedList):
"""Collection of Materials used for an OpenMC simulation.
This class corresponds directly to the materials.xml input file. It can be
thought of as a normal Python list where each member is a
:class:`Material`. It behaves like a list as the following example
demonstrates:
>>> fuel = openmc.Material()
>>> clad = openmc.Material()
>>> water = openmc.Material()
>>> m = openmc.Materials([fuel])
>>> m.append(water)
>>> m += [clad]
Parameters
----------
materials : Iterable of openmc.Material
Materials to add to the collection
Attributes
----------
@ -654,10 +670,12 @@ class Materials(object):
"""
def __init__(self):
self._materials = []
def __init__(self, materials=None):
super(Materials, self).__init__(Material, 'materials collection')
self._default_xs = None
self._materials_file = ET.Element("materials")
if materials is not None:
self += materials
@property
def default_xs(self):
@ -665,11 +683,14 @@ class Materials(object):
@default_xs.setter
def default_xs(self, xs):
check_type('default xs', xs, basestring)
cv.check_type('default xs', xs, basestring)
self._default_xs = xs
def add_material(self, material):
"""Add a material to the file.
"""Append material to collection
.. deprecated:: 0.8
Use :meth:`Materials.append` instead.
Parameters
----------
@ -677,51 +698,72 @@ class Materials(object):
Material to add
"""
if not isinstance(material, Material):
msg = 'Unable to add a non-Material "{0}" to the ' \
'Materials instance'.format(material)
raise ValueError(msg)
self._materials.append(material)
warnings.warn("Materials.add_material(...) has been deprecated and may be "
"removed in a future version. Use Material.append(...) "
"instead.", DeprecationWarning)
self.append(material)
def add_materials(self, materials):
"""Add multiple materials to the file.
"""Add multiple materials to the collection
.. deprecated:: 0.8
Use compound assignment instead.
Parameters
----------
materials : tuple or list of openmc.Material
materials : Iterable of openmc.Material
Materials to add
"""
if not isinstance(materials, Iterable):
msg = 'Unable to create OpenMC materials.xml file from "{0}" which ' \
'is not iterable'.format(materials)
raise ValueError(msg)
warnings.warn("Materials.add_materials(...) has been deprecated and may be "
"removed in a future version. Use compound assignment "
"instead.", DeprecationWarning)
for material in materials:
self.add_material(material)
self.append(material)
def append(self, material):
"""Append material to collection
Parameters
----------
material : openmc.Material
Material to append
"""
super(Materials, self).append(material)
def insert(self, index, material):
"""Insert material before index
Parameters
----------
index : int
Index in list
material : openmc.Material
Material to insert
"""
super(Materials, self).insert(index, material)
def remove_material(self, material):
"""Remove a material from the file
.. deprecated:: 0.8
Use :meth:`Materials.remove` instead.
Parameters
----------
material : openmc.Material
Material to remove
"""
if not isinstance(material, Material):
msg = 'Unable to remove a non-Material "{0}" from the ' \
'Materials instance'.format(material)
raise ValueError(msg)
self._materials.remove(material)
warnings.warn("Materials.remove_material(...) has been deprecated and "
"may be removed in a future version. Use "
"Materials.remove(...) instead.", DeprecationWarning)
self.remove(material)
def make_isotropic_in_lab(self):
for material in self._materials:
for material in self:
material.make_isotropic_in_lab()
def _create_material_subelements(self):
@ -729,7 +771,7 @@ class Materials(object):
subelement = ET.SubElement(self._materials_file, "default_xs")
subelement.text = self._default_xs
for material in self._materials:
for material in self:
xml_element = material.get_material_xml()
self._materials_file.append(xml_element)

2
openmc/mgxs/library.py
View file

@ -370,7 +370,7 @@ class Library(object):
for mgxs_type in self.mgxs_types:
mgxs = self.get_mgxs(domain, mgxs_type)
for tally_id, tally in mgxs.tallies.items():
tallies_file.add_tally(tally, merge=merge)
tallies_file.append(tally, merge=merge)
def load_from_statepoint(self, statepoint):
"""Extracts tallies in an OpenMC StatePoint with the data needed to

76
openmc/plots.py
View file

@ -2,6 +2,7 @@ from collections import Iterable
from numbers import Real, Integral
from xml.etree import ElementTree as ET
import sys
import warnings
import numpy as np
@ -401,43 +402,90 @@ class Plot(object):
return element
class Plots(object):
"""Collection of Plots used for an OpenMC simulation. Corresponds directly to
the plots.xml input file.
class Plots(cv.CheckedList):
"""Collection of Plots used for an OpenMC simulation.
This class corresponds directly to the plots.xml input file. It can be
thought of as a normal Python list where each member is a :class:`Plot`. It
behaves like a list as the following example demonstrates:
>>> xz_plot = openmc.Plot()
>>> big_plot = openmc.Plot()
>>> small_plot = openmc.Plot()
>>> p = openmc.Plots((xz_plot, big_plot))
>>> p.append(small_plot)
>>> small_plot = p.pop()
Parameters
----------
plots : Iterable of openmc.Plot
Plots to add to the collection
"""
def __init__(self):
self._plots = []
def __init__(self, plots=None):
super(Plots, self).__init__(Plot, 'plots collection')
self._plots_file = ET.Element("plots")
if plots is not None:
self += plots
def add_plot(self, plot):
"""Add a plot to the file.
.. deprecated:: 0.8
Use :meth:`Plots.append` instead.
Parameters
----------
plot : openmc.Plot
Plot to add
"""
warnings.warn("Plots.add_plot(...) has been deprecated and may be "
"removed in a future version. Use Plots.append(...) "
"instead.", DeprecationWarning)
self.append(plot)
if not isinstance(plot, Plot):
msg = 'Unable to add a non-Plot "{0}" to the Plots instance'.format(plot)
raise ValueError(msg)
def append(self, plot):
"""Append plot to collection
self._plots.append(plot)
Parameters
----------
plot : openmc.Plot
Plot to append
"""
super(Plots, self).append(plot)
def insert(self, index, plot):
"""Insert plot before index
Parameters
----------
index : int
Index in list
plot : openmc.Plot
Plot to insert
"""
super(Plots, self).insert(index, plot)
def remove_plot(self, plot):
"""Remove a plot from the file.
.. deprecated:: 0.8
Use :meth:`Plots.remove` instead.
Parameters
----------
plot : openmc.Plot
Plot to remove
"""
self._plots.remove(plot)
warnings.warn("Plots.remove_plot(...) has been deprecated and may be "
"removed in a future version. Use Plots.remove(...) "
"instead.", DeprecationWarning)
self.remove(plot)
def colorize(self, geometry, seed=1):
"""Generate a consistent color scheme for each domain in each plot.
@ -455,7 +503,7 @@ class Plots(object):
"""
for plot in self._plots:
for plot in self:
plot.colorize(geometry, seed)
@ -481,11 +529,11 @@ class Plots(object):
"""
for plot in self._plots:
for plot in self:
plot.highlight_domains(geometry, domains, seed, alpha, background)
def _create_plot_subelements(self):
for plot in self._plots:
for plot in self:
xml_element = plot.get_plot_xml()
if len(plot._name) > 0:

137
openmc/tallies.py
View file

@ -3419,65 +3419,108 @@ class Tally(object):
return new_tally
class Tallies(object):
"""Collection of Tallies used for an OpenMC simulation. Corresponds directly to
the tallies.xml input file.
class Tallies(cv.CheckedList):
"""Collection of Tallies used for an OpenMC simulation.
This class corresponds directly to the tallies.xml input file. It can be
thought of as a normal Python list where each member is a :class:`Tally`. It
behaves like a list as the following example demonstrates:
>>> t1 = openmc.Tally()
>>> t2 = openmc.Tally()
>>> t3 = openmc.Tally()
>>> tallies = openmc.Tallies([t1])
>>> tallies.append(t2)
>>> tallies += [t3]
Parameters
----------
tallies : Iterable of openmc.Tally
Tallies to add to the collection
"""
def __init__(self):
self._tallies = []
self._meshes = []
def __init__(self, tallies=None):
super(Tallies, self).__init__(Tally, 'tallies collection')
self._tallies_file = ET.Element("tallies")
@property
def tallies(self):
return self._tallies
@property
def meshes(self):
return self._meshes
if tallies is not None:
self += tallies
def add_tally(self, tally, merge=False):
"""Add a tally to the file
"""Append tally to collection
.. deprecated:: 0.8
Use :meth:`Tallies.append` instead.
Parameters
----------
tally : openmc.Tally
Tally to add to file
Tally to add
merge : bool
Indicate whether the tally should be merged with an existing tally,
if possible. Defaults to False.
"""
warnings.warn("Tallies.add_tally(...) has been deprecated and may be "
"removed in a future version. Use Tallies.append(...) "
"instead.", DeprecationWarning)
self.append(tally, merge)
def append(self, tally, merge=False):
"""Append tally to collection
Parameters
----------
tally : openmc.Tally
Tally to append
merge : bool
Indicate whether the tally should be merged with an existing tally,
if possible. Defaults to False.
"""
if not isinstance(tally, Tally):
msg = 'Unable to add a non-Tally "{0}" to the Tallies instance'.format(tally)
raise ValueError(msg)
raise TypeError(msg)
if merge:
merged = False
# Look for a tally to merge with this one
for i, tally2 in enumerate(self._tallies):
for i, tally2 in enumerate(self):
# If a mergeable tally is found
if tally2.can_merge(tally):
# Replace tally 2 with the merged tally
merged_tally = tally2.merge(tally)
self._tallies[i] = merged_tally
self[i] = merged_tally
merged = True
break
# If not mergeable tally was found, simply add this tally
if not merged:
self._tallies.append(tally)
super(Tallies, self).append(tally)
else:
self._tallies.append(tally)
super(Tallies, self).append(tally)
def insert(self, index, item):
"""Insert tally before index
Parameters
----------
index : int
Index in list
item : openmc.Tally
Tally to insert
"""
super(Tallies, self).insert(index, item)
def remove_tally(self, tally):
"""Remove a tally from the file
"""Remove a tally from the collection
.. deprecated:: 0.8
Use :meth:`Tallies.remove` instead.
Parameters
----------
@ -3485,8 +3528,11 @@ class Tallies(object):
Tally to remove
"""
warnings.warn("Tallies.remove_tally(...) has been deprecated and may "
"be removed in a future version. Use Tallies.remove(...) "
"instead.", DeprecationWarning)
self._tallies.remove(tally)
self.remove(tally)
def merge_tallies(self):
"""Merge any mergeable tallies together. Note that n-way merges are
@ -3494,8 +3540,8 @@ class Tallies(object):
"""
for i, tally1 in enumerate(self._tallies):
for j, tally2 in enumerate(self._tallies):
for i, tally1 in enumerate(self):
for j, tally2 in enumerate(self):
# Do not merge the same tally with itself
if i == j:
continue
@ -3504,10 +3550,10 @@ class Tallies(object):
if tally1.can_merge(tally2):
# Replace tally 1 with the merged tally
merged_tally = tally1.merge(tally2)
self._tallies[i] = merged_tally
self[i] = merged_tally
# Remove tally 2 since it is no longer needed
self._tallies.pop(j)
self.pop(j)
# Continue iterating from the first loop
break
@ -3515,6 +3561,10 @@ class Tallies(object):
def add_mesh(self, mesh):
"""Add a mesh to the file
.. deprecated:: 0.8
Meshes that appear in a tally are automatically added to the
collection.
Parameters
----------
mesh : openmc.Mesh
@ -3522,36 +3572,43 @@ class Tallies(object):
"""
if not isinstance(mesh, Mesh):
msg = 'Unable to add a non-Mesh "{0}" to the Tallies instance'.format(mesh)
raise ValueError(msg)
self._meshes.append(mesh)
warnings.warn("Tallies.add_mesh(...) has been deprecated and may be "
"removed in a future version. Meshes that appear in a "
"tally are automatically added to the collection.",
DeprecationWarning)
def remove_mesh(self, mesh):
"""Remove a mesh from the file
.. deprecated:: 0.8
Meshes do not need to be managed explicitly.
Parameters
----------
mesh : openmc.Mesh
Mesh to remove from the file
"""
self._meshes.remove(mesh)
warnings.warn("Tallies.remove_mesh(...) has been deprecated and may be "
"removed in a future version. Meshes do not need to be "
"managed explicitly.", DeprecationWarning)
def _create_tally_subelements(self):
for tally in self._tallies:
for tally in self:
xml_element = tally.get_tally_xml()
self._tallies_file.append(xml_element)
def _create_mesh_subelements(self):
for mesh in self._meshes:
if len(mesh._name) > 0:
self._tallies_file.append(ET.Comment(mesh._name))
already_written = set()
for tally in self:
for f in tally.filters:
if f.type == 'mesh' and f.mesh not in already_written:
if len(f.mesh.name) > 0:
self._tallies_file.append(ET.Comment(f.mesh.name))
xml_element = mesh.get_mesh_xml()
self._tallies_file.append(xml_element)
xml_element = f.mesh.get_mesh_xml()
self._tallies_file.append(xml_element)
already_written.add(f.mesh)
def export_to_xml(self):
"""Create a tallies.xml file that can be used for a simulation.

8
tests/input_set.py
View file

@ -267,9 +267,9 @@ class InputSet(object):
# Define the materials file.
self.materials.default_xs = '71c'
self.materials.add_materials((fuel, clad, cold_water, hot_water,
rpv_steel, lower_rad_ref, upper_rad_ref, bot_plate, bot_nozzle,
top_nozzle, top_fa, bot_fa))
self.materials += (fuel, clad, cold_water, hot_water, rpv_steel,
lower_rad_ref, upper_rad_ref, bot_plate,
bot_nozzle, top_nozzle, top_fa, bot_fa)
# Define surfaces.
s1 = openmc.ZCylinder(R=0.41, surface_id=1)
@ -590,7 +590,7 @@ class MGInputSet(InputSet):
# Define the materials file.
self.materials.default_xs = '71c'
self.materials.add_materials((uo2, clad, water))
self.materials += (uo2, clad, water)
# Define surfaces.

7
tests/test_asymmetric_lattice/test_asymmetric_lattice.py
View file

@ -54,12 +54,11 @@ class AsymmetricLatticeTestHarness(PyAPITestHarness):
# Initialize the tallies
tally = openmc.Tally(name='distribcell tally', tally_id=27)
tally.add_filter(distrib_filter)
tally.add_score('nu-fission')
tally.filters.append(distrib_filter)
tally.scores.append('nu-fission')
# Initialize the tallies file
tallies_file = openmc.Tallies()
tallies_file.add_tally(tally)
tallies_file = openmc.Tallies([tally])
# Assign the tallies file to the input set
self._input_set.tallies = tallies_file

3
tests/test_distribmat/test_distribmat.py
View file

@ -28,9 +28,8 @@ class DistribmatTestHarness(PyAPITestHarness):
light_fuel.set_density('g/cc', 2.0)
light_fuel.add_nuclide('U-235', 1.0)
mats_file = openmc.Materials()
mats_file = openmc.Materials([moderator, dense_fuel, light_fuel])
mats_file.default_xs = '71c'
mats_file.add_materials([moderator, dense_fuel, light_fuel])
mats_file.export_to_xml()

2
tests/test_mg_max_order/test_mg_max_order.py
View file

@ -27,7 +27,7 @@ class MGNuclideInputSet(MGInputSet):
# Define the materials file.
self.materials.default_xs = '71c'
self.materials.add_materials((uo2, clad, water))
self.materials += (uo2, clad, water)
# Define surfaces.

2
tests/test_mg_nuclide/test_mg_nuclide.py
View file

@ -26,7 +26,7 @@ class MGNuclideInputSet(MGInputSet):
# Define the materials file.
self.materials.default_xs = '71c'
self.materials.add_materials((uo2, clad, water))
self.materials += (uo2, clad, water)
# Define surfaces.

21
tests/test_mg_tallies/test_mg_tallies.py
View file

@ -27,24 +27,15 @@ class MGTalliesTestHarness(PyAPITestHarness):
mat_filter = openmc.Filter(type='material', bins=[1,2,3])
tally1 = openmc.Tally(tally_id=1)
tally1.add_filter(mesh_filter)
tally1.add_score('total')
tally1.add_score('absorption')
tally1.add_score('flux')
tally1.add_score('fission')
tally1.add_score('nu-fission')
tally1.filters = [mesh_filter]
tally1.scores = ['total', 'absorption', 'flux',
'fission', 'nu-fission']
tally2 = openmc.Tally(tally_id=2)
tally2.add_filter(mat_filter)
tally2.add_filter(energy_filter)
tally2.add_filter(energyout_filter)
tally2.add_score('scatter')
tally2.add_score('nu-scatter')
tally2.filters = [mat_filter, energy_filter, energyout_filter]
tally2.scores = ['scatter', 'nu-scatter']
self._input_set.tallies = openmc.Tallies()
self._input_set.tallies.add_mesh(mesh)
self._input_set.tallies.add_tally(tally1)
self._input_set.tallies.add_tally(tally2)
self._input_set.tallies = openmc.Tallies([tally1, tally2])
super(MGTalliesTestHarness, self)._build_inputs()

3
tests/test_resonance_scattering/test_resonance_scattering.py
View file

@ -17,9 +17,8 @@ class ResonanceScatteringTestHarness(PyAPITestHarness):
mat.add_nuclide('Pu-239', 0.02)
mat.add_nuclide('H-1', 20.0)
mats_file = openmc.Materials()
mats_file = openmc.Materials([mat])
mats_file.default_xs = '71c'
mats_file.add_material(mat)
mats_file.export_to_xml()
# Geometry

3
tests/test_source/test_source.py
View file

@ -16,8 +16,7 @@ class SourceTestHarness(PyAPITestHarness):
mat1 = openmc.Material(material_id=1)
mat1.set_density('g/cm3', 4.5)
mat1.add_nuclide(openmc.Nuclide('U-235', '71c'), 1.0)
materials = openmc.Materials()
materials.add_material(mat1)
materials = openmc.Materials([mat1])
materials.export_to_xml()
sphere = openmc.Sphere(surface_id=1, R=10.0, boundary_type='vacuum')

41
tests/test_tallies/test_tallies.py
View file

@ -42,7 +42,8 @@ class TalliesTestHarness(PyAPITestHarness):
mesh_2x2.lower_left = [-182.07, -182.07]
mesh_2x2.upper_right = [182.07, 182.07]
mesh_2x2.dimension = [2, 2]
mesh_filter = Filter(type='mesh', bins=(1,))
mesh_filter = Filter(type='mesh')
mesh_filter.mesh = mesh_2x2
azimuthal_tally4 = Tally()
azimuthal_tally4.filters = [azimuthal_filter2, mesh_filter]
azimuthal_tally4.scores = ['flux']
@ -171,32 +172,18 @@ class TalliesTestHarness(PyAPITestHarness):
all_nuclide_tallies[0].estimator = 'collision'
self._input_set.tallies = Tallies()
self._input_set.tallies.add_tally(azimuthal_tally1)
self._input_set.tallies.add_tally(azimuthal_tally2)
self._input_set.tallies.add_tally(azimuthal_tally3)
self._input_set.tallies.add_tally(azimuthal_tally4)
self._input_set.tallies.add_tally(cellborn_tally)
self._input_set.tallies.add_tally(dg_tally)
self._input_set.tallies.add_tally(energy_tally)
self._input_set.tallies.add_tally(energyout_tally)
self._input_set.tallies.add_tally(transfer_tally)
self._input_set.tallies.add_tally(material_tally)
self._input_set.tallies.add_tally(mu_tally1)
self._input_set.tallies.add_tally(mu_tally2)
self._input_set.tallies.add_tally(mu_tally3)
self._input_set.tallies.add_tally(polar_tally1)
self._input_set.tallies.add_tally(polar_tally2)
self._input_set.tallies.add_tally(polar_tally3)
self._input_set.tallies.add_tally(polar_tally4)
self._input_set.tallies.add_tally(universe_tally)
[self._input_set.tallies.add_tally(t) for t in score_tallies]
[self._input_set.tallies.add_tally(t) for t in flux_tallies]
self._input_set.tallies.add_tally(scatter_tally1)
self._input_set.tallies.add_tally(scatter_tally2)
[self._input_set.tallies.add_tally(t) for t in total_tallies]
self._input_set.tallies.add_tally(questionable_tally)
[self._input_set.tallies.add_tally(t) for t in all_nuclide_tallies]
self._input_set.tallies.add_mesh(mesh_2x2)
self._input_set.tallies += (
[azimuthal_tally1, azimuthal_tally2, azimuthal_tally3,
azimuthal_tally4, cellborn_tally, dg_tally, energy_tally,
energyout_tally, transfer_tally, material_tally, mu_tally1,
mu_tally2, mu_tally3, polar_tally1, polar_tally2, polar_tally3,
polar_tally4, universe_tally])
self._input_set.tallies += score_tallies
self._input_set.tallies += flux_tallies
self._input_set.tallies += (scatter_tally1, scatter_tally2)
self._input_set.tallies += total_tallies
self._input_set.tallies.append(questionable_tally)
self._input_set.tallies += all_nuclide_tallies
self._input_set.export()

5
tests/test_tally_aggregation/test_tally_aggregation.py
View file

@ -15,9 +15,6 @@ class TallyAggregationTestHarness(PyAPITestHarness):
# The summary.h5 file needs to be created to read in the tallies
self._input_set.settings.output = {'summary': True}
# Initialize the tallies file
tallies_file = openmc.Tallies()
# Initialize the nuclides
u235 = openmc.Nuclide('U-235')
u238 = openmc.Nuclide('U-238')
@ -33,7 +30,7 @@ class TallyAggregationTestHarness(PyAPITestHarness):
tally.filters = [energy_filter, distrib_filter]
tally.scores = ['nu-fission', 'total']
tally.nuclides = [u235, u238, pu239]
tallies_file.add_tally(tally)
tallies_file = openmc.Tallies([tally])
# Export tallies to file
self._input_set.tallies = tallies_file

5
tests/test_tally_arithmetic/test_tally_arithmetic.py
View file

@ -43,14 +43,13 @@ class TallyArithmeticTestHarness(PyAPITestHarness):
tally.filters = [material_filter, energy_filter, distrib_filter]
tally.scores = ['nu-fission', 'total']
tally.nuclides = [u235, pu239]
tallies_file.add_tally(tally)
tallies_file.append(tally)
tally = openmc.Tally(name='tally 2')
tally.filters = [energy_filter, mesh_filter]
tally.scores = ['total', 'fission']
tally.nuclides = [u238, u235]
tallies_file.add_tally(tally)
tallies_file.add_mesh(mesh)
tallies_file.append(tally)
# Export tallies to file
self._input_set.tallies = tallies_file

4
tests/test_tally_slice_merge/test_tally_slice_merge.py
View file

@ -70,9 +70,7 @@ class TallySliceMergeTestHarness(PyAPITestHarness):
distribcell_tally.add_nuclide(nuclide)
# Add tallies to a Tallies object
tallies_file = openmc.Tallies()
tallies_file.add_tally(tallies[0])
tallies_file.add_tally(distribcell_tally)
tallies_file = openmc.Tallies((tallies[0], distribcell_tally))
# Export tallies to file
self._input_set.tallies = tallies_file