Fixes to example ipynb files I found while testing examples

.github/environment.yml

@@ -5,10 +5,10 @@ dependencies:
   - jupyter=1.0
   - nbformat
   - nbconvert
-  - numpy=1.18
+  - numpy
+  - seaborn
   - requests=2.24.0
   - pytest
   - vtk
   - pyevtk
-  - xlwt
   - swig

.github/workflows/ci.yml

@@ -33,15 +33,15 @@ jobs:
         with:
           activate-environment: jupyter-actions
           environment-file: .github/environment.yml
-          python-version: 3.7
+          python-version: 3.9
 
       - name: Install OpenMC
         shell: bash -l {0}
         run: |
           conda install -n base conda-libmamba-solver
           conda config --set solver libmamba
           conda activate jupyter-actions
-          conda install -c conda-forge openmc
+          conda install -c conda-forge openmc==0.14.0
 
       - name: Cache Cross Sections
         id: xs-cache

.test/test_notebooks.py

@@ -20,7 +20,7 @@ def process_notebook(notebook_filename, html_directory='notebook-html'):
     with open(notebook_filename) as f:
         nb = nbformat.read(f, as_version=4)
 
-    ep = ExecutePreprocessor(timeout=600,
+    ep = ExecutePreprocessor(timeout=1200,
                              kernel_name='python3')
 
     try:
@@ -41,7 +41,11 @@ def find_notebooks():
                  .splitlines())
 
     # Get just the notebooks from the git files
-    return [fn for fn in git_files if fn.endswith(".ipynb")]
+    notebooks = [fn for fn in git_files if fn.endswith(".ipynb")]
+    # remove the MGXS notebooks that use on OepnMOC for now
+    notebooks.remove('mgxs-part-ii.ipynb')
+    notebooks.remove('mgxs-part-iii.ipynb')
+    return notebooks
 
 
 @pytest.mark.parametrize('notebook', find_notebooks())

cad-based-geometry.ipynb

@@ -273,7 +273,7 @@
    },
    "outputs": [],
    "source": [
-    "model.settings.source = openmc.Source(space=openmc.stats.Box([-4., -4., -4.],\n",
+    "model.settings.source = openmc.IndependentSource(space=openmc.stats.Box([-4., -4., -4.],\n",
     "                                                       [ 4.,  4.,  4.]))\n",
     "model.export_to_model_xml()"
    ]
@@ -610,7 +610,7 @@
     "settings.run_mode = \"fixed source\"\n",
     "\n",
     "space = openmc.stats.Box(*geometry.bounding_box)\n",
-    "my_source = openmc.Source(space=space, domains=[water])\n",
+    "my_source = openmc.IndependentSource(space=space, domains=[water])\n",
     "my_source.energy = openmc.stats.Discrete(x=[12.0,], p=[1.0,])\n",
     "\n",
     "settings.source = my_source"
@@ -1064,7 +1064,7 @@
    ],
    "source": [
     "# creates a simple isotropic neutron source in the center with 14MeV neutrons\n",
-    "source = openmc.Source()\n",
+    "source = openmc.IndependentSource()\n",
     "# the distribution of radius is just a single value at the plasma major radius\n",
     "radius = openmc.stats.Discrete([250.], [1])\n",
     "# the distribution of source z values is just a single value\n",
@@ -1077,7 +1077,7 @@
     "# sets the direction to isotropic\n",
     "source.angle = openmc.stats.Isotropic()\n",
     "# sets the energy distribution to a Muir distribution of D-T fusion neutrons\n",
-    "source.energy = openmc.stats.Muir(e0=14_000_000.0, m_rat=5.0, kt=20000.0)"
+    "source.energy = openmc.stats.muir(e0=14_000_000.0, m_rat=5.0, kt=20000.0)"
    ]
   },
   {

candu.ipynb

@@ -359,7 +359,7 @@
     "settings.particles = 1000\n",
     "settings.batches = 20\n",
     "settings.inactive = 10\n",
-    "settings.source = openmc.Source(space=openmc.stats.Point())\n",
+    "settings.source = openmc.IndependentSource(space=openmc.stats.Point())\n",
     "settings.export_to_xml()"
    ]
   },

capi.ipynb

@@ -257,7 +257,7 @@
     "settings_file.inactive = 10\n",
     "settings_file.particles = 10000\n",
     "settings_file.temperature = {'multipole': True, 'method': 'interpolation', 'range': [290, 2500]}\n",
-    "settings_file.source = openmc.source.Source(space=uniform_dist)\n",
+    "settings_file.source = openmc.source.IndependentSource(space=uniform_dist)\n",
     "settings_file.export_to_xml()"
    ]
   },

chain_simple.xml

@@ -1,5 +1,9 @@
 <?xml version="1.0"?>
 <depletion_chain>
+  <nuclide name="Ba135" decay_modes="0" reactions="0" />
+  <nuclide name="Ba136" decay_modes="0" reactions="0" />
+  <nuclide name="Ba138" decay_modes="0" reactions="0" />
+  <nuclide name="I134" decay_modes="0" reactions="0" />
   <nuclide name="I135" decay_modes="1" reactions="1" half_life="2.36520E+04">
     <decay type="beta" target="Xe135" branching_ratio="1.0" />
     <reaction type="(n,gamma)" Q="0.0" target="Xe136" /> <!-- Not precisely true, but whatever -->
@@ -8,9 +12,14 @@
     <decay type=" beta" target="Cs135" branching_ratio="1.0" />
     <reaction type="(n,gamma)" Q="0.0" target="Xe136" />
   </nuclide>
+  <nuclide name="Xe134" decay_modes="0" reactions="0" />
   <nuclide name="Xe136" decay_modes="0" reactions="0" />
+  <nuclide name="Xe137" decay_modes="0" reactions="0" />
   <nuclide name="Cs135" decay_modes="0" reactions="0" />
-  <nuclide name="Gd157" decay_modes="0" reactions="1"  >
+  <nuclide name="Cs136" decay_modes="0" reactions="0" />
+  <nuclide name="Cs138" decay_modes="0" reactions="0" />
+  <nuclide name="Cs138_m1" decay_modes="0" reactions="0" />
+  <nuclide name="Gd157" decay_modes="0" reactions="1">
     <reaction type="(n,gamma)" Q="0.0" target="Nothing" />
   </nuclide>
     <nuclide name="Cs137" half_life="949252600.0" decay_modes="2" decay_energy="179447.6443" reactions="6">

depletion.ipynb

@@ -4,7 +4,7 @@
    "cell_type": "markdown",
    "metadata": {},
    "source": [
-    "# Depletion\n",
+    "# Pincell Depletion\n",
     "This notebook is intended to introduce the reader to the depletion interface contained in OpenMC. It is recommended that you are moderately familiar with building models using the OpenMC Python API. The earlier examples are excellent starting points, as this notebook will not focus heavily on model building.\n",
     "\n",
     "If you have a real power reactor, the fuel composition is constantly changing as fission events produce energy, remove some fissile isotopes, and produce fission products. Other reactions, like $(n, \\alpha)$ and $(n, \\gamma)$ will alter the composition as well. Furthermore, some nuclides undergo spontaneous decay with widely ranging frequencies. Depletion is the process of modeling this behavior.\n",
@@ -105,8 +105,8 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "bound_box = openmc.rectangular_prism(1.24, 1.24, boundary_type=\"reflective\")\n",
-    "root_cell = openmc.Cell(fill=pin_univ, region=bound_box)\n",
+    "bound_box = openmc.model.RectangularPrism(1.24, 1.24, boundary_type=\"reflective\")\n",
+    "root_cell = openmc.Cell(fill=pin_univ, region=-bound_box)\n",
     "geometry = openmc.Geometry([root_cell])"
    ]
   },

expansion-filters.ipynb

@@ -63,16 +63,16 @@
    "source": [
     "# Define surfaces used to construct regions\n",
     "zmin, zmax = -10., 10.\n",
-    "box = openmc.model.rectangular_prism(10., 10., boundary_type='reflective')\n",
+    "box = openmc.model.RectangularPrism(10., 10., boundary_type='reflective')\n",
     "bottom = openmc.ZPlane(z0=zmin, boundary_type='vacuum')\n",
     "boron_lower = openmc.ZPlane(z0=-0.5)\n",
     "boron_upper = openmc.ZPlane(z0=0.5)\n",
     "top = openmc.ZPlane(z0=zmax, boundary_type='vacuum')\n",
     "\n",
     "# Create three cells and add them to geometry\n",
-    "fuel1 = openmc.Cell(fill=fuel, region=box & +bottom & -boron_lower)\n",
-    "absorber = openmc.Cell(fill=b4c, region=box & +boron_lower & -boron_upper)\n",
-    "fuel2 = openmc.Cell(fill=fuel, region=box & +boron_upper & -top)\n",
+    "fuel1 = openmc.Cell(fill=fuel, region=-box & +bottom & -boron_lower)\n",
+    "absorber = openmc.Cell(fill=b4c, region=-box & +boron_lower & -boron_upper)\n",
+    "fuel2 = openmc.Cell(fill=fuel, region=-box & +boron_upper & -top)\n",
     "geom = openmc.Geometry([fuel1, absorber, fuel2])"
    ]
   },
@@ -91,7 +91,7 @@
    "source": [
     "settings = openmc.Settings()\n",
     "spatial_dist = openmc.stats.Box(*geom.bounding_box)\n",
-    "settings.source = openmc.Source(space=spatial_dist)\n",
+    "settings.source = openmc.IndependentSource(space=spatial_dist)\n",
     "settings.batches = 210\n",
     "settings.inactive = 10\n",
     "settings.particles = 1000"
@@ -506,7 +506,7 @@
    "source": [
     "settings = openmc.Settings()\n",
     "spatial_dist = openmc.stats.Box(*geom.bounding_box)\n",
-    "settings.source = openmc.Source(space=spatial_dist)\n",
+    "settings.source = openmc.IndependentSource(space=spatial_dist)\n",
     "settings.batches = 100\n",
     "settings.inactive = 20\n",
     "settings.particles = 100000"

hexagonal-lattice.ipynb

@@ -373,7 +373,7 @@
     }
    ],
    "source": [
-    "main_cell.region = openmc.model.hexagonal_prism(\n",
+    "main_cell.region = -openmc.model.HexagonalPrism(\n",
     "    edge_length=4*lattice.pitch[0],\n",
     "    orientation='x',\n",
     "    boundary_type='vacuum'\n",

mdgxs-part-ii.ipynb

@@ -297,7 +297,7 @@
     "# Create an initial uniform spatial source distribution over fissionable zones\n",
     "bounds = [-10.71, -10.71, -10, 10.71, 10.71, 10.]\n",
     "uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)\n",
-    "settings.source = openmc.Source(space=uniform_dist)"
+    "settings.source = openmc.IndependentSource(space=uniform_dist)"
    ]
   },
   {
@@ -376,12 +376,10 @@
    "outputs": [],
    "source": [
     "# Instantiate a 20-group EnergyGroups object\n",
-    "energy_groups = openmc.mgxs.EnergyGroups()\n",
-    "energy_groups.group_edges = np.logspace(-3, 7.3, 21)\n",
+    "energy_groups = openmc.mgxs.EnergyGroups(group_edges=np.logspace(-3, 7.3, 21))\n",
     "\n",
     "# Instantiate a 1-group EnergyGroups object\n",
-    "one_group = openmc.mgxs.EnergyGroups()\n",
-    "one_group.group_edges = np.array([energy_groups.group_edges[0], energy_groups.group_edges[-1]])"
+    "one_group = openmc.mgxs.EnergyGroups(group_edges=np.array([energy_groups.group_edges[0], energy_groups.group_edges[-1]]))"
    ]
   },
   {

mg-mode-part-i.ipynb

@@ -506,7 +506,7 @@
     "# Create an initial uniform spatial source distribution over fissionable zones\n",
     "bounds = [-32.13, -10.71, -1e50, 10.71, 32.13, 1e50]\n",
     "uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)\n",
-    "settings.source = openmc.Source(space=uniform_dist)\n",
+    "settings.source = openmc.IndependentSource(space=uniform_dist)\n",
     "\n",
     "# Tell OpenMC we want to run in eigenvalue mode\n",
     "settings.run_mode = 'eigenvalue'\n",

mg-mode-part-ii.ipynb

@@ -358,7 +358,7 @@
     "# Create an initial uniform spatial source distribution over fissionable zones\n",
     "bounds = [-10.71, -10.71, -10, 10.71, 10.71, 10.]\n",
     "uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)\n",
-    "settings.source = openmc.Source(space=uniform_dist)\n",
+    "settings.source = openmc.IndependentSource(space=uniform_dist)\n",
     "\n",
     "model.settings = settings"
    ]
@@ -914,24 +914,24 @@
    "source": [
     "# First lets plot the fuel data\n",
     "# We will first add the continuous-energy data\n",
-    "fig = openmc.plot_xs(fuel, ['total'])\n",
+    "fig = openmc.plot_xs(reactions={fuel: ['total']})\n",
     "\n",
     "# We will now add in the corresponding multi-group data and show the result\n",
-    "openmc.plot_xs(fuel_mg, ['total'], plot_CE=False, mg_cross_sections='mgxs.h5', axis=fig.axes[0])\n",
+    "openmc.plot_xs(reactions={fuel_mg: ['total']}, plot_CE=False, mg_cross_sections='mgxs.h5', axis=fig.axes[0])\n",
     "fig.axes[0].legend().set_visible(False)\n",
     "plt.show()\n",
     "plt.close()\n",
     "\n",
     "# Then repeat for the zircaloy data\n",
-    "fig = openmc.plot_xs(zircaloy, ['total'])\n",
-    "openmc.plot_xs(zircaloy_mg, ['total'], plot_CE=False, mg_cross_sections='mgxs.h5', axis=fig.axes[0])\n",
+    "fig = openmc.plot_xs(reactions={zircaloy: ['total']})\n",
+    "openmc.plot_xs(reactions={zircaloy_mg: ['total']}, plot_CE=False, mg_cross_sections='mgxs.h5', axis=fig.axes[0])\n",
     "fig.axes[0].legend().set_visible(False)\n",
     "plt.show()\n",
     "plt.close()\n",
     "\n",
     "# And finally repeat for the water data\n",
-    "fig = openmc.plot_xs(water, ['total'])\n",
-    "openmc.plot_xs(water_mg, ['total'], plot_CE=False, mg_cross_sections='mgxs.h5', axis=fig.axes[0])\n",
+    "fig = openmc.plot_xs(reactions={water: ['total']})\n",
+    "openmc.plot_xs(reactions={water_mg: ['total']}, plot_CE=False, mg_cross_sections='mgxs.h5', axis=fig.axes[0])\n",
     "fig.axes[0].legend().set_visible(False)\n",
     "plt.show()\n",
     "plt.close()"

mg-mode-part-iii.ipynb

@@ -445,7 +445,7 @@
     "# Create an initial uniform spatial source distribution over fissionable zones\n",
     "bounds = [pin_pitch, pin_pitch, 10, length - pin_pitch, length - pin_pitch, 10]\n",
     "uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)\n",
-    "settings.source = openmc.Source(space=uniform_dist)\n",
+    "settings.source = openmc.IndependentSource(space=uniform_dist)\n",
     "\n",
     "model.settings = settings"
    ]
@@ -466,8 +466,7 @@
    "outputs": [],
    "source": [
     "# Instantiate a 2-group EnergyGroups object\n",
-    "groups = openmc.mgxs.EnergyGroups()\n",
-    "groups.group_edges = np.array([0., 0.625, 20.0e6])"
+    "groups = openmc.mgxs.EnergyGroups(group_edges=np.array([0., 0.625, 20.0e6]))"
    ]
   },
   {