Fix for plotting model with multi-group cross sections

openmc/material.py

@@ -163,6 +163,8 @@ def __init__(
         # The single instance of Macroscopic data present in this material
         # (only one is allowed, hence this is different than _nuclides, etc)
         self._macroscopic = None
+
+        self._cross_sections = None
 
         # If specified, a list of table names
         self._sab = []
@@ -208,6 +210,15 @@ def __repr__(self) -> str:
 
         return string
 
+    @property
+    def cross_sections(self) -> Path | None:
+        return self._cross_sections
+
+    @cross_sections.setter
+    def cross_sections(self, cross_sections):
+        if cross_sections is not None:
+            self._cross_sections = input_path(cross_sections)        
+
     @property
     def name(self) -> str | None:
         return self._name
@@ -1918,6 +1929,30 @@ def cross_sections(self) -> Path | None:
     def cross_sections(self, cross_sections):
         if cross_sections is not None:
             self._cross_sections = input_path(cross_sections)
+            for mat in self:
+                mat.cross_sections = self.cross_sections
+
+    def _setup_cross_sections(self, material):
+        """Copy cross_sections to material if exists
+           and copy cross_sections from material if not exists
+
+        Parameters
+        ----------
+        material : openmc.Material
+
+
+        """
+        if self.cross_sections is not None:
+            if material.cross_sections is not None:
+                if material.cross_sections != self.cross_sections:
+                    warn("Material {material.id} cross_sections value has been overriden.")
+            material.cross_sections = self.cross_sections
+        elif material.cross_sections is not None:
+            self.cross_sections = material.cross_sections
+
+    def __setitem__(self, index: int, material):
+        self._setup_cross_sections(material)
+        super().__setitem__(index, material)
 
     def append(self, material):
         """Append material to collection
@@ -1928,6 +1963,7 @@ def append(self, material):
             Material to append
 
         """
+        self._setup_cross_sections(material)
         super().append(material)
 
     def insert(self, index: int, material):
@@ -1941,6 +1977,7 @@ def insert(self, index: int, material):
             Material to insert
 
         """
+        self._setup_cross_sections(material)
         super().insert(index, material)
 
     def make_isotropic_in_lab(self):

openmc/model/model.py

@@ -1159,6 +1159,17 @@ def plot(
         x_max = (origin[x] + 0.5*width[0]) * axis_scaling_factor[axis_units]
         y_min = (origin[y] - 0.5*width[1]) * axis_scaling_factor[axis_units]
         y_max = (origin[y] + 0.5*width[1]) * axis_scaling_factor[axis_units]
+
+        # Determine whether any materials contains macroscopic data and if so,
+        # set energy mode accordingly
+        for mat in self.geometry.get_all_materials().values():
+            if mat._macroscopic is not None:
+                self.settings.energy_mode = 'multi-group'
+                break
+
+        # Convert cross_section path to absolute
+        if self.materials.cross_sections is not None:
+            self.materials.cross_sections = Path(self.materials.cross_sections).resolve()
 
         # Get ID map from the C API
         id_map = self.id_map(

openmc/universe.py

@@ -337,6 +337,7 @@ def plot(self, *args, **kwargs):
         """
         model = openmc.Model()
         model.geometry = openmc.Geometry(self)
+        model.materials = openmc.Materials(self.get_all_materials().values())
         return model.plot(*args, **kwargs)
 
     def get_nuclides(self):

tests/unit_tests/test_universe.py

@@ -6,6 +6,46 @@
 from tests.unit_tests import assert_unbounded
 
 
+@pytest.fixture
+def mg_lib():
+    groups = openmc.mgxs.EnergyGroups(group_edges=[1e-5, 20.0e6])
+    h2o_xsdata = openmc.XSdata('LWTR', groups)
+    h2o_xsdata.order = 0
+    h2o_xsdata.set_total([1.0])
+    h2o_xsdata.set_absorption([0.5])
+    scatter_matrix = np.array([[[0.5]]])
+    scatter_matrix = np.rollaxis(scatter_matrix, 0, 3)
+    h2o_xsdata.set_scatter_matrix(scatter_matrix)
+    mg_cross_sections_file = openmc.MGXSLibrary(groups)
+    mg_cross_sections_file.add_xsdatas([h2o_xsdata])
+    mg_cross_sections_file.export_to_hdf5()
+    return "mgxs.h5"
+
+
+@pytest.fixture
+def mg_model(mg_lib):
+    model = openmc.Model()
+
+    # Create materials for the problem
+    h2o_data = openmc.Macroscopic('LWTR')
+
+    water = openmc.Material(name='Water')
+    water.set_density('macro', 1.0)
+    water.add_macroscopic(h2o_data)
+
+    # Instantiate a Materials collection and export to XML
+    model.materials = openmc.Materials([water])
+    model.materials.cross_sections = mg_lib
+    H = 1.0
+    L = 100
+
+    sph00 = openmc.Sphere(r=10, boundary_type = "vacuum")
+    cell00 = openmc.Cell(region = -sph00 ,  fill = water)
+    univ = openmc.Universe(cells = [cell00], universe_id = 1)
+    model.geometry = openmc.Geometry(univ)
+    return model
+
+
 def test_basic():
     c1 = openmc.Cell()
     c2 = openmc.Cell()
@@ -102,6 +142,12 @@ def test_plot(run_in_tmpdir, sphere_model):
     # Close plots to avoid warning
     import matplotlib.pyplot as plt
     plt.close('all')
+
+
+def test_mg_plot(mg_model):    
+    univ = mg_model.geometry.root_universe
+    univ.plot(width=(200, 200), basis='yz', color_by='cell')
+    univ.plot(width=(200, 200), basis='yz', color_by='material')
 
 
 def test_get_nuclides(uo2):