odw.Model

examples/regular_2d_mesh_tally_example.py

@@ -24,12 +24,10 @@
 # So a set of 6 CSG surfaces are automatically made and added to the geometry
 geometry = odw.Geometry(h5m_filename=h5m_filename)
 
-# Creates the materials to use in the problem using by linking the material
-# tags in the DAGMC h5m file with material definitions in the
+# Creates the materials to use in the problem with material definitions in the
 # neutronics-material-maker. One could also use openmc.Material or nmm.Material
 # objects instead of the strings used here
 materials = odw.Materials(
-    h5m_filename=h5m_filename,
     correspondence_dict={
         "blanket_mat": "Li4SiO4",
         "blanket_rear_wall_mat": "Be",
@@ -43,11 +41,6 @@
     },
 )
 
-# makes use of the dagmc-bound-box package to get the corners of the bounding
-# box. This will be used to set the bounding box for the tally. This can be
-# expanded with the expand keyword if needed
-my_bounding_box = geometry.corners()
-
 
 # A MeshTally2D tally allows a set of standard tally types (made from filters
 # and scores) to be applied to the DAGMC geometry. By default the mesh will be
@@ -56,12 +49,12 @@
 
 tally1 = odw.MeshTally2D(
     tally_type="photon_effective_dose",
-    plane="xy",
-    bounding_box=my_bounding_box)
+    plane="xy"
+)
 tally2 = odw.MeshTally2D(
     tally_type="neutron_effective_dose",
-    plane="xy",
-    bounding_box=my_bounding_box)
+    plane="xy"
+)
 
 # no modifications are made to the default openmc.Tallies
 tallies = openmc.Tallies([tally1, tally2])
@@ -79,7 +72,7 @@
 settings.source = FusionRingSource(fuel="DT", radius=350)
 
 # no modifications are made to the default openmc.Model object
-my_model = openmc.Model(
+my_model = odw.Model(
     materials=materials, geometry=geometry, settings=settings, tallies=tallies
 )
 statepoint_file = my_model.run()

examples/regular_3d_mesh_tally_example.py

@@ -25,12 +25,10 @@
 # So a set of 6 CSG surfaces are automatically made and added to the geometry
 geometry = odw.Geometry(h5m_filename=h5m_filename)
 
-# Creates the materials to use in the problem using by linking the material
-# tags in the DAGMC h5m file with material definitions in the
+# Creates the materials to use in the problem with material definitions in the
 # neutronics-material-maker. One could also use openmc.Material or nmm.Material
 # objects instead of the strings used here
 materials = odw.Materials(
-    h5m_filename=h5m_filename,
     correspondence_dict={
         "blanket_mat": "Li4SiO4",
         "blanket_rear_wall_mat": "Be",
@@ -44,17 +42,13 @@
     },
 )
 
-# makes use of the dagmc-bound-box package to get the corners of the bounding
-# box. This will be used to set the bounding box for the tally. This can be
-# expanded with the expand keyword if needed
-my_bounding_box = geometry.corners()
 
 # A MeshTally3D tally allows a set of standard tally types (made from filters
 # and scores) to be applied to the DAGMC geometry. By default the mesh will be
 # applied across the entire geomtry with and the size of the geometry is
 # automatically found.
 tally1 = odw.MeshTally3D(
-    tally_type="neutron_effective_dose", bounding_box=my_bounding_box
+    tally_type="neutron_effective_dose"
 )
 
 # no modifications are made to the default openmc.Tallies
@@ -72,7 +66,7 @@
 settings.source = FusionRingSource(fuel="DT", radius=350)
 
 # no modifications are made to the default openmc.Model object
-my_model = openmc.Model(
+my_model = odw.Model(
     materials=materials, geometry=geometry, settings=settings, tallies=tallies
 )
 statepoint_file = my_model.run()

openmc_dagmc_wrapper/Geometry.py

@@ -117,6 +117,7 @@ def create_sphere_of_vacuum_surface(self):
         be used as an alternative to the traditionally DAGMC graveyard cell"""
 
         if self.graveyard_box is None:
+            # TODO this could be refactored using self.corners()
             from dagmc_bounding_box import DagmcBoundingBox
 
             self.graveyard_box = DagmcBoundingBox(self.h5m_filename).corners()

openmc_dagmc_wrapper/Materials.py

@@ -10,18 +10,14 @@ class Materials(openmc.Materials):
     input formats.
 
     Args:
-        h5m_filename: the filename of the h5m file containing the DAGMC
-            geometry
         correspondence_dict: A dictionary that maps the material tags present
             within the DAGMC file with materials. Materials can be provided in
             a variety of formats including neutronics_material_maker.Material
             objects, strings or openmc.Material objects.
     """
 
-    def __init__(self, h5m_filename: str, correspondence_dict: dict):
+    def __init__(self, correspondence_dict: dict):
         self.correspondence_dict = correspondence_dict
-        self.h5m_filename = h5m_filename
-        self.checks()
         self.set_openmc_materials()
         super().__init__(list(self.openmc_materials.values()))
 
@@ -43,14 +39,14 @@ def set_openmc_materials(self):
 
         self.openmc_materials = openmc_materials
 
-    def checks(self):
-        materials_in_h5m = di.get_materials_from_h5m(self.h5m_filename)
+    def checks(self, h5m_filename):
+        materials_in_h5m = di.get_materials_from_h5m(h5m_filename)
         # # checks all the required materials are present
         for reactor_material in self.correspondence_dict.keys():
             if reactor_material not in materials_in_h5m:
                 msg = (
                     f"material with tag {reactor_material} was not found in "
-                    f"the dagmc h5m file. The DAGMC file {self.h5m_filename} "
+                    f"the dagmc h5m file. The DAGMC file {h5m_filename} "
                     f"contains the following material tags {materials_in_h5m}."
                 )
                 raise ValueError(msg)

openmc_dagmc_wrapper/__init__.py

@@ -9,3 +9,5 @@
 from .tallies import *
 
 from .Settings import FusionSettings
+
+from .model import Model

openmc_dagmc_wrapper/mesh.py

@@ -9,14 +9,16 @@ def __init__(
         plane="xy",
         resolution=(400, 400),
         plane_slice_location=(-1, 1),
-        bounding_box=None  # TODO replace this by [(xmin, xmax), (ymin, ymax)]
+        # TODO replace this by bounds=[(xmin, xmax), (ymin, ymax)]
+        bounding_box=None
     ):
         self.plane = plane
         self.resolution = resolution
         self.plane_slice_location = plane_slice_location
         super().__init__(mesh_id, name)
         self.set_dimension()
-        self.set_bounds(bounding_box)
+        if bounding_box is not None:
+            self.set_bounds(bounding_box)
 
     def set_dimension(self):
         if self.plane == "xy":

openmc_dagmc_wrapper/model.py

@@ -0,0 +1,41 @@
+import openmc
+
+import openmc_dagmc_wrapper as odw
+
+
+class Model(openmc.Model):
+    def __init__(
+        self,
+        geometry=None,
+        materials=None,
+        settings=None,
+        tallies=None,
+        plots=None
+    ):
+        """Inits Model
+
+        Args:
+            geometry (odw.Geometry, optional): Geometry information. Defaults
+                to None.
+            materials (odw.Materials, optional): Materials information.
+                Defaults to None.
+            settings (openmc.Settings, optional): Settings information.
+                Defaults to None.
+            tallies (openmc.Tallies, optional): Tallies information.
+                Defaults to None.
+            plots (openmc.Plots, optional): Plot information. Defaults to None.
+        """
+        super().__init__(geometry, materials, settings, tallies, plots)
+        self.materials.checks(self.geometry.h5m_filename)
+        self.check_tallies_meshes_corners()
+
+    def check_tallies_meshes_corners(self):
+        """Iterates through tallies and check if they have a RegularMesh2D.
+        If the RegularMesh2D doesn't have corners, add them from the geometry.
+        """
+        for tally in self.tallies:
+            for filter in tally.filters:
+                if isinstance(filter, openmc.MeshFilter):
+                    if isinstance(filter.mesh, odw.RegularMesh2D):
+                        if filter.mesh.lower_left is None:
+                            filter.mesh.set_bounds(self.geometry.corners())