Merge pull request #110 from fusion-energy/odw_model

odw.Model

.github/workflows/black.yml

@@ -0,0 +1,32 @@
+name: black
+
+on:
+  push:
+    paths:
+      - '**.py'
+
+defaults:
+  run:
+    shell: bash
+
+jobs:
+  black:
+    runs-on: ubuntu-latest
+    steps:
+      - uses: actions/checkout@v2
+        with:
+          ref: ${{ github.head_ref }}
+      - name: Setup Python
+        uses: actions/setup-python@v2
+        with:
+          python-version: 3.x
+      - name: Install black
+        run: |
+          python -m pip install --upgrade pip
+          pip install black
+      - name: Run black
+        run: |
+          black .
+      - uses: stefanzweifel/git-auto-commit-action@v4
+        with:
+          commit_message: "[skip ci] Apply formatting changes"

.gitignore

@@ -142,5 +142,11 @@ dmypy.json
 *.trelis
 *.out
 *.tar.gz
-examples/neutronics_workflow-0.0.2/
-tests/neutronics_workflow-0.0.2/
+
+tests/example_01_single_volume_cell_tally/
+tests/example_02_multi_volume_cell_tally/
+tests/example_04_multi_volume_regular_mesh_tally/
+tests/example_05_3D_unstructured_mesh_tally/
+tests/output_files_produced.zip
+
+_version.py

examples/cell_tally_example.py

@@ -47,16 +47,12 @@
 # scores) to be applied to a DAGMC material or a volume
 # This cell tally applies a TBR tally to the volume(s) labeled with the
 # blanket_mat tag in the DAGMC geometry
-tally1 = odw.CellTally(
-    tally_type="TBR",
-    target="blanket_mat",
-    materials=materials)
+tally1 = odw.CellTally(tally_type="TBR", target="blanket_mat", materials=materials)
 
 # This cell tally obtains the neutron fast flux on all volumes in the problem
 tally2 = odw.CellTallies(
-    tally_types=["neutron_fast_flux"],
-    targets="all_volumes",
-    h5m_filename=h5m_filename)
+    tally_types=["neutron_fast_flux"], targets="all_volumes", h5m_filename=h5m_filename
+)
 
 # no modifications are made to the default openmc.Tallies
 tallies = openmc.Tallies([tally1] + tally2.tallies)

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",
@@ -43,25 +41,14 @@
     },
 )
 
-# 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
 # applied across the entire geomtry with and the size of the geometry is
 # automatically found.
 
-tally1 = odw.MeshTally2D(
-    tally_type="photon_effective_dose",
-    plane="xy",
-    bounding_box=my_bounding_box)
-tally2 = odw.MeshTally2D(
-    tally_type="neutron_effective_dose",
-    plane="xy",
-    bounding_box=my_bounding_box)
+tally1 = odw.MeshTally2D(tally_type="photon_effective_dose", plane="xy")
+tally2 = odw.MeshTally2D(tally_type="neutron_effective_dose", plane="xy")
 
 # no modifications are made to the default openmc.Tallies
 tallies = openmc.Tallies([tally1, tally2])
@@ -79,7 +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()

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",
@@ -44,18 +42,12 @@
     },
 )
 
-# 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
-)
+tally1 = odw.MeshTally3D(tally_type="neutron_effective_dose")
 
 # no modifications are made to the default openmc.Tallies
 tallies = openmc.Tallies([tally1])
@@ -72,7 +64,7 @@
 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

@@ -2,39 +2,33 @@
 
 import dagmc_h5m_file_inspector as di
 import openmc
-from numpy import cos, sin
 from dagmc_bounding_box import DagmcBoundingBox
+from numpy import cos, sin
+
+from .utils import check_files_exists
 
 
 class Geometry(openmc.Geometry):
-    """A openmc.Geometry object with a DAGMC Universe. If the model
-    requires a graveyard bounding box this will be automatically added. When
-    simulating a sector model reflecting surfaces can be added to complete the
-    boundary conditions.
+    """A openmc.Geometry object with a DAGMC Universe. When simulating a sector
+    model reflecting surfaces can be added to complete the boundary conditions.
 
     Args:
         h5m_filename: the filename of the h5m file containing the DAGMC
             geometry.
         reflective_angles: if a sector model is being simulated this argument
             can be used to specify the angles (in radians) to use when
             creating reflecting surfaces for a sector model.
-        graveyard_box: If a certain size of graveyard is required then the
-            upper left and lower right corners can be specified. If this is not
-            specified then the code checks to see if a graveyard exists and if
-            none are found then it makes the graveyard to encompass the geometry
     """
 
     def __init__(
         self,
         h5m_filename: str,
         reflective_angles: Tuple[float, float] = None,
-        graveyard_box=None,
     ):
         self.h5m_filename = h5m_filename
+        check_files_exists(h5m_filename)
         self.reflective_angles = reflective_angles
-        self.graveyard_box = graveyard_box
         self.dagmc_bounding_box = DagmcBoundingBox(h5m_filename)
-
         super().__init__(root=self.make_root())
 
     def corners(
@@ -61,19 +55,15 @@ def make_root(self):
             # made
 
             if "graveyard" not in di.get_materials_from_h5m(self.h5m_filename):
-                # vac_surfs = self.create_cube_of_vacuum_surfaces()
-                # # creates a cube of surfaces for the boundary conditions
-                # region = +vac_surfs[0] & \
-                #          -vac_surfs[1] & \
-                #          +vac_surfs[2] & \
-                #          -vac_surfs[3] & \
-                #          +vac_surfs[4] & \
-                #          -vac_surfs[5]
-                vac_surf = self.create_sphere_of_vacuum_surface()
+                vac_surf = openmc.Sphere(
+                    r=1000000,  # set to 10km to be big enough for models
+                    surface_id=99999999,  # set to large surface id to avoid overlaps
+                    boundary_type="vacuum",
+                )
                 region = -vac_surf
 
                 containing_cell = openmc.Cell(
-                    cell_id=9999, region=region, fill=dag_univ
+                    cell_id=99999999, region=region, fill=dag_univ
                 )
                 root = [containing_cell]
             else:
@@ -105,27 +95,8 @@ def make_root(self):
                 vac_surf = self.create_sphere_of_vacuum_surface()
                 region = -vac_surf & -reflective_1 & +reflective_2
 
-            containing_cell = openmc.Cell(
-                cell_id=9999, region=region, fill=dag_univ)
+            containing_cell = openmc.Cell(cell_id=9999, region=region, fill=dag_univ)
 
             root = [containing_cell]
 
         return root
-
-    def create_sphere_of_vacuum_surface(self):
-        """Creates a single vacuum surfaces that surround the geometry and can
-        be used as an alternative to the traditionally DAGMC graveyard cell"""
-
-        if self.graveyard_box is None:
-            from dagmc_bounding_box import DagmcBoundingBox
-
-            self.graveyard_box = DagmcBoundingBox(self.h5m_filename).corners()
-        bbox = [[*self.graveyard_box[0]], [*self.graveyard_box[1]]]
-
-        largest_radius = 3 * max(max(bbox[0]), max(bbox[1]))
-
-        sphere_surface = openmc.Sphere(
-            r=largest_radius, surface_id=9999, boundary_type="vacuum"
-        )
-
-        return sphere_surface

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/Tally.py

@@ -120,9 +120,7 @@ def compute_filters(tally_type):
         energy_function_filter_n = openmc.EnergyFunctionFilter(
             energy_bins_n, dose_coeffs_n
         )
-        additional_filters = [
-            neutron_particle_filter,
-            energy_function_filter_n]
+        additional_filters = [neutron_particle_filter, energy_function_filter_n]
     elif tally_type == "photon_effective_dose":
         energy_function_filter_p = openmc.EnergyFunctionFilter(
             energy_bins_p, dose_coeffs_p

openmc_dagmc_wrapper/__init__.py

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