Merge pull request #61 from fusion-energy/develop

Major refactoring to use OpenMC classes

.circleci/config.yml

@@ -16,29 +16,36 @@ jobs:
             python setup.py install
 
       - run:
-          name: run test_NeutronicModel
+          name: run test_neutronics_utils
           command: 
-            pytest tests/test_neutronics_model.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
+            pytest tests/test_neutronics_utils.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
 
       - run:
-          name: run test_shape_neutronics
+          name: run tests Settings()
           command: 
-            pytest tests/test_shape_neutronics.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
+            pytest tests/test_settings.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
 
       - run:
-          name: run test_reactor_neutronics
+          name: run tests Materials()
           command: 
-            pytest tests/test_reactor_neutronics.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
+            pytest tests/test_materials.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
 
-      # - run:
-      #     name: run test_neutronics_utils
-      #     command: 
-      #       pytest tests/test_neutronics_utils.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
+      - run:
+          name: run tests Tallies
+          command: 
+            pytest tests/test_tallies/ -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
 
       - run:
-          name: run test_example_neutronics_simulations
+          name: System tests
           command: 
-            pytest tests/test_example_neutronics_simulations.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
+            pytest tests/test_system/ -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
+
+      # - run:
+      #     name: run test_example_neutronics_simulations
+      #     command:
+      #       pytest tests/test_example_neutronics_simulations.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
+
+
 
       # TODO add example notebooks
       # - run:

.github/workflows/ci_with_install.yml

@@ -25,9 +25,9 @@ jobs:
 
       - name: run tests
         run: |
+          pytest tests/test_neutronics_utils.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
           pytest tests/test_example_neutronics_simulations.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
-          pytest tests/test_neutronics_model.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
-          # pytest tests/test_neutronics_utils.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
-          pytest tests/test_reactor_neutronics.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
-          pytest tests/test_shape_neutronics.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
-          python tests/notebook_testing.py
+          pytest tests/test_settings.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
+          pytest tests/test_materials.py -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
+          pytest tests/test_tallies/ -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml
+          pytest tests/test_system/ -v --cov=openmc_dagmc_wrapper --cov-append --cov-report term --cov-report xml

examples/cell_tally_example.py

@@ -0,0 +1,87 @@
+# A minimal example that obtains TBR on the blanket and fast neutron flux on all
+# cells in the DAGMC geometry.
+# Particular emphasis is placed on explaining the openmc-dagmc-wrapper
+# extentions of openmc base classes.
+
+import tarfile
+import urllib.request
+
+import openmc
+import openmc_dagmc_wrapper as odw
+from openmc_plasma_source import FusionRingSource
+
+# downloads a dagmc file for use in the example
+url = "https://github.com/fusion-energy/neutronics_workflow/archive/refs/tags/v0.0.2.tar.gz"
+urllib.request.urlretrieve(url, "v0.0.2.tar.gz")
+tar = tarfile.open("v0.0.2.tar.gz", "r:gz")
+tar.extractall(".")
+tar.close()
+h5m_filename = "neutronics_workflow-0.0.2/example_02_multi_volume_cell_tally/stage_2_output/dagmc.h5m"
+
+
+# creates a geometry object from a DAGMC geometry.
+# In this case the geometry doen't have a graveyard cell.
+# 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
+# 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",
+        "center_column_shield_mat": "Be",
+        "divertor_mat": "Be",
+        "firstwall_mat": "Be",
+        "inboard_tf_coils_mat": "Be",
+        "pf_coil_case_mat": "Be",
+        "pf_coil_mat": "Be",
+        "tf_coil_mat": "Be",
+    },
+)
+
+# A cell tally allows a set of standard tally types (made from filters and
+# 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)
+
+# 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)
+
+# no modifications are made to the default openmc.Tallies
+tallies = openmc.Tallies([tally1] + tally2.tallies)
+
+# Creates and openmc settings object with the run mode set to 'fixed source'
+# and the number of inactivate particles set to zero. Setting these to values
+# by default means less code is needed by the user and less chance of simulating
+# batches that don't contribute to the tallies
+settings = odw.FusionSettings()
+settings.batches = 1
+settings.particles = 100
+# assigns a ring source of DT energy neutrons to the source using the
+# openmc_plasma_source package
+settings.source = FusionRingSource(fuel="DT", radius=350)
+
+
+# no modifications are made to the default openmc.Model object
+my_model = openmc.Model(
+    materials=materials, geometry=geometry, settings=settings, tallies=tallies
+)
+statepoint_file = my_model.run()
+
+# processes the output h5 file. the process_results function contains logic on
+# how to process each tally with respect to the tally multipliers and units
+# involved. The fusion power input allows tallies to be scaled from the units
+# of per source neutron to units such as Watts (for heating), Sv per second
+# (for dose) and other convenient units.
+odw.process_results(statepoint_file, fusion_power=1e9)

examples/regular_2d_mesh_tally_example.py

@@ -0,0 +1,86 @@
+# A minimal example that obtains TBR on the blanket and fast neutron flux on all
+# cells in the DAGMC geometry.
+# Particular emphasis is placed on explaining the openmc-dagmc-wrapper
+# extentions of openmc base classes.
+
+import tarfile
+import urllib.request
+
+import openmc
+import openmc_dagmc_wrapper as odw
+from openmc_plasma_source import FusionRingSource
+
+
+# downloads a dagmc file for use in the example
+# url = "https://github.com/fusion-energy/neutronics_workflow/archive/refs/tags/v0.0.2.tar.gz"
+# urllib.request.urlretrieve(url, "v0.0.2.tar.gz")
+# tar = tarfile.open("v0.0.2.tar.gz", "r:gz")
+# tar.extractall(".")
+# tar.close()
+h5m_filename = "neutronics_workflow-0.0.2/example_02_multi_volume_cell_tally/stage_2_output/dagmc.h5m"
+h5m_filename = "neutronics_workflow-0.0.2/example_02_multi_volume_cell_tally/stage_2_output/dagmc_no_grave_yard.h5m"
+
+
+# creates a geometry object from a DAGMC geometry.
+# In this case the geometry doen't have a graveyard cell.
+# 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
+# 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",
+        "center_column_shield_mat": "Be",
+        "divertor_mat": "Be",
+        "firstwall_mat": "Be",
+        "inboard_tf_coils_mat": "Be",
+        "pf_coil_case_mat": "Be",
+        "pf_coil_mat": "Be",
+        "tf_coil_mat": "Be",
+    },
+)
+
+# 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=h5m_filename
+)
+tally2 = odw.MeshTally2D(
+    tally_type="neutron_effective_dose", plane="xy", bounding_box=h5m_filename
+)
+
+# no modifications are made to the default openmc.Tallies
+tallies = openmc.Tallies([tally1, tally2])
+
+# Creates and openmc settings object with the run mode set to 'fixed source'
+# and the number of inactivate particles set to zero. Setting these to values
+# by default means less code is needed by the user and less chance of simulating
+# batches that don't contribute to the tallies
+settings = odw.FusionSettings()
+settings.batches = 2
+settings.particles = 100
+settings.photon_transport = True
+# assigns a ring source of DT energy neutrons to the source using the
+# openmc_plasma_source package
+settings.source = FusionRingSource(fuel="DT", radius=350)
+
+# no modifications are made to the default openmc.Model object
+my_model = openmc.Model(
+    materials=materials, geometry=geometry, settings=settings, tallies=tallies
+)
+statepoint_file = my_model.run()
+
+# processes the output h5 file. The process_results function contains logic on
+# how to process each tally with respect to the tally multipliers and units
+# involved. The fusion power input allows tallies to be scaled from the units
+# of per source neutron to units such as Watts (for heating), Sv per second
+# (for dose) and other convenient units.
+odw.process_results(statepoint_file, fusion_power=1e9)

examples/regular_3d_mesh_tally_example.py

@@ -0,0 +1,79 @@
+# A minimal example that obtains TBR on the blanket and fast neutron flux on all
+# cells in the DAGMC geometry.
+# Particular emphasis is placed on explaining the openmc-dagmc-wrapper
+# extentions of openmc base classes.
+
+import tarfile
+import urllib.request
+
+import openmc
+import openmc_dagmc_wrapper as odw
+from openmc_plasma_source import FusionRingSource
+
+# downloads a dagmc file for use in the example
+# url = "https://github.com/fusion-energy/neutronics_workflow/archive/refs/tags/v0.0.2.tar.gz"
+# urllib.request.urlretrieve(url, "v0.0.2.tar.gz")
+# tar = tarfile.open("v0.0.2.tar.gz", "r:gz")
+# tar.extractall(".")
+# tar.close()
+h5m_filename = "neutronics_workflow-0.0.2/example_02_multi_volume_cell_tally/stage_2_output/dagmc.h5m"
+
+
+# creates a geometry object from a DAGMC geometry.
+# In this case the geometry doen't have a graveyard cell.
+# 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
+# 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",
+        "center_column_shield_mat": "Be",
+        "divertor_mat": "Be",
+        "firstwall_mat": "Be",
+        "inboard_tf_coils_mat": "Be",
+        "pf_coil_case_mat": "Be",
+        "pf_coil_mat": "Be",
+        "tf_coil_mat": "Be",
+    },
+)
+
+# 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=h5m_filename)
+
+# no modifications are made to the default openmc.Tallies
+tallies = openmc.Tallies([tally1])
+
+# Creates and openmc settings object with the run mode set to 'fixed source'
+# and the number of inactivate particles set to zero. Setting these to values
+# by default means less code is needed by the user and less chance of simulating
+# batches that don't contribute to the tallies
+settings = odw.FusionSettings()
+settings.batches = 1
+settings.particles = 100
+# assigns a ring source of DT energy neutrons to the source using the
+# openmc_plasma_source package
+settings.source = FusionRingSource(fuel="DT", radius=350)
+
+# no modifications are made to the default openmc.Model object
+my_model = openmc.Model(
+    materials=materials, geometry=geometry, settings=settings, tallies=tallies
+)
+statepoint_file = my_model.run()
+
+# processes the output h5 file. The process_results function contains logic on
+# how to process each tally with respect to the tally multipliers and units
+# involved. The fusion power input allows tallies to be scaled from the units
+# of per source neutron to units such as Watts (for heating), Sv per second
+# (for dose) and other convenient units.
+odw.process_results(statepoint_file, fusion_power=1e9)