simpler examples

.github/workflows/ci.yml

@@ -24,21 +24,10 @@ jobs:
           source "${HOME}/conda/etc/profile.d/mamba.sh"
           mamba activate
           mamba install -y -c conda-forge openmc
-          pip install openmc_data_downloader
-          openmc_data_downloader -l ENDFB-7.1-NNDC -i Fe56 Be9 Li6 Li7
-
-      - name: Install
-        run: |
-          python3 -m pip install .
-          python3 -c "import openmc_plasma_source"
-          
-      - name: Run tests
-        run: |
-          python3 -m pip install .[tests]
-          python3 -m pytest -v tests
-
-      - name: Run examples
-        run: |
-          python3 examples/point_source_example.py
-          python3 examples/ring_source_example.py
-          python3 examples/tokamak_source_example.py
+          pip install .
+          python -c "import openmc_plasma_source"
+          pip install .[tests]
+          pytest -v tests
+          python examples/point_source_example.py
+          python examples/ring_source_example.py
+          python examples/tokamak_source_example.py

examples/point_source_example.py

@@ -1,23 +1,17 @@
 import openmc
 from openmc_plasma_source import FusionPointSource
 
-my_source = FusionPointSource(coordinate=(0, 0, 0), temperature=20000.0, fuel="DT")
-
-# Create a single material
-iron = openmc.Material()
-iron.set_density("g/cm3", 5.0)
-iron.add_element("Fe", 1.0)
-mats = openmc.Materials([iron])
-
-# Create a 5 cm x 5 cm box filled with iron
-cells = []
-inner_box1 = openmc.ZCylinder(r=600.0)
-inner_box2 = openmc.ZCylinder(r=1400.0)
-outer_box = openmc.model.rectangular_prism(4000.0, 4000.0, boundary_type="vacuum")
-cells += [openmc.Cell(fill=iron, region=-inner_box1)]
-cells += [openmc.Cell(fill=None, region=+inner_box1 & -inner_box2)]
-cells += [openmc.Cell(fill=iron, region=+inner_box2 & outer_box)]
-geometry = openmc.Geometry(cells)
+# minimal geometry
+sphere_surface = openmc.Sphere(r=1000.0, boundary_type='vacuum')
+cell = openmc.Cell(region=-sphere_surface)
+geometry = openmc.Geometry([cell])
+
+# define the source
+my_source = FusionPointSource(
+    coordinate=(0, 0, 0),
+    temperature=20000.0,
+    fuel="DT"
+)
 
 # Tell OpenMC we're going to use our custom source
 settings = openmc.Settings()
@@ -26,19 +20,9 @@
 settings.particles = 1000
 settings.source = my_source
 
-# Finally, define a mesh tally so that we can see the resulting flux
-mesh = openmc.RegularMesh()
-mesh.lower_left = (-2000.0, -2000.0)
-mesh.upper_right = (2000.0, 2000.0)
-mesh.dimension = (50, 50)
-
-tally = openmc.Tally()
-tally.filters = [openmc.MeshFilter(mesh)]
-tally.scores = ["flux"]
-tallies = openmc.Tallies([tally])
 
 model = openmc.model.Model(
-    materials=mats, geometry=geometry, settings=settings, tallies=tallies
+    materials=None, geometry=geometry, settings=settings
 )
 
 model.run()

examples/ring_source_example.py

@@ -2,48 +2,27 @@
 from openmc_plasma_source import FusionRingSource
 import math
 
+# minimal geometry
+sphere_surface = openmc.Sphere(r=1000.0, boundary_type='vacuum')
+cell = openmc.Cell(region=-sphere_surface)
+geometry = openmc.Geometry([cell])
+
+# define the source
 my_source = FusionRingSource(
     radius=700,
     angles=(0.0, 2 * math.pi),  # 360deg source
     temperature=20000.0,
 )
 
-# Create a single material
-iron = openmc.Material()
-iron.set_density("g/cm3", 5.0)
-iron.add_element("Fe", 1.0)
-mats = openmc.Materials([iron])
-
-# Create a 5 cm x 5 cm box filled with iron
-cells = []
-inner_box1 = openmc.ZCylinder(r=600.0)
-inner_box2 = openmc.ZCylinder(r=1400.0)
-outer_box = openmc.model.rectangular_prism(4000.0, 4000.0, boundary_type="vacuum")
-cells += [openmc.Cell(fill=iron, region=-inner_box1)]
-cells += [openmc.Cell(fill=None, region=+inner_box1 & -inner_box2)]
-cells += [openmc.Cell(fill=iron, region=+inner_box2 & outer_box)]
-geometry = openmc.Geometry(cells)
-
-# Tell OpenMC we're going to use our custom source
 settings = openmc.Settings()
 settings.run_mode = "fixed source"
 settings.batches = 10
 settings.particles = 1000
+# tell OpenMC we're going to use our custom source
 settings.source = my_source
 
-# Finally, define a mesh tally so that we can see the resulting flux
-mesh = openmc.RegularMesh()
-mesh.lower_left = (-2000.0, -2000.0)
-mesh.upper_right = (2000.0, 2000.0)
-mesh.dimension = (50, 50)
-
-tally = openmc.Tally()
-tally.filters = [openmc.MeshFilter(mesh)]
-tally.scores = ["flux"]
-tallies = openmc.Tallies([tally])
-
 model = openmc.model.Model(
-    materials=mats, geometry=geometry, settings=settings, tallies=tallies
+    materials=None, geometry=geometry, settings=settings
 )
 
 model.run()

examples/tokamak_source_example.py

@@ -1,6 +1,10 @@
 import openmc
 from openmc_plasma_source import TokamakSource
 
+# minimal geometry
+sphere_surface = openmc.Sphere(r=1000.0, boundary_type='vacuum')
+cell = openmc.Cell(region=-sphere_surface)
+geometry = openmc.Geometry([cell])
 
 # create a plasma source
 my_plasma = TokamakSource(
@@ -22,43 +26,16 @@
     ion_temperature_beta=6,
 )
 
-
-# Create a single material
-iron = openmc.Material()
-iron.set_density("g/cm3", 5.0)
-iron.add_element("Fe", 1.0)
-mats = openmc.Materials([iron])
-
-# Create a 5 cm x 5 cm box filled with iron
-cells = []
-inner_box1 = openmc.ZCylinder(r=600.0)
-inner_box2 = openmc.ZCylinder(r=1400.0)
-outer_box = openmc.model.rectangular_prism(4000.0, 4000.0, boundary_type="vacuum")
-cells += [openmc.Cell(fill=iron, region=-inner_box1)]
-cells += [openmc.Cell(fill=None, region=+inner_box1 & -inner_box2)]
-cells += [openmc.Cell(fill=iron, region=+inner_box2 & outer_box)]
-geometry = openmc.Geometry(cells)
-
 # Tell OpenMC we're going to use our custom source
 settings = openmc.Settings()
 settings.run_mode = "fixed source"
 settings.batches = 10
 settings.particles = 1000
 settings.source = my_plasma.sources
 
-# Finally, define a mesh tally so that we can see the resulting flux
-mesh = openmc.RegularMesh()
-mesh.lower_left = (-2000.0, -2000.0)
-mesh.upper_right = (2000.0, 2000.0)
-mesh.dimension = (50, 50)
-
-tally = openmc.Tally()
-tally.filters = [openmc.MeshFilter(mesh)]
-tally.scores = ["flux"]
-tallies = openmc.Tallies([tally])
 
 model = openmc.model.Model(
-    materials=mats, geometry=geometry, settings=settings, tallies=tallies
+    materials=None, geometry=geometry, settings=settings
 )
 
 model.run()