added create statepoint file to example folder

.github/workflows/ci_with_install.yml

@@ -45,8 +45,7 @@ jobs:
       - name: Run examples
         run: |
           cd examples
-          cp ../tests/statepoint.2.h5 .
-          cp ../tests/statepoint.1.h5 .
+          python tests/create_statepoint_file_for_testing.py --batches 2 --particles 100
           python processing_2d_mesh_effective_dose_tally.py
           python processing_cell_damage-energy_tally.py
           python processing_cell_effective_dose_tally.py

examples/create_statepoint_file_for_testing.py

@@ -0,0 +1,181 @@
+# This minimal example makes a 3D volume and exports the shape to a stp file
+# A surrounding volume called a graveyard is needed for neutronics simulations
+
+import openmc
+import openmc_dagmc_wrapper as odw
+import openmc_plasma_source as ops
+import openmc_data_downloader as odd
+import argparse
+
+parser = argparse.ArgumentParser()
+
+parser.add_argument('-b', '--batches', type=int, default=2,
+                    help='number of batches')
+parser.add_argument('-p', '--particles', type=int, default=1000000,
+                    help='number of particles')
+args = parser.parse_args()
+
+# MATERIALS
+breeder_material = openmc.Material(1, "PbLi")  # Pb84.2Li15.8
+breeder_material.add_element("Pb", 84.2, percent_type="ao")
+breeder_material.add_element(
+    "Li",
+    15.8,
+    percent_type="ao",
+    enrichment=50.0,
+    enrichment_target="Li6",
+    enrichment_type="ao",
+)  # 50% enriched
+breeder_material.set_density("atom/b-cm", 3.2720171e-2)  # around 11 g/cm3
+
+iron = openmc.Material(name="iron")
+iron.set_density("g/cm3", 7.75)
+iron.add_element("Pb", 0.95, percent_type="wo")
+
+materials = openmc.Materials([breeder_material, iron])
+
+odd.just_in_time_library_generator(libraries="TENDL-2019", materials=materials)
+
+# GEOMETRY
+
+# surfaces
+vessel_inner = openmc.Sphere(r=500)
+first_wall_outer_surface = openmc.Sphere(r=510)
+breeder_blanket_outer_surface = openmc.Sphere(r=610, boundary_type="vacuum")
+
+
+# cells
+inner_vessel_region = -vessel_inner
+inner_vessel_cell = openmc.Cell(region=inner_vessel_region)
+
+first_wall_region = -first_wall_outer_surface & +vessel_inner
+first_wall_cell = openmc.Cell(region=first_wall_region)
+first_wall_cell.fill = iron
+
+breeder_blanket_region = +first_wall_outer_surface & -breeder_blanket_outer_surface
+breeder_blanket_cell = openmc.Cell(region=breeder_blanket_region)
+breeder_blanket_cell.fill = breeder_material
+
+universe = openmc.Universe(
+    cells=[inner_vessel_cell, first_wall_cell, breeder_blanket_cell]
+)
+geometry = openmc.Geometry(universe)
+
+
+tally1 = odw.CellTally(
+    tally_type="flux",
+    target=2,
+)
+
+tally2 = odw.CellTally(
+    tally_type="neutron_spectra",
+    target=2,
+)
+
+tally3 = odw.CellTally(
+    tally_type="TBR",
+    target=2,
+)
+
+tally4 = odw.CellTally(
+    tally_type="heating",
+    target=2,
+)
+
+tally5 = odw.CellTally(
+    tally_type="neutron_effective_dose",
+    target=2,
+)
+
+tally6 = odw.CellTally(
+    tally_type="photon_effective_dose",
+    target=2,
+)
+
+tally7 = odw.CellTally(
+    tally_type="neutron_fast_flux",
+    target=2,
+)
+
+tally8 = odw.CellTally(
+    tally_type="photon_fast_flux",
+    target=2,
+)
+
+tally9 = odw.CellTally(
+    tally_type="neutron_spectra",
+    target=2,
+)
+
+tally10 = odw.CellTally(
+    tally_type="neutron_spectra",
+    target=2,
+)
+
+tally11 = odw.CellTally(
+    tally_type="neutron_spectra",
+    target=3,
+)
+
+tally12 = odw.CellTally(
+    tally_type="photon_spectra",
+    target=2,
+)
+
+tally13 = odw.CellTally(
+    tally_type="(n,total)",
+    target=2,
+)
+
+tally14 = odw.CellTally(
+    tally_type="damage-energy",
+    target=2,
+)
+
+tally15 = odw.MeshTally2D(
+    tally_type="neutron_effective_dose",
+    plane="xy",
+    mesh_resolution=(10, 5),
+    bounding_box=[(-100, -100, 0), (100, 100, 1)],
+)
+
+tally16 = odw.MeshTally3D(
+    mesh_resolution=(100, 100, 100),
+    bounding_box=[(-100, -100, 0), (100, 100, 1)],
+    tally_type="neutron_effective_dose",
+)
+
+tallies = openmc.Tallies(
+    [
+        tally1,
+        tally2,
+        tally3,
+        tally4,
+        tally5,
+        tally6,
+        tally7,
+        tally8,
+        tally9,
+        tally10,
+        tally11,
+        tally12,
+        tally13,
+        tally14,
+        tally15,
+        tally16,
+    ]
+)
+
+
+settings = odw.FusionSettings()
+settings.batches = args.batches
+settings.particles = args.particles
+# assigns a ring source of DT energy neutrons to the source using the
+# openmc_plasma_source package
+settings.source = ops.FusionPointSource()
+
+
+my_model = openmc.model.Model(
+    materials=materials, geometry=geometry, settings=settings, tallies=tallies
+)
+statepoint_file = my_model.run()