Merge pull request #12 from fusion-energy/develop

added requested_units to functions

CITATION.cff

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+cff-version: 1.2.0
+message: "If you use this software, please cite it as below."
+authors:
+- family-names: "Shimwell"
+  given-names: "Jonathan"
+  orcid: "https://orcid.org/0000-0001-6909-0946"
+title: "A Python package for plotting regular mesh tally results from neutronics simulations."
+version: 0.1.0
+date-released: 2021-10-30
+url: "https://github.com/fusion-energy/regular_mesh_plotter"

examples/create_statepoint_file_with_meshes_openmc.py

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+# 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
+
+
+# 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("Li", 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.MeshTally2D(
+    tally_type="neutron_effective_dose",
+    plane="xy",
+    mesh_resolution=(10, 5),
+    bounding_box=[(-100, -100, 0), (100, 100, 1)],
+)
+
+tally2 = odw.MeshTally3D(
+    mesh_resolution=(100, 100, 100),
+    bounding_box=[(-100, -100, 0), (100, 100, 1)],
+    tally_type="neutron_effective_dose",
+)
+
+tally3 = odw.MeshTally2D(
+    tally_type="neutron_flux",
+    plane="xy",
+    mesh_resolution=(10, 5),
+    bounding_box=[(-100, -100, 0), (100, 100, 1)],
+)
+
+tallies = openmc.Tallies(
+    [
+        tally1,
+        tally2,
+        tally3,
+    ]
+)
+
+settings = odw.FusionSettings()
+settings.batches = 2
+settings.particles = 1000000
+# 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()

examples/create_statepoint_file_with_meshes_openmc_dagmc.py

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+# 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 paramak
+from stl_to_h5m import stl_to_h5m
+
+my_shape = paramak.ExtrudeStraightShape(
+    points=[(1, 1), (1, 200), (600, 200), (600, 1)],
+    distance=180,
+)
+
+my_shape.export_stl("example.stl")
+
+# This script converts the CAD stl files generated into h5m files that can be
+# used in DAGMC enabled codes. h5m files created in this way are imprinted,
+# merged, faceted and ready for use in OpenMC. One of the key aspects of this
+# is the assignment of materials to the volumes present in the CAD files.
+
+stl_to_h5m(
+    files_with_tags=[("example.stl", "mat1")],
+    h5m_filename="dagmc.h5m",
+)
+
+# makes use of the previously created neutronics geometry (h5m file) and assigns
+# actual materials to the material tags. Sets simulation intensity and specifies
+# the neutronics results to record (know as tallies).
+
+geometry = odw.Geometry(
+    h5m_filename="dagmc.h5m",
+)
+
+materials = odw.Materials(
+    h5m_filename="dagmc.h5m", correspondence_dict={"mat1": "eurofer"}
+)
+
+tally1 = odw.MeshTally2D(
+    tally_type="neutron_effective_dose",
+    plane="xy",
+    mesh_resolution=(10, ),
+    bounding_box="dagmc.h5m",
+)
+tally2 = odw.MeshTally2D(
+    tally_type="neutron_effective_dose",
+    plane="yz",
+    mesh_resolution=(10, 5),
+    bounding_box="dagmc.h5m",
+)
+tally3 = odw.MeshTally2D(
+    tally_type="neutron_effective_dose",
+    plane="xz",
+    mesh_resolution=(10, 5),
+    bounding_box="dagmc.h5m",
+)
+tally4 = odw.MeshTally2D(
+    tally_type="neutron_effective_dose",
+    plane="xy",
+    mesh_resolution=(10, ),
+    bounding_box="dagmc.h5m",
+)
+tally5 = odw.MeshTally2D(
+    tally_type="neutron_effective_dose",
+    plane="yz",
+    mesh_resolution=(10, 5),
+    bounding_box="dagmc.h5m",
+)
+tally6 = odw.MeshTally2D(
+    tally_type="neutron_effective_dose",
+    plane="xz",
+    mesh_resolution=(10, 5),
+    bounding_box="dagmc.h5m",
+)
+
+# tally2 = odw.MeshTally3D(
+#     mesh_resolution=(100, 100, 100),
+#     bounding_box="dagmc.h5m",
+#     tally_type="neutron_effective_dose",
+# )
+
+tallies = openmc.Tallies(
+    [
+        tally1,
+        tally2,
+        tally3,
+        tally4,
+        tally5,
+        tally6,
+    ]
+)
+
+settings = odw.FusionSettings()
+settings.batches = 2
+settings.particles = 1000
+# 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(
+    materials=materials, geometry=geometry, settings=settings, tallies=tallies
+)
+statepoint_file = my_model.run()

examples/plot_regular_mesh_dose_tally.py

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+import regular_mesh_plotter as rmp
+import openmc
+
+# loads in the statepoint file containing tallies
+statepoint = openmc.StatePoint(filepath="statepoint.2.h5")
+
+# gets one tally from the available tallies
+my_tally = statepoint.get_tally(name="2_neutron_effective_dose")
+
+# creates a plot of the mesh tally
+my_plot = rmp.plot_regular_mesh_dose_tally(
+    tally=my_tally,  # the openmc tally object to plot, must be a 2d mesh tally
+    filename='plot_regular_mesh_dose_tally.png',  # the filename of the picture file saved
+    scale=None,  # LogNorm(),
+    vmin=None,
+    label="",
+    x_label="X [cm]",
+    y_label="Y [cm]",
+    rotate_plot = 0,
+    required_units='picosievert cm **2 / simulated_particle',
+    source_strength = None,
+)
+
+# displays the plot
+my_plot.show()

examples/plot_regular_mesh_dose_tally_with_geometry.py

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+import regular_mesh_plotter as rmp
+import openmc
+
+# loads in the statepoint file containing tallies
+statepoint = openmc.StatePoint(filepath="statepoint.2.h5")
+
+# gets one tally from the available tallies
+my_tally = statepoint.get_tally(name="2_neutron_effective_dose")
+
+# creates a plot of the mesh
+my_plot = rmp.plot_regular_mesh_dose_tally_with_geometry(
+    tally=my_tally,
+    dagmc_file_or_trimesh_object="dagmc.h5m",
+    filename = 'plot_regular_mesh_dose_tally_with_geometry.png',
+    scale=None,  # LogNorm(),
+    vmin=None,
+    label="",
+    x_label="X [cm]",
+    y_label="Y [cm]",
+    plane_origin = None,  # this could be skipped as it defaults to None, which uses the center of the mesh
+    plane_normal = [0, 0, 1],
+    rotate_mesh = 0,
+    rotate_geometry = 0,
+    required_units='picosievert cm **2 / simulated_particle',
+    source_strength = None,
+):
+
+my_plot.show()

examples/plot_regular_mesh_tally.py

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+
+import regular_mesh_plotter as rmp
+import openmc
+
+# loads in the statepoint file containing tallies
+statepoint = openmc.StatePoint(filepath="statepoint.2.h5")
+
+# gets one tally from the available tallies
+my_tally = statepoint.get_tally(name="2_neutron_effective_dose")
+
+# creates a plot of the mesh
+plot_regular_mesh_tally(
+    tally,
+    filename: Optional[str] = None,
+    scale=None,  # LogNorm(),
+    vmin=None,
+    label="",
+    base_plt=None,
+    x_label="X [cm]",
+    y_label="Y [cm]",
+    rotate_plot: float = 0,
+    required_units: str = None,
+    source_strength: float = None,

examples/plot_regular_mesh_tally_with_geometry.py

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+
+import regular_mesh_plotter as rmp
+import openmc
+
+# loads in the statepoint file containing tallies
+statepoint = openmc.StatePoint(filepath="statepoint.2.h5")
+
+# gets one tally from the available tallies
+my_tally = statepoint.get_tally(name="2_neutron_effective_dose")
+
+# creates a plot of the mesh
+my_plot = rmp.plot_regular_mesh_tally_with_geometry(
+    tally=my_tally,
+    dagmc_file_or_trimesh_object='dagmc.h5m',
+    std_dev_or_tally_value='tally_value',
+    filename='plot_regular_mesh_tally_with_geometry.png',
+    scale=None,  # LogNorm(),
+    vmin=None,
+    label="",
+    x_label="X [cm]",
+    y_label="Y [cm]",
+    plane_origin = [0,0,0],
+    plane_normal = [0, 0, 1],
+    rotate_mesh = 0,
+    rotate_geometry = 0,
+    required_units=None,
+    source_strength = None
+)
+
+my_plot.show()

examples/plot_regular_mesh_values.py

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+
+import regular_mesh_plotter as rmp
+import openmc
+
+# loads in the statepoint file containing tallies
+statepoint = openmc.StatePoint(filepath="statepoint.2.h5")
+
+# gets one tally from the available tallies
+my_tally = statepoint.get_tally(name="2_neutron_effective_dose")
+
+# creates a plot of the mesh
+plot_regular_mesh_values(
+    values: np.ndarray,
+    filename: Optional[str] = None,
+    scale=None,  # LogNorm(),
+    vmin=None,
+    label="",
+    base_plt=None,
+    extent=None,
+    x_label="X [cm]",
+    y_label="Y [cm]",
+    rotate_plot: float = 0,
+)

examples/plot_regular_mesh_values_with_geometry.py

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+
+import regular_mesh_plotter as rmp
+import openmc
+
+# loads in the statepoint file containing tallies
+statepoint = openmc.StatePoint(filepath="statepoint.2.h5")
+
+# gets one tally from the available tallies
+my_tally = statepoint.get_tally(name="2_neutron_effective_dose")
+
+
+
+def plot_regular_mesh_values_with_geometry(
+    values: np.ndarray,
+    dagmc_file_or_trimesh_object,
+    filename: Optional[str] = None,
+    scale=None,  # LogNorm(),
+    vmin=None,
+    label="",
+    extent=None,
+    x_label="X [cm]",
+    y_label="Y [cm]",
+    plane_origin: List[float] = None,
+    plane_normal: List[float] = [0, 0, 1],
+    rotate_mesh: float = 0,
+    rotate_geometry: float = 0,
+):