Merge pull request #280 from fusion-energy/adding_unstructured_mesh_r…

…2s_shut_down_dose_rate

adding shut down dose rate simulation using a DAGMC geometry and an unstructured mesh

tasks/task_18_CAD_shut_down_dose_rate/1_unstrucutred_mesh_R2S_shut_down_dose_rate.py

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+from cad_to_dagmc import CadToDagmc
+import cadquery as cq
+import openmc
+from matplotlib.colors import LogNorm
+import openmc.deplete
+
+# makes a CAD geometry to use for the neutronics geometry
+s = cq.Workplane("XY")
+sPnts = [
+    (2.75, 1.5),
+    (2.5, 1.75),
+    (2.0, 1.5),
+    (1.5, 1.0),
+    (1.0, 1.25),
+    (0.5, 1.0),
+    (0, 1.0),
+]
+r = s.lineTo(3.0, 0).lineTo(3.0, 1.0).spline(sPnts, includeCurrent=True).close()
+result = r.extrude(0.5)
+
+my_model = CadToDagmc()
+
+my_model.add_cadquery_object(result)
+
+# this makes the tet mesh used for the unstructured mesh tally which is overlaid on the geometry
+my_model.export_unstructured_mesh_file(filename="umesh.h5m", max_mesh_size=1, min_mesh_size=0.1)
+
+# this makes the surface mesh used for the material volume
+my_model.export_dagmc_h5m_file(filename="dagmc.h5m", max_mesh_size=1, min_mesh_size=0.1, material_tags=['mat1'])
+
+# the unstructured mesh to overlay on the DAGMC geometry
+umesh = openmc.UnstructuredMesh("umesh.h5m", library="moab")
+
+# filters for use inf the tally
+mesh_filter = openmc.MeshFilter(umesh)
+neutron_particle_filter = openmc.ParticleFilter(['neutron'])
+energy_filter = openmc.EnergyFilter.from_group_structure('CCFE-709')
+
+# sets up a neutron spectra tally on the unstructured mesh 
+tally = openmc.Tally(name="unstructured_mesh_neutron_spectra_tally")
+tally.filters = [mesh_filter, energy_filter, neutron_particle_filter]
+tally.scores = ["flux"]
+tally.estimator = "tracklength"
+my_tallies = openmc.Tallies([tally])
+
+my_material = openmc.Material(name='mat1')
+my_material.add_nuclide("H1", 1, percent_type="ao")
+my_material.set_density("g/cm3", 0.001)
+my_materials = openmc.Materials([my_material])
+
+universe = openmc.DAGMCUniverse("dagmc.h5m").bounded_universe()
+my_geometry = openmc.Geometry(universe)
+
+my_settings = openmc.Settings()
+my_settings.batches = 10
+my_settings.inactive = 0
+my_settings.particles = 5000
+my_settings.run_mode = "fixed source"
+
+# Create a DT point source
+my_source = openmc.IndependentSource()
+my_source.space = openmc.stats.Point((0, 0, 0))
+my_source.angle = openmc.stats.Isotropic()
+my_source.energy = openmc.stats.Discrete([14e6], [1])
+my_settings.source = my_source
+
+model = openmc.model.Model(my_geometry, my_materials, my_settings, my_tallies)
+sp_filename = model.run()
+
+
+sp = openmc.StatePoint(sp_filename)
+
+tally_result = sp.get_tally(name="unstrucutred_mesh_tally")
+
+# normally with regular meshes I would get the mesh from the tally
+# but with unstrucutred meshes the tally does not contain the mesh
+# however we can get it from the statepoint file
+# umesh = tally_result.find_filter(openmc.MeshFilter)
+umesh_from_sp = sp.meshes[1]
+
+# these trigger internal code in the mesh object so that its centroids and volumes become known.
+# centroids and volumes are needed for the get_values and write_data_to_vtk steps
+centroids = umesh_from_sp.centroids
+mesh_vols = umesh_from_sp.volumes
+
+flux_in_each_group_for_each_voxel = tally_result.get_values(scores=["flux"], value="mean")
+
+all_sources = []
+
+for flux_in_each_group, mesh_vol in zip(flux_in_each_group_for_each_voxel, mesh_vols):
+    micro_xs = openmc.deplete.MicroXS.from_multigroup_flux(
+        energies='CCFE-709',
+        multigroup_flux=flux_in_each_group,
+        temperature=294, # endf 8.0 has ['1200K', '2500K', '250K', '294K', '600K', '900K']
+        chain_file= openmc.config['chain_file'],
+        nuclides=my_material.get_nuclides()
+    )
+
+    # constructing the operator, note we pass in the flux and micro xs
+    operator = openmc.deplete.IndependentOperator(
+        materials=my_materials,
+        fluxes=[sum(flux_in_each_group)*my_material.volume],  # Flux in each group in [n-cm/src] for each domain
+        micros=[micro_xs],
+        reduce_chain=True,  # reduced to only the isotopes present in depletable materials and their possible progeny
+        reduce_chain_level=5,
+        normalization_mode="source-rate"
+    )
+
+    integrator = openmc.deplete.PredictorIntegrator(
+        operator=operator,
+        timesteps=[5, 60, 60],
+        source_rates=[1e20, 0 , 0], # a 5 second pulse of neutrons followed by 120 seconds of decay
+        timestep_units='s'
+    )
+
+    integrator.integrate()
+
+    results = openmc.deplete.Results.from_hdf5("depletion_results.h5")
+    last_time_step=result[-1]
+    activated_material = last_time_step.get_material(my_material.id)
+
+    activated_material.volume = mesh_vol
+
+    energy = activated_material.get_decay_photon_energy(
+        clip_tolerance = 1e-6,  # cuts out a small fraction of the very low energy (and hence negligible dose contribution) photons
+        units = 'Bq',
+    )
+    strength = energy.integral()
+
+
+mesh_source = openmc.MeshSource(
+    mesh=umesh_from_sp,
+    sources=all_sources,
+)
+
+my_gamma_settings = openmc.Settings()
+my_gamma_settings.run_mode = "fixed source"
+my_gamma_settings.batches = 100
+my_gamma_settings.particles = 100000
+my_gamma_settings.source = mesh_source
+
+energies, pSv_cm2 = openmc.data.dose_coefficients(particle="photon", geometry="AP")
+dose_filter = openmc.EnergyFunctionFilter(
+    energies, pSv_cm2, interpolation="cubic"  # interpolation method recommended by ICRP
+)
+
+regularmesh = openmc.RegularMesh().from_domain(
+    my_geometry,
+    dimension=[10, 10, 10],  # 10 voxels in each axis direction (x, y, z)
+)
+
+particle_filter = openmc.ParticleFilter(["photon"])
+mesh_filter = openmc.MeshFilter(regularmesh)
+dose_tally = openmc.Tally()
+dose_tally.filters = [mesh_filter, dose_filter, particle_filter]
+dose_tally.scores = ["flux"]
+dose_tally.name = "photon_dose_on_mesh"
+tallies = openmc.Tallies([dose_tally])
+
+
+model_gamma = openmc.Model(my_geometry, my_materials, my_gamma_settings, tallies)
+
+model_gamma.run(cwd="photons")
+
+# You may wish to plot the dose tally on a mesh, this package makes it easy to include the geometry with the mesh tally
+from openmc_regular_mesh_plotter import plot_mesh_tally
+with openmc.StatePoint('photons/statepoint.100.h5') as statepoint:
+    photon_tally = statepoint.get_tally(name="photon_dose_on_mesh")
+
+    # normalising this tally is a little different to other examples as the source strength has been using units of photons per second.
+    # tally.mean is in units of pSv-cm3/source photon.
+    # as source strength is in photons_per_second this changes units to pSv-/second
+
+    # multiplication by pico_to_micro converts from (pico) pSv/s to (micro) uSv/s
+    # dividing by mesh voxel volume cancles out the cm3 units
+    # could do the mesh volume scaling on the plot and vtk functions but doing it here instead
+    pico_to_micro = 1e-6
+    seconds_to_hours = 60*60
+    scaling_factor = (seconds_to_hours * pico_to_micro) / regularmesh.volumes[0][0][0]
+
+    plot = plot_mesh_tally(
+            tally=photon_tally,
+            basis="xz",
+            # score='flux', # only one tally so can make use of default here
+            value="mean",
+            colorbar_kwargs={
+                'label': "Decay photon dose [µSv/h]",
+            },
+            norm=LogNorm(),
+            volume_normalization=False,  # this is done in the scaling_factor
+            scaling_factor=scaling_factor,
+        )
+    plot.figure.savefig(f'shut_down_dose_map_timestep.png')