Plotting flux spectrum after depletion in openmc

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1

Here is how I can view the neutron flux at the output. There is an application in Jupyter notebook that is made like this. @paulromano @Shimwell @pshriwise

sp_path = model.run(output=False)

How can I change this for the following integrator?

# Instantiate a Settings object, set all runtime parameters, and export to XML
settings = openmc.Settings()
settings.batches = 500
settings.inactive = 20
settings.particles = 500

# Create an initial uniform spatial source distribution over fissionable zones
settings.source = openmc.source.Source(space=openmc.stats.Point())

tally = dict()
tally['heating'] = openmc.Tally(name="heating")
tally['heating'].scores.append('heating-local')
e_min, e_max = 1e-5, 20e6
groups = 500
energies = np.logspace(log10(e_min), log10(e_max), groups + 1)
energy_filter = openmc.EnergyFilter(energies)
particle_filter = openmc.ParticleFilter(['neutron'])
cell_filter = openmc.MaterialFilter([uo2])
tally['flux']= openmc.Tally(name="flux")
tally['flux'].filters = [energy_filter, particle_filter] 
tally['flux'].scores = ['flux']
mesh = openmc.RegularMesh()
mesh.dimension = [500, 500, 1]
mesh.lower_left = [-100, -100, 50]
mesh.upper_right = [100, 100, 200]
mesh_filter = openmc.MeshFilter(mesh)
tally['mesh'] = openmc.Tally(name="Mesh")
tally['mesh'].scores = ['flux','absorption','fission']
tally['mesh'].filters = [mesh_filter]
tally['mesh'].filters.append(particle_filter)
tallies = openmc.Tallies(tally.values())

###############################################################################
#                   Initialize and run depletion calculation
###############################################################################

with open('/depletion/serpent_fissq.json', 'r') as f:
    serpent_fission_q = json.load(f)

model = openmc.Model(geometry=geom, settings=settings, tallies=tallies)
model.export_to_xml()
chain_file = '/depletion-comparison/data/depletion/chain.xml'
op = openmc.deplete.Operator(model, chain_file,
    fission_q=serpent_fission_q,
    fission_yield_mode="average")

# cumulative steps in MWd/kg
burnup_cum = np.array([0.1, 0.5, 1.0, 2.0, 3.0, 4.0, 5.0, 6.0, 7.0, 8.0, 9.0, 10.0, 11.0, 12.5, 15.0, 17.5, 20.0, 22.5, 25.0, 27.5, 30.0, 32.5, 35.0, 37.5, 40.0, 42.5, 45.0])
burnup = np.diff(burnup_cum, prepend=0.0)
power = 131040  

integrator = openmc.deplete.PredictorIntegrator(op, burnup, power, timestep_units='MWd/kg')
integrator.integrate()
2

hi prof electron, the specified tally will be reported on the statepoint file of each depletion steps, so to read your tally output like from the second depletion steps, you could read your statepoint_n2.h5, and so on. so as long as your tally has been added to the model for your transport operator being used at the integrator, then it will be good.
model = openmc.Model(geometry, materials, settings, tallies)
op = openmc.deplete.Operator(model, …
you can try it with small particle history just to make sure that everything is as you want.

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