Mesh plot problem

User Support
1

Hi everyone,

I met a strange problem when I was using mesh tally to present the neutron flux distribution. The geometry of this model is four UO2 assemblies array surrounded by water, and a 2D mesh tally was defined to cover the assemblies and water region. But in the results visualization process, the assemblies appeared at different areas with the geometry definition.

I tried to redefine the mesh plot and check the codes, but I can not find a clue to fix it. I’ll put the geometry plot, mesh plot and the tally setting files in the attachment. If anybody knows where is the problem, please let me know.

# Load the statepoint file
sp=openmc.StatePoint(‘statepoint.200.h5’)

tally=sp.get_tally(name=‘flux tally’)

flux=tally.get_slice(scores=[‘flux’])
thermal_flux=flux.get_slice(filters=[openmc.EnergyFilter], filter_bins=[((0, 0.625),)])

thermal_flux.mean.shape=(173,173)
plt.figure()
plt.imshow(thermal_flux.mean, interpolation=‘none’)
plt.colorbar()

Above is part of the results visualization code. If there is something error here, please let me know. Thanks for your help in advance.

Best wishes,

Yiming

PWR_assembly.png
PWR_assembly.png3000×3000 11 KB

tallies.xml (533 Bytes)

tallies.out (733 KB)

thermal_neutron_flux.png

2

Hi Yiming,

The imshow function in matplotlib default to putting the (0,0) value in the upper-left corner rather than the lower-left. If you pass the keyword argument origin=‘lower’ to imshow, it will vertically flip the image to match what the data that OpenMC gives you.

Best regards,
Paul