Geometry plotting not working as expected

User Support
1

Hello,

I am trying to plot my cylindrical geometry for the simple input below. I am following the instructions here. When I open the vti file in paraview, I only see a cube of side length 50 instead of my cylindrical geometry. I am running the file with python, then running openmc --plot, then running openmc-voxel-to-vtk plot_1.h5 What am I doing wrong? Thanks!

import numpy as np

import openmc
import openmc.mgxs as mgxs

# create a model object to tie geometry, materials, settings, and tallies together
model = openmc.Model()

# Instantiate a Material and register the Nuclides
inf_medium = openmc.Material(name=‘moderator’)
inf_medium.set_density(‘g/cc’, 5.)
inf_medium.add_nuclide(‘C12’, 0.01450188)
inf_medium.add_nuclide(‘U235’, 0.000114142)
inf_medium.add_nuclide(‘U238’, 0.006886019)

# Instantiate a Materials collection and export to XML
model.materials = openmc.Materials([inf_medium])
model.materials.export_to_xml()

# Lengths in cm
thickness = 1.27
inner_radius = 36.83
inner_height = 251.46

outer_radius = inner_radius + thickness
total_height = 2 * thickness + inner_height

# Instantiate boundary Planes
#inner_tube = openmc.ZCylinder(r=inner_radius)
outer_tube = openmc.ZCylinder(r=outer_radius, boundary_type=‘vacuum’)

bottom = openmc.ZPlane(z0=0.0, boundary_type=‘vacuum’)
#cavity_start = openmc.ZPlane(z0=thickness)
#cavity_end = openmc.ZPlane(z0=thickness + inner_height)
top = openmc.ZPlane(z0=total_height, boundary_type=‘vacuum’)

# Instantiate a Cell
cell = openmc.Cell(cell_id=1, name=‘cell’)

# Register bounding Surfaces with the Cell
cell.region = +bottom & -top & -outer_tube

# Fill the Cell with the Material
cell.fill = inf_medium

# Create root universe
root_universe = openmc.Universe(name=‘root universe’, cells=[cell])

# Create Geometry and set root Universe
model.geometry = openmc.Geometry(root_universe)
model.geometry.export_to_xml()

# OpenMC simulation parameters
batches = 50
inactive = 10
particles = 2500
#generations_per_batch = 1

# Instantiate a Settings object
settings = openmc.Settings()
settings.batches = batches
settings.inactive = inactive
settings.particles = particles
#settings.generations_per_batch = generations_per_batch
settings.output = {‘tallies’: True}

# Create an initial uniform spatial source distribution over fissionable zones
r_dist = openmc.stats.Uniform(0, inner_radius)
phi_dist = openmc.stats.Uniform(0, 2*np.pi)
z_dist = openmc.stats.Uniform(thickness, thickness + inner_height)
uniform_dist = openmc.stats.CylindricalIndependent(r_dist, phi_dist, z_dist)
settings.source = openmc.Source(space=uniform_dist)

model.settings = settings
settings.export_to_xml()

# Instantiate a 2-group EnergyGroups object
groups = mgxs.EnergyGroups(group_edges=np.array([0., 0.625, 20.0e6]))

# Instantiate a few different sections
total = mgxs.TotalXS(domain=cell, energy_groups=groups)
absorption = mgxs.AbsorptionXS(domain=cell, energy_groups=groups)
scattering = mgxs.ScatterXS(domain=cell, energy_groups=groups)

# Note that if we wanted to incorporate neutron multiplication in the
# scattering cross section we would write the previous line as:
# scattering = mgxs.ScatterXS(domain=cell, energy_groups=groups, nu=True)

# Instantiate an empty Tallies object
tallies = openmc.Tallies()

# Add total tallies to the tallies file
tallies += total.tallies.values()

# Add absorption tallies to the tallies file
tallies += absorption.tallies.values()

# Add scattering tallies to the tallies file
tallies += scattering.tallies.values()

model.tallies = tallies

tallies.export_to_xml()

# Geometry plot
vox_plot = openmc.Plot()
vox_plot.type = ‘voxel’
vox_plot.width = (100., 100., 100.)
vox_plot.pixels = (400, 400, 400)

plots = openmc.Plots([vox_plot])
plots.export_to_xml()

2

@pbehne Paraview automatically shows all geometry cells specified in OpenMC, including vacuum cells. The IDs of these cells might start from negative values. To visualize only your desired cylindrical geometry, you can use the Threshold filter in Paraview. This filter allows you to select a specific range of cell IDs to display. By setting the threshold to exclude the vacuum cells, you should be able to see your cylindrical geometry correctly.

image
image1644×883 125 KB

3

Thank you, @Pranto! Now that I can see my cylindrical geometry, I have a follow-up question. In the input file, I specified a cylinder of radius ~38 cm and height ~254 cm. However, it looks like I am getting a cylinder with the correct radius, but only half the expected height. When I look at a slice through entire geometry, I see the correct height, but the cylinder ends halfway down:

image
image896×2192 134 KB

Do you know what is going on here?

4

The issue you are facing with the height of the cylinder is due to the default origin used in the plot. In OpenMC, the width of the voxels gets halved in each direction from the origin. As a result, when you visualize the cylindrical geometry with the default origin (0, 0, 0), it leads to the widths of the voxels being divided in half, resulting in a cylinder with only half the expected height. To get the full height, you can change the origin of the plot to (0, 0, H/2), where H is the height of the cylinder. This will ensure that the cylinder is centered correctly and spans the entire height.

A modified code snippet below that incorporates this suggested change:

vox_plot = openmc.Plot()
vox_plot.type = 'voxel'
vox_plot.origin = (0., 0., total_height/2)
vox_plot.width = (outer_radius*2, outer_radius*2, total_height)

I hope this helps! Please let me know if you have any questions.

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