# Building Geometry and Plotting Issues **URL:** \ **Category:** User Support\ **Created:** [January 16, 2025, 2:13am UTC](https://openmc.discourse.group/t/building-geometry-and-plotting-issues/5145 "2025-01-16T02:13:55Z")\ **Posts on this page:** 3\ **Page:** 1
**Author:** ![AbdullahAlharbi](https://avatars.discourse-cdn.com/v4/letter/a/45deac/32.png) [@AbdullahAlharbi](https://openmc.discourse.group/u/AbdullahAlharbi)\ **Post date:** [January 16, 2025, 2:13am UTC](https://openmc.discourse.group/t/building-geometry-and-plotting-issues/5145/1 "2025-01-16T02:13:55Z")
Hello, I just installed OpenMC a few days ago and I am working on a big project using it, I’m running into a few issues and I would benefit a lot from clarifications: 1- What is the best IDE to write and execute OpenMC simulation, I am used to PyCharm, is it a good IDE for OpenMC ?, I see most people are using Jupiter notebook. 2- When I try to plot a piece of geometry, a cell or a universe, I always get this error: ===\> fuel\_assembly\_cell.plot() ===\> OSError: \path\. so: cannot open shared object file: No such file or directory 3- after I build the geometry, it gets exported to xml just fine, but when I try to plot it through: ===\> plot = mc.Plot() ===\>plots = openmc.Plots([plot]) ===\>plots.export\_to\_xml() ===\>openmc.plot\_geometry() or run a simulation through openmc.run(), I get the following error: Simulating batch 1 ERROR: Particle -1 left lattice 2, but it has no outer definition. ERROR: Particle -1 left lattice 2, but it has no outer definition. Traceback (most recent call last): File “/mnt/c/openMC/UFTR/test.py”, line 126, in mc.run() File “/home/eastdusty/openmc\_env/lib/python3.11/site-packages/openmc/executor.py”, line 314, in run \_run(args, output, cwd) File “/home/eastdusty/openmc\_env/lib/python3.11/site-packages/openmc/executor.py”, line 125, in \_run raise RuntimeError(error\_msg) RuntimeError: Particle -1 left lattice 2, but it has no outer definition. ERROR: Particle -1 left lattice 2, but it has no outer definition. Since I can not include my code in a pdf file, I will include it below as plain text, and please let me know if there is an issue with geometry definition or program installation, I am new to OpenMC and trying to learn, and has been stuck for about two days now with this issue. import os import matplotlib.pyplot as plt os.environ[‘OPENMC\_CROSS\_SECTIONS’] = ‘/home/eastdusty/openmc\_env/share/data/endfb80/endfb-viii.0-hdf5/cross\_sections.xml’ import openmc as mc import openmc.plotter as plotter U3Si2Al = mc.Material(name=‘U3Si2Al’) U3Si2Al.add\_nuclide(‘U238’, 3 \* 0.8025) U3Si2Al.add\_nuclide(‘U235’, 3 \* 0.1975) U3Si2Al.add\_nuclide(‘Si28’, 2 \* 0.922545) U3Si2Al.add\_nuclide(‘Si29’, 2 \* 0.04672) U3Si2Al.add\_nuclide(‘Si30’, 2 \* 0.030735) U3Si2Al.add\_nuclide(‘Al27’, 1) U3Si2Al.set\_density(‘g/cm3’, 5.55) Al6061 = mc.Material(name=‘Al6061’) Al6061.add\_nuclide(‘Al27’, 1) Al6061.set\_density(‘g/cm3’, 2.7) water = mc.Material(name=‘Water’) water.add\_nuclide(‘H1’, 0.111877) water.add\_nuclide(‘H2’, 0.000013) water.add\_nuclide(‘O16’, 0.88811) water.set\_density(‘g/cm3’, 1.0) materials = mc.Materials([U3Si2Al, Al6061, water]) materials.export\_to\_xml() fuel\_top = mc.ZPlane(z0=30) fuel\_bottom = mc.ZPlane(z0=-30) fuel\_right = mc.XPlane(x0=2.98) fuel\_left = mc.XPlane(x0=-2.98) fuel\_front = mc.YPlane(y0=0.0255) fuel\_back = mc.YPlane(y0=-0.0255) fuel\_region = +fuel\_bottom & -fuel\_top & +fuel\_left & -fuel\_right & -fuel\_front & +fuel\_back fuel\_cell = mc.Cell(region=fuel\_region, fill=U3Si2Al) clad\_top = mc.ZPlane(z0=32.55, boundary\_type = ‘vacuum’) clad\_bottom = mc.ZPlane(z0=-32.55, boundary\_type = ‘vacuum’) clad\_right = mc.XPlane(x0=3.615, boundary\_type = ‘vacuum’) clad\_left = mc.XPlane(x0=-3.615, boundary\_type = ‘vacuum’) clad\_front = mc.YPlane(y0=0.0636) clad\_back = mc.YPlane(y0=-0.0636) clad\_region = +clad\_bottom & -clad\_top & +clad\_left & -clad\_right & -clad\_front & +clad\_back & ~fuel\_region clad\_cell = mc.Cell(region=clad\_region, fill=Al6061) coolant\_top = mc.ZPlane(z0=32.55) coolant\_bottom = mc.ZPlane(z0=-32.55) coolant\_front = mc.YPlane(y0=0.2046) coolant\_back = mc.YPlane(y0=-0.2046) coolant\_left = clad\_left coolant\_right = clad\_right coolant\_region = -coolant\_top & + coolant\_bottom & -coolant\_front & +coolant\_back & +coolant\_left & -coolant\_right & ~clad\_region & ~ fuel\_region coolant = mc.Cell(region=coolant\_region, fill=water) fuel\_plate = mc.Universe(cells=[fuel\_cell, clad\_cell, coolant]) fuel\_assembly = mc.RectLattice(name = ‘fuel assembly’) fuel\_assembly.lower\_left = (0,-2.8644) fuel\_assembly.pitch = (1, 0.2046) fuel\_assembly.universes = [[fuel\_plate, fuel\_plate, fuel\_plate, fuel\_plate, fuel\_plate, fuel\_plate, fuel\_plate, fuel\_plate, fuel\_plate, fuel\_plate, fuel\_plate, fuel\_plate]] fuel\_assembly\_cell = mc.Cell() fuel\_assembly\_cell.fill = fuel\_assembly fuel\_assembly\_universe = mc.Universe(cells=[fuel\_assembly\_cell]) front\_boundary = mc.YPlane(y0=2.8644, boundary\_type = ‘vacuum’) back\_boundary = mc.YPlane(y0=-2.8644, boundary\_type = ‘vacuum’) geometry = mc.Geometry(fuel\_assembly\_universe) geometry.export\_to\_xml() color = {fuel\_cell: ‘red’, clad\_cell: ‘green’, coolant: ‘blue’} plot = mc.Plot() plot.basis = ‘xy’ plot.origin = (0, 0, 0) plot.width = (7.25, 5.73) plot.pixels = (500, 500) plot.colors = color plot.filename = ‘geometry\_plot’ plot.filetype = ‘png’ plots = mc.Plots([plot]) plots.export\_to\_xml() mc.plot\_geometry() settings = mc.Settings() settings.run\_mode = ‘fixed source’ source = mc.Source() source.space = mc.stats.Point((0.0, 0.0, 0.0)) source.angle = mc.stats.Isotropic() source.energy = mc.stats.Discrete([1.0e6], [1.0]) settings.source = source settings.batches = 50 settings.particles = 1000 settings.inactive = 10 settings.export\_to\_xml() mc.run() ---
**Author:** ![skyawlker](https://avatars.discourse-cdn.com/v4/letter/s/a587f6/32.png) [@skyawlker](https://openmc.discourse.group/u/skyawlker)\ **Post date:** [January 16, 2025, 1:35pm UTC](https://openmc.discourse.group/t/building-geometry-and-plotting-issues/5145/2 "2025-01-16T13:35:58Z")
> [@AbdullahAlharbi](#): > > 2- When I try to plot a piece of geometry, a cell or a universe, I always get this error: > ===\> fuel\_assembly\_cell.plot() > ===\> OSError: \path. so: cannot open shared object file: No such file or directory fuel\_assembly\_cell.plot() is worked when we use Jupyter notebook. If you don’t use it, you can use `openmc -p` in terminal to plot. > [@AbdullahAlharbi](#): > > 3- after I build the geometry, it gets exported to xml just fine, but when I try to plot it through: > ===\> plot = mc.Plot() > ===\>plots = openmc.Plots([plot]) > ===\>plots.export\_to\_xml() > ===\>openmc.plot\_geometry() > or run a simulation through openmc.run(), I get the following error: > Simulating batch 1 You should define the outer when the lattice was used. Or this error will happen when the region of the cell small than the lattice which filld with. ---
**Author:** ![AbdullahAlharbi](https://avatars.discourse-cdn.com/v4/letter/a/45deac/32.png) [@AbdullahAlharbi](https://openmc.discourse.group/u/AbdullahAlharbi)\ **Post date:** [January 20, 2025, 4:12am UTC](https://openmc.discourse.group/t/building-geometry-and-plotting-issues/5145/4 "2025-01-20T04:12:40Z")
Yes that worked, after I defined the outer boundary everything was fine.