# Pincell model error while attempting to calulate keff **URL:** \ **Category:** User Support\ **Created:** [August 13, 2024, 11:11pm UTC](https://openmc.discourse.group/t/pincell-model-error-while-attempting-to-calulate-keff/4682 "2024-08-13T23:11:37Z")\ **Posts on this page:** 4\ **Page:** 1
**Author:** ![AsukuA](https://avatars.discourse-cdn.com/v4/letter/a/49beb7/32.png) [@AsukuA](https://openmc.discourse.group/u/AsukuA)\ **Post date:** [August 13, 2024, 11:11pm UTC](https://openmc.discourse.group/t/pincell-model-error-while-attempting-to-calulate-keff/4682/1 "2024-08-13T23:11:37Z")
Hello, I am relatively new to openmc. I have created a test pincell model. But i tried to account for the height of the pincell and this led to termination of the process with the following error message. #error message WARNING: After particle 251 crossed surface 511 it could not be located in any cell and it did not leak. WARNING: After particle 126 crossed surface 511 it could not be located in any cell and it did not leak. WARNING: After particle 626 crossed surface 511 it could not be located in any WARNING: After particle 501 crossed surface 510 it could not be located in any… import openmc import numpy as np import matplotlib.pyplot as plt import matplotlib.image as mpimg openmc.config.cross\_sections = ‘/openmc/cross-sections/cross\_sections.xml’ # Define Materials materials = openmc.Materials() # Active Fuel Material uo2 = openmc.Material(name=‘uo2’) uo2.set\_density(‘g/cm3’, 10.66) uo2.add\_nuclide(‘U235’, 0.114313381) uo2.add\_nuclide(‘U238’, 0.761063318) uo2.add\_nuclide(‘U234’, 0.001758667) uo2.add\_nuclide(‘U236’, 0.002198334) uo2.add\_nuclide(‘O16’, 0.11839) uo2.add\_element(‘Al’, 0.000250000000) uo2.add\_element(‘C’, 0.000100000000) uo2.add\_element(‘Ca’, 0.000100000000) uo2.add\_element(‘Mg’, 0.000100000000) uo2.add\_element(‘Cl’, 0.000025000000) uo2.add\_element(‘Cr’, 0.000250000000) uo2.add\_element(‘Co’, 0.000100000000) uo2.add\_element(‘F’, 0.000015000000) uo2.add\_element(‘H’, 0.000001300000) uo2.add\_element(‘Fe’, 0.000500000000) uo2.add\_element(‘Ni’, 0.000250000000) uo2.add\_element(‘N’, 0.000075000000) uo2.add\_element(‘Si’, 0.000500000000) uo2.add\_element(‘Th’, 0.000010000000) materials.append(uo2) # Fuel Clad/Zircaloy-4 Material zr4 = openmc.Material(name=‘Zr4’) zr4.set\_density(‘g/cm3’, 6.5) zr4.add\_element(‘Zr’, 0.9792948) zr4.add\_element(‘Fe’, 0.0024000) zr4.add\_element(‘Cr’, 0.0013000) zr4.add\_element(‘Sn’, 0.0169999) zr4.add\_element(‘B’, 0.000002971) zr4.add\_element(‘B’, 0.0000006713) zr4.add\_element(‘He’, 0.000001652) materials.append(zr4) # Moderator Material h2o = openmc.Material(name=‘h20’) h2o.set\_density(‘g/cm3’, 0.99825) h2o.add\_nuclide(‘H1’, 0.111898) h2o.add\_nuclide(‘O16’, 0.888102) h2o.add\_s\_alpha\_beta(‘c\_H\_in\_H2O’) materials.append(h2o) # Export materials to XML materials.export\_to\_xml() # Define Cylindrical Surfaces fuel\_outer\_radius = openmc.ZCylinder(r=0.215) clad\_inner\_radius = openmc.ZCylinder(r=0.215) clad\_outer\_radius = openmc.ZCylinder(r=0.275) water\_inner\_radius = openmc.ZCylinder(r=0.6192) # Define Z-Planes for height z\_min = -11.5 # cm z\_max = 11.5 # cm z\_top = openmc.ZPlane(z\_max) z\_bottom = openmc.ZPlane(z\_min) # Define Regions fuel\_region = -fuel\_outer\_radius & +z\_bottom & -z\_top clad\_region = +clad\_inner\_radius & -clad\_outer\_radius & +z\_bottom & -z\_top # Create Cells fuel = openmc.Cell(name=‘fuel’) fuel.fill = uo2 fuel.region = fuel\_region clad = openmc.Cell(name=‘clad’) clad.fill = zr4 clad.region = clad\_region pitch = 0.6192 \*2 left = openmc.XPlane(-pitch/2, boundary\_type=‘reflective’) right = openmc.XPlane(pitch/2, boundary\_type=‘reflective’) bottom = openmc.YPlane(-pitch/2, boundary\_type=‘reflective’) top = openmc.YPlane(pitch/2, boundary\_type=‘reflective’) # Update water\_region to use boundary conditions water\_region = +left & -right & +bottom & -top & +clad\_outer\_radius & +z\_bottom & -z\_top moderator = openmc.Cell(name=‘moderator’) moderator.fill = h2o moderator.region = water\_region # Define Root Universe root\_universe = openmc.Universe(cells=[fuel, clad, moderator]) # Define Geometry geometry = openmc.Geometry() geometry.root\_universe = root\_universe # Export geometry to XML geometry.export\_to\_xml() # Define Source source = openmc.Source() source.space = openmc.stats.Box( lower\_left=[-11.5, -11.5, -11.5], upper\_right=[11.5, 11.5, 11.5] ) source.strength = 1.0 # Normalized value or based on empirical data # Define Settings settings = openmc.Settings() settings.batches = 100 settings.inactive = 10 settings.particles = 1000 settings.run\_mode = ‘eigenvalue’ # Define Tally tally = openmc.Tally() tally.filters = [openmc.CellFilter([fuel.id, clad.id, moderator.id])] energy\_filter = openmc.EnergyFilter([1e-5, 0.625, 0.825, 20.0]) tally.filters.append(energy\_filter) tally.scores = [‘flux’] # Add Tally to Tallies Collection tallies = openmc.Tallies([tally]) tallies.export\_to\_xml() settings.export\_to\_xml() # Run the simulation openmc.run() please note that the code runs without any error when i remove the height variable that defines the height of the cylinder. please suggestions on how to resolve this will be appreciated. Thank you. ---
**Author:** ![wahidluthfi](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/wahidluthfi/32/3375_2.png) [@wahidluthfi](https://openmc.discourse.group/u/wahidluthfi)\ **Post date:** [August 14, 2024, 12:32am UTC](https://openmc.discourse.group/t/pincell-model-error-while-attempting-to-calulate-keff/4682/2 "2024-08-14T00:32:47Z")
Hi Asukua, welcome back to the community. As you said before, the problem might come from the reality that you forgot to add boundary conditions to your z plane so openmc thinks that this surface has transmission BC. Try to use reflective BC since 2×11.5 is quite short for a PWR fuel pin height, but if you use the full height of your reactor fuel pin, you can use vacuum BC. ```auto z_top = openmc.ZPlane(z_max, boundary_type=‘reflective’) z_bottom = openmc.ZPlane(z_min, boundary_type=‘reflective’) ``` Also, you can check your geometry by using openmc geometry debug, it will check the cell overlaps and more ```auto openmc.run(geometry_debug=True) ``` ---
**Author:** ![AsukuA](https://avatars.discourse-cdn.com/v4/letter/a/49beb7/32.png) [@AsukuA](https://openmc.discourse.group/u/AsukuA)\ **Post date:** [August 14, 2024, 1:28am UTC](https://openmc.discourse.group/t/pincell-model-error-while-attempting-to-calulate-keff/4682/3 "2024-08-14T01:28:53Z")
Thanks so much. It worked perfectly. The pincell wasnt intended to be that of a PWR. Its for actually small. Its for an MNSR, a tank-in-pool type reactor similar to SLOWPOKE reactors. ---
**Author:** ![wahidluthfi](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/wahidluthfi/32/3375_2.png) [@wahidluthfi](https://openmc.discourse.group/u/wahidluthfi)\ **Post date:** [August 14, 2024, 1:42am UTC](https://openmc.discourse.group/t/pincell-model-error-while-attempting-to-calulate-keff/4682/4 "2024-08-14T01:42:02Z")
okay, then the vacuum boundary condition might fit your requirements. glad to hear that.