# Must all universes share the same origin? **URL:** \ **Category:** User Support\ **Created:** [November 10, 2023, 11:22am UTC](https://openmc.discourse.group/t/must-all-universes-share-the-same-origin/3618 "2023-11-10T11:22:16Z")\ **Posts on this page:** 1\ **Page:** 1
**Author:** ![emilio.castro](https://avatars.discourse-cdn.com/v4/letter/e/b38774/32.png) [@emilio.castro](https://openmc.discourse.group/u/emilio.castro)\ **Post date:** [November 10, 2023, 11:22am UTC](https://openmc.discourse.group/t/must-all-universes-share-the-same-origin/3618/1 "2023-11-10T11:22:16Z")
Hello. I am trying to place an universe (consisting on a cylinder) inside a cell to place it in my problem universe at a different axial location. First, I define my isolated cylinder as follows, between z= -1.02555 and 1.02555, and it works: ```python #!/usr/bin/env python3 # -*- coding: utf-8 -*- """ Created on Mon Oct 9 10:45:00 2023 @author: ecasglez """ import openmc #################### # Define materials # #################### mat_1 = openmc.Material() mat_1.add_element('U', 1.0) mat_1.set_density('g/cm3', 10) mat_1.temperature = 293.15 mat_2 = openmc.Material() mat_2.add_element('Zr', 1.0) mat_2.set_density('g/cm3', 6) mat_2.temperature = 293.15 mats = openmc.Materials([mat_1, mat_2]) mats.export_to_xml() ################### # Define geometry # ################### case_A_cyl_in = openmc.ZCylinder(r=7.5050) case_A_cyl_out = openmc.ZCylinder(r=7.5438, boundary_type = 'reflective') case_A_z_in_top = openmc.ZPlane(1.00455) case_A_z_in_bottom = openmc.ZPlane(-1.00455) case_A_z_out_top = openmc.ZPlane(1.02555, boundary_type = 'reflective') case_A_z_out_bottom = openmc.ZPlane(-1.02555, boundary_type = 'reflective') case_A_in = openmc.Cell() case_A_in.fill = mat_1 case_A_in.region = -case_A_cyl_in & -case_A_z_in_top & +case_A_z_in_bottom case_A_out = openmc.Cell() case_A_out.fill = mat_2 case_A_out.region = -case_A_cyl_out & +case_A_z_out_bottom & -case_A_z_out_top & ~ case_A_in.region case_A_u = openmc.Universe(cells=[case_A_in, case_A_out]) geom = openmc.Geometry(root = case_A_u) geom.export_to_xml() ##################### # Source Definition # ##################### point = openmc.stats.Point((0, 0, 0)) src = openmc.Source(space=point) ############ # Settings # ############ settings = openmc.Settings() settings.source = src settings.batches = 100 settings.inactive = 20 settings.particles = 1000 settings.export_to_xml() ####### # Run # ####### openmc.run() ``` Now, I want to put this cylinder inside another cylinder taller that the previous one but with the same diameter. I put it in a cell with the same height of my original cylinder but using a different origin of coordinates. So the first cylinder is defined between z = -1.02555 and 1.02555, and placed then in the problem universe between Z = 0, 1.02555 \* 2, as follows: ```python #!/usr/bin/env python3 # -*- coding: utf-8 -*- """ Created on Mon Oct 9 10:45:00 2023 @author: ecasglez """ import openmc #################### # Define materials # #################### mat_1 = openmc.Material() mat_1.add_element('U', 1.0) mat_1.set_density('g/cm3', 10) mat_1.temperature = 293.15 mat_2 = openmc.Material() mat_2.add_element('Zr', 1.0) mat_2.set_density('g/cm3', 6) mat_2.temperature = 293.15 mat_3 = openmc.Material() mat_3.add_element('H', 1.0) mat_3.set_density('g/cm3', 1.0) mat_3.temperature = 293.15 mats = openmc.Materials([mat_1, mat_2, mat_3]) mats.export_to_xml() ################### # Define geometry # ################### case_A_cyl_in = openmc.ZCylinder(r=7.5050) case_A_cyl_out = openmc.ZCylinder(r=7.5438) case_A_z_in_top = openmc.ZPlane(1.00455) case_A_z_in_bottom = openmc.ZPlane(-1.00455) case_A_z_out_top = openmc.ZPlane(1.02555) case_A_z_out_bottom = openmc.ZPlane(-1.02555) case_A_in = openmc.Cell() case_A_in.fill = mat_1 case_A_in.region = -case_A_cyl_in & -case_A_z_in_top & +case_A_z_in_bottom case_A_out = openmc.Cell() case_A_out.fill = mat_2 case_A_out.region = -case_A_cyl_out & +case_A_z_out_bottom & -case_A_z_out_top & ~ case_A_in.region case_A_u = openmc.Universe(cells=[case_A_in, case_A_out]) ################################################################## # Place case_A_u in my problem universe at different coordinates # ################################################################## #My problem consists of a taller cylinder with the same diameter #and the previous case_A_u is placed there at some axial position problem_cyl_out = openmc.ZCylinder(r=7.5438, boundary_type = 'reflective') height_case_A = 1.02555 * 2 problem_z_case_A_top = openmc.ZPlane(height_case_A) problem_z_case_A_bottom = openmc.ZPlane(0) problem_z_top = openmc.ZPlane(10, boundary_type = 'reflective') problem_z_bottom = openmc.ZPlane(-10, boundary_type = 'reflective') problem_cell_1 = openmc.Cell() problem_cell_1.fill = case_A_u problem_cell_1.region = -problem_cyl_out & +problem_z_case_A_bottom & -problem_z_case_A_top problem_cell_2 = openmc.Cell() problem_cell_2.fill = mat_3 problem_cell_2.region = -problem_cyl_out & +problem_z_bottom & -problem_z_top & ~ problem_cell_1.region problem_universe = openmc.Universe(cells=[problem_cell_1, problem_cell_2]) geom = openmc.Geometry(root = problem_universe, surface_precision = 6) geom.export_to_xml() ##################### # Source Definition # ##################### point = openmc.stats.Point((0, 0, 1)) src = openmc.Source(space=point) ############ # Settings # ############ settings = openmc.Settings() settings.source = src settings.batches = 100 settings.inactive = 20 settings.particles = 1000 settings.export_to_xml() ####### # Run # ####### openmc.run() ``` So the first cylinder is defined between z = -1.02555 and 1.02555. The cell in the second universe to include this first cylinder is between 0 and 2\*1.02555. If I run openmc I get this ```plaintext WARNING: After particle 13 crossed surface 5 it could not be located in any cell and it did not leak. ERROR: Maximum number of lost particles has been reached. ``` When running openmc-plotter I get this, where the black color means that it has not been filled with any material. This makes me think that all universes must share the same origin. ![imagen](https://global.discourse-cdn.com/free1/uploads/openmc/original/2X/7/7c0bd87b8883da11662d458c78c9703a23e25c79.png) Is it required that the origin of coordinates is the same for both universes? That is, do I need to define the first cylinder Z planes related to the main problem universe? If this is the problem, is there a way to define the pieces of the problem individually, to add them later to the problem universe for a case like this where I will change the axial location several times? I can do it by calculating the values using Python, but I wonder if there is an OpenMC way.