# Could not find the cell containing particle 2 **URL:** \ **Category:** User Support\ **Created:** [June 2, 2021, 1:27am UTC](https://openmc.discourse.group/t/could-not-find-the-cell-containing-particle-2/1230 "2021-06-02T01:27:31Z")\ **Posts on this page:** 4\ **Page:** 1
**Author:** ![HaoPF](https://avatars.discourse-cdn.com/v4/letter/h/ba9def/32.png) [@HaoPF](https://openmc.discourse.group/u/HaoPF)\ **Post date:** [June 2, 2021, 1:27am UTC](https://openmc.discourse.group/t/could-not-find-the-cell-containing-particle-2/1230/1 "2021-06-02T01:27:31Z")
Hello, everyone, I’m a beginner to OpenMC, and now I’ve built a model of my own.But an error occurred while I was running it. I found that OpenMC works well when my model is infinite in volume (with no upper and lower boundaries).But when I add the upper and lower bounds, OpenMC gets an error.The error reporting page is as follows: ![1.png](https://global.discourse-cdn.com/free1/uploads/openmc/original/2X/7/75ceb09b95acc6bb3a0d9b2bb2fdf0d807b7fab7.jpeg) The interface that works well in infinite volume is as follows: ![2.png](https://global.discourse-cdn.com/free1/uploads/openmc/original/2X/e/eb1020fdff00df1b898606c46af02590b1513fa1.jpeg) Specifically, I changed the upper and lower boundaries in the geometry.This is as follows: ![3.png](https://global.discourse-cdn.com/free1/uploads/openmc/original/2X/0/09ad1ef9d4a45c102ab416c266ec708a5ccaa0f5.jpeg) The resulting image looks pretty good.I’m sorry I’m so stupid I don’t know where the mistake is. Sincerely hope to get your help. thanks HaoPF ---
**Author:** ![HaoPF](https://avatars.discourse-cdn.com/v4/letter/h/ba9def/32.png) [@HaoPF](https://openmc.discourse.group/u/HaoPF)\ **Post date:** [June 2, 2021, 1:54am UTC](https://openmc.discourse.group/t/could-not-find-the-cell-containing-particle-2/1230/2 "2021-06-02T01:54:17Z")
Here is the code for my model: import openmc material\_Na = openmc.Material(name=‘Na’) material\_Na.temperature = 300 material\_Na.set\_density(‘g/cm3’, 0.968) material\_Na.add\_element(‘Na’, 1.0) material\_K = openmc.Material(name=‘K’) material\_K.temperature = 300 material\_K.set\_density(‘g/cm3’, 0.862) material\_K.add\_element(‘K’, 1.0) coolant = openmc.Material.mix\_materials([material\_Na, material\_K], [0.22, 0.78], ‘wo’) coolant\_pipe = openmc.Material(name=‘Re’) coolant\_pipe.temperature = 300 coolant\_pipe.set\_density(‘g/cm3’, 0.732) coolant\_pipe.add\_element(‘Re’, 1.0) fuel = openmc.Material(name=‘UN’) fuel.temperature = 300 fuel.set\_density(‘g/cm3’, 14.264) fuel.add\_element(‘U’, 1.0, enrichment=55.0) fuel.add\_element(‘N’, 1.0) material\_Mo = openmc.Material(name=‘Mo’) material\_Mo.temperature = 300 material\_Mo.set\_density(‘g/cm3’, 10.200) material\_Mo.add\_element(‘Mo’, 1.0) material\_Re = openmc.Material(name=‘Re’) material\_Re.temperature = 300 material\_Re.set\_density(‘g/cm3’, 21.020) material\_Re.add\_element(‘K’, 1.0) cladding = openmc.Material.mix\_materials([material\_Mo, material\_Re], [0.70, 0.30], ‘wo’) materials = openmc.Materials([coolant, coolant\_pipe, fuel, cladding]) materials.export\_to\_xml() # Define problem geometry s01 = openmc.ZCylinder(r=0.3000) s02 = openmc.ZCylinder(r=0.3800) s03 = openmc.ZCylinder(r=1.6325) cell01 = openmc.Cell(name=‘Cell 01’) cell02 = openmc.Cell(name=‘Cell 02’) cell03 = openmc.Cell(name=‘Cell 03’) cell04 = openmc.Cell(name=‘Cell 04’) cell11 = openmc.Cell(name=‘Cell 11’) cell01.region = -s01 cell02.region = +s01 & -s02 cell03.region = +s02 & -s03 cell04.region = +s03 cell01.fill = coolant cell02.fill = coolant\_pipe cell03.fill = fuel cell04.fill = cladding cell11.fill = cladding universe01 = openmc.Universe() universe02 = openmc.Universe() universe01.add\_cells([cell01, cell02, cell03, cell04]) universe02.add\_cells([cell11]) lattice1 = openmc.HexLattice() lattice1.center = (0., 0.) lattice1.pitch = (3.4,) lattice1.outer = universe02 ring\_1 = [universe01]\*24 ring\_2 = [universe01]\*18 ring\_3 = [universe01]\*12 ring\_4 = [universe01]\*6 ring\_5 = [universe01]\*1 lattice1.universes = [ring\_1, ring\_2, ring\_3, ring\_4, ring\_5] outer\_surface1 = openmc.ZCylinder(r=16) outer\_surface2 = openmc.ZPlane(z0=0.0) outer\_surface3 = openmc.ZPlane(z0=43.2) outer\_surface1.boundary\_type = ‘vacuum’ outer\_surface2.boundary\_type = ‘vacuum’ outer\_surface3.boundary\_type = ‘vacuum’ main\_cell = openmc.Cell(fill=lattice1, region=-outer\_surface1 & +outer\_surface2 & -outer\_surface3) geometry = openmc.Geometry([main\_cell]) geometry.export\_to\_xml() # OpenMC simulation parameters batches = 100 inactive = 30 particles = 10000 settings\_file = openmc.Settings() settings\_file.batches = batches settings\_file.inactive = inactive settings\_file.particles = particles settings\_file.export\_to\_xml() ---
**Author:** ![paulromano](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/paulromano/32/486_2.png) [@paulromano](https://openmc.discourse.group/u/paulromano)\ **Post date:** [June 2, 2021, 12:14pm UTC](https://openmc.discourse.group/t/could-not-find-the-cell-containing-particle-2/1230/3 "2021-06-02T12:14:13Z")
@HaoPF You haven’t specified a source distribution at all. Please see my response to another recent post with the same issue: > [@Critical Boron Concentration in Burnup Calculation](https://openmc.discourse.group/t/critical-boron-concentration-in-burnup-calculation/1173/6): > > @ayaz It looks like you haven’t specified a source distribution at all. By default, OpenMC will source particles at the origin, (0, 0, 0). The z-dimension in your geometry starts at z=0, so particles would be sourced directly on the boundary, which can cause issues like this. You should define a starting source that covers the entire length of your fuel bundle. See the user’s guide section on [external source distributions](https://docs.openmc.org/en/latest/usersguide/settings.html#external-source-distributions) for further information. ---
**Author:** ![HaoPF](https://avatars.discourse-cdn.com/v4/letter/h/ba9def/32.png) [@HaoPF](https://openmc.discourse.group/u/HaoPF)\ **Post date:** [June 3, 2021, 1:22am UTC](https://openmc.discourse.group/t/could-not-find-the-cell-containing-particle-2/1230/4 "2021-06-03T01:22:09Z")
Thanks，@paulromano As you said, I define a starting source that covers the entire length of your fuel bundle and it works well. I really appreciate your help. HaoPF