Could not find the cell containing particle 2

[paulromano]

@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

@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 for further information.