Hello all
i have attached the geometry.xml file along with tallies.xml and tallies.out file with this message. can anyone help in deciphering the out of distribcell i.e. how to find the flux value for individual pin when one single universe for fuel pin is being used in multiple assemblies?
thanks for your help in advance
tallies.out (1.02 MB)
tallies.xml (314 Bytes)
geometry.xml (11.3 KB)
Let’s take the first label as an example:
0->1000->1000(0,-4,1)->9->2->8->1->10(0,-6,1)->1->11
This means universe 0 (base universe) --> cell 1000 --> lattice 1000 position (x=0,alpha=-4,z=1) --> universe 9 --> cell 2 --> universe 8 --> cell 1 --> lattice 10 position (x=0,alpha=-6,z=1) --> universe 1 --> cell 11
I’ve attached a PDF showing how the hexagonal lattice indexing works.
Best,
Paul
hex_lat.pdf (26.3 KB)
Thanks for your reply. The problem which i am facing is that the cell over which we are tallying i.e. cell 11 which makes universe 1 is in being used in hex_lattice 10 and 11 both. But using the guide you provided i can’t see any entries for lattice 11. also what about the entries at the last i.e.
Distributed Cell 0->1000->1000(-1,4,1)->9->2->
8->1->10
Total Material
Flux 4.26368E-04 +/- 8.06462E-05
Distributed Cell 0->1000->1000
Total Material
Flux 6.94903E-04 +/- 9.50518E-05
thank you very much for your help 
Hello King,
I think I can answer the question why you don’t see any entries for hex_lattice 11. Lattice 11 is contained in universe 6, but I don’t see universe 6 anywhere else in the geometry. OpenMC will only output information for cells that neutrons make it into. Since universe 6 isn’t contained in the geometry anywhere, no neutrons enter it, and it (along with lattice 11) will not appear in the tallies.out.
I don’t know why cell 11 isn’t present in the last entries of that output. It might be a bug with OpenMC. I’m investigating it now, and I’ll let you know when I figure it out.
I think we’ve squashed the bug. If you checkout the latest “develop” branch you should see proper results.
Thank you