hexagonal lattice

[Ahmed K. Madani]

I'm trying to simulate  VVER reactor assembly but i couldn't understand how to set hexagonal lattice in openmc the assembly VVER contain 312 fuel pins  and 19 guide tubes  , so can you share with me any way to understand how to set it up because i cannot understand the concept of the hexagonal lattice in openmc   

Thanks 

[Paul Romano]

Ahmed,

If you are writing XML files directly, you should take a look at the documentation for the <hex_lattice> element. If you are using the Python API, there is an example of creating a hexagonal lattice here and here. Let us know if you are still confused after looking over the documentation and the examples.

Best regards,

Paul

On Tue, Mar 14, 2017 at 4:06 PM, Ahmed K. Madani <ahmedk...@gmail.com> wrote:

I'm trying to simulate  VVER reactor assembly but i couldn't understand how to set hexagonal lattice in openmc the assembly VVER contain 312 fuel pins  and 19 guide tubes  , so can you share with me any way to understand how to set it up because i cannot understand the concept of the hexagonal lattice in openmc   

Thanks 

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[Marcin Rowinski]

Hi!

Have you solved you problem? I am currently do very similar thing, an assembly of VVER reactor. I have managed to create the geometry, picture made in openmc attached. But cannot make it work.

I believe that the problem is either in some surface, region or settings definition, however, I am running out of ideas how to fix it.  My base for geometry was this workshop suggested by Paul. The calculations work fine when I use cylinder or rectangular region. But when I use openmc.get_hexagonal_prism(edge_length, orientation='y', boundary_type='reflective') it does not. I keep receiving following error:

WARNING: After particle 3 crossed surface 10019 it could not be located in any
          cell and it did not leak.
 WARNING: After particle 3 crossed surface 10019 it could not be located in any
          cell and it did not leak.
 WARNING: After particle 4 crossed surface 10018 it could not be located in any
          cell and it did not leak.
 WARNING: After particle 6 crossed surface 10019 it could not be located in any
          cell and it did not leak.
 WARNING: After particle 7 crossed surface 10019 it could not be located in any
          cell and it did not leak.
 WARNING: After particle 8 crossed surface 10018 it could not be located in any
          cell and it did not leak.
 WARNING: After particle 9 crossed surface 10019 it could not be located in any
          cell and it did not leak.
 WARNING: After particle 11 crossed surface 10019 it could not be located in any
          cell and it did not leak.
 WARNING: After particle 12 crossed surface 10018 it could not be located in any
          cell and it did not leak.
 WARNING: After particle 13 crossed surface 10019 it could not be located in any
          cell and it did not leak.
 ERROR: Maximum number of lost particles has been reached.
ERROR STOP 

I would appreciate any help in this matter. Is there any example of similar geometry? Please let me know if you have any questions?

Best,
Marcin

[Paul Romano]

Hi Marcin,

Without having a look at your inputs, it's hard to say where the problem lies. Looking at the attached picture, I am a little bit suspicious of the fuel pins. The regions in the fuel pin that are white appear to be voids -- have you actually assigned cells there? From the Python API, if you want to create a cell with no material in it, simply do not assign anything to the fill, e.g.

void_cell = openmc.Cell(region=...)

or if you were writing XML directly, you would need to put material="void", e.g.

<cell id="1" material="void" region="..." />

If you're comfortable sharing your inputs, I can try to help diagnose further.

Best,

Paul

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