Primarly neutrons distribution throughout fuel rods of a hexagonal lattice

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1

Dear OpenMC users and experts,

I’m wondered whether it’s possible to distribute primarly neutrons only across fuel rods of a hexagonal fuel assembly. I’m working on a WWER-440 fuel assembly modeling and would like to use a source distribution similar to KENO-VI start data type 0 (NST=0): primaries are uniformly distributed throughout fissile rods (4 types of “red” circles on the geometry image below).

What kind of external source would you suggest to use there? Does it require using of a compiled source? Probably, there is a recommended approach for setting the source for hexagonal assemblies?

Best regards,

Ihor

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The best way to get a uniform distribution through fuel pins is to use a source constraint. Something like the following should work:

min_extents = (-1., -1., -1.)
max_extents = (1., 1., 1.)
spatial = openmc.stats.Box(lower_left=min_extents, upper_right=max_extents)
source = openmc.IndependentSource(space=spatial, constraints={‘domains‘ : [FUEL_MATERIAL], ‘fissionable‘ : True})

Replace min_extents and max_extents with the bounding box for your geometry, and FUEL_MATERIAL with the openmc.Material representing the fuel. You can also use an openmc.Cell or openmc.Universe in FUEL_MATERIAL.

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