(probably stupid) question about current tally backscattering

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1

Hello all,

I am running a tritium breeding simulation on a blanket geometry and would like to calculate the TBR but for this I need to know the number of neutrons incident into the blanket. I am planning to use the current score on it’s source-facing surfaces in order to achieve this.

Let’s assume for the time being my geometry looks something like the image below; a point source, and a current tally on a surface of a cell. I’d like to know the number of neutrons incident on said surface (in this case, neutrons 1, 2, & 3). However, my question is: how does the current score treat backscattering? If neutron 1 is backscattered out of the cell, would this effectively ‘cancel’ it from the tally? Or is it a case that once it crosses the surface, it’s counted.

Thanks!

IMG_2334
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2

I’ve also just realised that OpenMC surfaces are infinite planes/cylinders so tallying the current across one would not tell me the fraction of neutrons entering into the cell itself…. anyone have any ideas how I could get this?

3

I don’t have much experience with using them, but I believe you can get around the issue of backscatter by using a openmc.CellFromFilter - you can specify the cell that the particles should be travelling from, and thus only particles travelling from that direction will be tallied.

4

I believe the surface only exists to the extent it is part of the defined geometry. That’s how it works in certain other codes. Once a particle enters another cell, it only sees the surfaces defining that cell.