Difficulties obtaining a flux or current tally score

User Support
1

Hello

I’m running OpenMC on a human phantom geometry that consists of an outer volume and multiple inner volumes. I will add a picture for reference.

output
output221×736 29.6 KB

I’m trying to set the model up so that I can place the source in any inner volume (for example the brain) and have the model calculate the flux/current for any other chosen volume (for example the heart). But I’m running into a problem where I only get a numbered tally result for the most outer volume (skin) and any other volume inside gives me the tally result of 0. I have tried both with SurfaceFilter and MaterialFilter but the result is the same. I also tried adding SurfaceFromFilter to counter this result but with that added even the most outer volume gives me the 0 result.

At this point I can’t seem to figure out where this error could be resulting from and how to move on from this. Any help and insight would be greatly appreciated.

2

Hi Krislin,

You could overlay a mesh tally and confirm that particles are transporting within the voxel geometry.

I suspect there’s an issue with how the tallies are defined, perhaps too many filters.
When I tally voxels I typically use a single particle filter, and a single distribcell filter.
Can you provide your XMLs?

Perry

3

Hi Perry,

thank you for the response.
So far I have only been using the surface filter, photon particle filter and energy function filter in my tallies. Occasionally adding cell from filter as well for testing purposes.

Here are my XML files:
geometry.xml (719 Bytes)
materials.xml (69.3 KB)
settings.xml (366 Bytes)
tallies.xml (1.3 KB)

Krislin

4

Hmm, I’m not familiar with CAD/mesh geometry.

I would try:

  • score flux
  • photon particle filter
  • material filter, but only one material filter specifying one material, per tally
    (have individual tallies with individual filters for each material)

If this works, then add the energy-function filter.

5

Thank you for the suggestion but using material filter and removing energy function filter still gives the same results- 0 flux tally for inner materials and numbered result for the outer material.

6

Make sure it isn’t a tally sensitiity issue and make your source really hot to see if it returns anything. I’ve had that issue in the past.

7

Increasing the source energy doesn’t bring any changes. But how would I go about checking the flux tally sensitivity?

8

Check openmc.trigger. There are ways to adjust the tally threshhold (if that is the problem).

9

It seems like tally sensitivity isn’t the problem here. I have been playing around with different energy levels and triggers but the results remain the same. Is there any way to conduct some sort of ‘diagnostics’? Without any error messages I feel like I’m just stumbling around in the dark.

10

I have had a similar problem in the past twice with reactor simulations, but both times I was using local geometry, not a dagmc import. The first time I had my regions defined wrong and it ended up overriding all the cells into water (even though the voxel slice showed the correct geometry), which sounds similar to your issue where you are only getting skin registrations. If there is a way to check whether all of the source particles are interacting with the skin when that shouldn’t be the case, that might be something worth checking. The second time I had the issue I was generating multi-group cross sections for a full reactor simulation, but I wasn’t running enough particles (not particle energy, particle count) for certain pieces of aluminum to register any values. These are the only possible causes that come to mind, as I have no experience with imported geometries. Hope this helps.