Photon flux with a layer of water and air

[wahidluthfi]

Hi Hedi,
I am not into photon transport since I am usually doing neutron transport for core modeling, so for my case (core modeling) the tally normalization factor can be found from the core thermal power.
But since photon transport tally outputs are dependent on source strength declared in simulation settings, then I think you could find other sources in discussion for photon tally normalization such as this

Photon Normalisation - Photon count/dose seems far too low

Hello! I am using photon and neutron tallies, loosely following the code given in the example: https://github.com/fusion-energy/neutronics-workshop When I output the dose tallies, after normalising by (power/heating [ev/particle]), as suggested for a non fixed-source calc, I get the gamma dose coming out to be much too small (10x smaller than neutron dose with no gamma shielding). I’m wondering now, if there is a different method for normalising the photon tallies, do you need to do a heating…

and some examples mentioned in this discussion might help you later

How to simulate reactor shielding in OpenMC?

Hi Kaspar Welcome to the community. I have a few examples for question 1 (How do I define an absorbed dose tally?) here

neutronics-workshop/tasks at main · fusion-energy/neutronics-workshop

regarding this statement “Let’s take the simple case where the box is filled with vacuum, in my code I’m simulating 10,000 particles so at the end I have to find my 10,000 particles”
I think, since you talk about vacuum and monodirectional photons, then your photon source will not interact with any material, and it will move through space ‘to infinity and beyond’. hehe, I recommend you use air instead of vacuum since it feels more practical for me.

it will be different when you use a material since the interaction between photons and material will have various products: photoelectric, Compton scattering, and pair production which will change the total number of photons in the system. correct me if I’m wrong.

also, regarding the tally dimension, when using score: ‘flux’ on the mesh tally, I think the dimension is #-cm/source since when using mesh tally, openmc counts the number of particles in a specific mesh volume. so to bring back the flux into the typical flux dimension (#/cm2-sec), you will need to multiply this tally with the source and divide it by mesh volume, so #-cm/source. #/sec. 1/cm3. But that is what I understand when doing core calculations, I don’t know if it is also applicable to photon transport. somehow when tallying flux or power distribution, I only focus on the calculated trends or ratio, not the absolute value of each output.
documentation on tally: 8. Specifying Tallies — OpenMC Documentation

finally, regarding the mesh tally dimension, that’s right, and I think I could show how it could be done to plot along the x-axis with this notebook. but you will need to find other sources in other discussions regarding tally normalization for photon transport since I didn’t have any experience with it. Sorry Hedi
layerwater.ipynb (217.8 KB)

Wahid