Single Energy Photon Source

Good day,

I am currently using OpenMC to do dose calculations and comparing it to data produced by a colleague. I have performed some initial calculations and had a large percentage difference when it came to our photon source calculations. This was due to the lack of energy on my side of the code. However, when I included the energy that my supervisor had input on her file, which was 4.4 MeV only, my simulation only ran on one batch for at least 2 hrs and did not proceed to a second batch. I used openmc.stats.Tabular for this energy for this run. Then recently I attempted to run with the same energy but then using openmc.stats.Discrete and had a result but had a larger percentage difference. What would be the best way for me to set the photon energy for my source?


Update: I just read the discussion the execution settings and it seems that using openmc.stats.Discrete is the way to do the energy. I am now trying to check on my data again and see if there are errors on my code the might have produced the large percentage difference. Maybe someone knows what could be the source of it? Thanks again.

@aeandallo Yes, the correct way of specifying a monoenergetic source would be to use the openmc.stats.Discrete class as follows:

energy_dist = openmc.stats.Discrete([4.4e6], [1.0])  # energies, probabilities
source = openmc.Source(energy=energy_dist)

Without any further information about your model or your colleagues, it’s not possible to speculate on what could be causing differences in results.

Hello, I can’t really share the model but we both had the same input for strength, energy distribution, space distribution, and dose coefficients. If it can possibly help, I am more confident with my data regarding the neutron dose calculation (same strength, energy distribution, space distribution, and dose coefficients too) and I ended up with a percentage difference of around 10%

Hello again,

I did some updated calculations to the same model. The model is a cylindrical AmBe source within a capsule and is surrounded by air (within a 3m x 3m x 3m box). The doses were measured in phantom (air) cells that are 30cm x 10cm x 50cm boxes situated on 50cm and 100 cm away from the source. As mentioned earlier, the source is a monoenergetic photon source with an energy of 4.4Mev and strength of 5.96e6. To calculate for the dose, I have used ICRP-74 and was a fixed source calculation for a photon source only. These details came from the MCNP data that my supervisor and I tried to recreate it using OpenMC.

I calculated for the percentage difference from both calculations from both simulations and at the difference phantom cells and got around 51% percentage difference. I am not sure at this point on what could have caused this difference.

This was not our first time comparing dose from a monoenergetic photon source and we had a relatively lower percentage difference compared to this one.

@aeandallo Again, without much information on your models it’s hard to speculate what might be the source of the differences. First, you should ask yourself:

  • Is the geometry/materials definition exactly the same between the two codes?
  • Are the codes using the same cross section data?
  • Are the physics treatments/options/cutoffs the same in both codes?
  • Are the units of your results consistent?
  • Have you accounted for uncertainties in the results for both codes?

If you’ve gone through all of these questions and believe everything is consistent, what I would suggest doing is systematically simplifying the models through a series of small changes (e.g., removing materials, removing cells) until you find a configuration where the results match.

I just had some discussions with my supervisor and we compared our codes. I already modified my code to follow my supervisors MCNP code, which were minor changes in material settings and total number of particles for the simulation. I am including my neutron source file with photon transport since we also wanted to investigate the differences in the photon values. Here is the summary of our percentage difference

Neutron Dose: ~3.8%
Primary Photon: ~50.8%
Secondary Photon: ~38.5%

I am attaching both OpenMC input files that I have used. For the post-processing of data. I just extract the values from the tallies and divide by the volume of the cell. (The conversions that I have used came from my supervisor.)

Thanks again. (7.7 KB) (11.8 KB)

*We are just curious on to why my results are somewhat off from her values. I am the first and the only one using OpenMC in my group so I could possibly have some part of my code wrong since I have only been using it for the past months.