I got the changed materials in different burnup steps using export_to_materials(). In theory, the mox fuel will change density during the burnup (fision…) in practical. However, the result was:
Initial mox materials:
burnup step1 & step16:
As shown in Figures, the “sum density” did not change, always 0.06987922652
Did I set something wrong? I’d really appreciate any suggestions.
Hi Jiang,
I had a similar experience with mine, I think its just the way OpenMC does depletion, although someone with more knowledge can correct me if I’m wrong. OpenMC assumes mass density is constant over depletion, therefore assuming that irradiation induced densification doesn’t occur. The sum is it re-normalizing the nuclide density with respect to your previously stated mass density, so even if you had the expected increase in density early in the burnup, this is normalized out of the results. I’m a little surprised it does it this way as I would expect it to cause errors in the depletion results, but maybe they are insignificant for the purposes of nuclear simulations. If you wanted to fix this I imagine you would need another script coupled with OpenMC to recalculate your mass density between each depletion stage, re-normalize the nuclide densities with respect to your new mass density, and input these nuclides and your new mass density into the next depletion step.
Have you checked the mass density on each burnup step? As you already know, the same atomic density, such as 0.06987922652 #/barn-cm, could end up with a different mass density caused by the change in fuel composition. Also, from my calculation, the atomic density was increased, which I think came from the fission product, but in my calculation, I burned it for more than 50% burned U235.
You might also notice the reduction of fissile mass during burnup calculation, especially in a non-breeder reactor. In general, the total heavy metal density was reduced during the burnup calculations.
