How to get local power?

User Support
1

Hi,

I am completely new here, and a bit unfamiliar with openmc units.
I’m modeling a nuclear core reactor, and trying to predict the axial and radial power profiles knowing its geometry and the equilibrium concentrations of the nuclides. To do this, I am trying to get the local power in each bin of a mesh I created (I just want something in Watt for each bin):

1/ what tallies should I use? I am currently hesitating between ‘fission’ (to do something like : Power = ‘fission_rate’ * Q[200MeV/fission] * theorical_number_of_fissions), or ‘heating’ (to do Power = ‘heating’ / duration_of_simulation), but I’m not sure.

2/ in most of discussions I found (e.g. here or here), we assume that we know the total power of the reactor, and we use it to calibrate our local power value. Can I do my work without using this technique? I think we should have enough information to do without it, and it should be a way to confirm our results (by checking that the core power is indeed equal to the sum of the local powers)

Thanks a lot,
Jack

2

  1. The first is standard.

  2. No, if you need local power in Watts, you will need to impose a total power in Watts. Your OpenMC solution is valid at any flux level. To put an absolute number on it requires an assumption or measurement.

1 Like
3

Thanks @tjlaboss. I’m still wondering:

  1. I don’t know how to get theorical_number_of_fissions. I guess it’s a bit tedious to go through the fission probability of each nuclide that makes up our materials;
  2. do you think I can try this: Power(voxel_i) = Power_tot * e(i)f(i)/[sum_i e(i)f(i)], where e(i) = nb_neutrons_entering_voxel_i = flux(voxel_i), and f(i) = pbty_of_fission_into_voxel_i = fission(i)? I am not sure about the second equlities in each case.

Thanks a lot.

4

It is rarely necessary to evaluate the contribution to fission power from each nuclide. You can do it if you like, and it’s a good exercise. Assuming 200 MeV/fission for uranium is pretty common.