Flux and Fission spectrum not giving proper values

User Support
1

Hello, everyone.

I am working on the fission and flux distribution for a reactor core, my k-eff value seems to match with the benchmark file that I am following, however, the flux and fission distribution are way off from the reference value. Could anyone help me figure out the reason behind this?

2

Hi,

Are you normalising the tally results? As outlined in: 8. Specifying Tallies — OpenMC Documentation

Typically you need to add some form of heating tally which is used in this normalisation

3

Which tally should I use for estimating the heating? I can see quite a few suggestions like “heating”, “heating-local” and “Fission-q-recoverable”. It is getting pretty confusing.

4

Hi Pearl,

I tend to just use “heating-local” if just doing an eigenvalue calculation with no photons:

Total nuclear heating in units of eV per source particle assuming energy from secondary photons is deposited locally. Note that this score should only be used for incident neutrons. See Heating and Energy Deposition.

Then “heating” if using photon transport. I have not got any experience with “Fission-q-recoverable” if I am honest…

Here is an example of how to use it (I am only showing the code including the heating tally)…
ref: 8. Specifying Tallies — OpenMC Documentation

heating = openmc.Tally(name="heating")
heating.scores = ["heating-local"]

with openmc.StatePoint(statepoint_file) as sp:
    heating_tally = sp.get_tally(name="heating")
    H = heating_tally.mean.ravel()[0]  # Heating rate in [eV/source particle]

Then from here you can follow the normalisation as in the referenced link.

5

Thanks! Currently running the code for the heating tally, let’s hope the values match.

Thanks again for your help.

6

Fingers crossed! Just remember that fission rate and neutron flux are normalised slightly differently.

7

I think for neutron flux, I might need to divide the value by volume whereas for fission rate volume isn’t required.