Burn-up calculation different materials

Hi all,

I am running a depletion calculation with lots of different materials and need to determine the burn-up of each material at the end of irradiation and would like to confirm my method of working.

I run the calculation using normalization_mode=‘fission-q’ in the CoupledOperator (Not sure if I should use “energy-deposition” mode?). Then I tally the heating scores of each depletable material H and convert these to J/source: H'=H\cdot 1.602\times 10^{-19}.
After which I determine the normalization factor f by dividing the total power given to the integrator P by the sum of all heating tallies \sum_{mat-depl} H' from ‘tallies.out’.
Then I can calculate the burn-up using: BU [MWd/tHMi]= \frac{H'\cdot f \cdot t_{irr}}{m_{HMi}}

Is this correct? I know there will be discrepancies due to not taking photon transport into account, but I am not yet interested in this. Thanks for helping me out.

It is clear to me now. I have to use normalization_mode=‘energy-deposition’ and because it is a neutron only calculation, the ‘heating-local’ tally should be used. Excuse me for the confusion as it is clearly explained here: Energy deposition