Hi Criss,
I think you need to declare the probability in the unit per eV for each neutron energy bin.
I am making a small notebook to show you the input and how it works in a cell filled with voids.
You can check if the neutron energy distribution matched between one you specified in the input and calculated by openmc if we use a voided cell.
tallyingE.ipynb (161.9 KB)
I am also adding some material if you want to fill the cells with other materials and see how the neutron spectrum changes/shifts.
I also added some lines if you want to use discrete neutron energy to check if the neutron energy calculated by openmc matches your discrete energy specified in input.
This discussion helped me on specifying the energy probability table.