how to calculate neutrons per fission in MOX or impure fuel

[connor]

Hi, sorry if I posted this in the wrong place, I'm new to this site.
I was wondering if there is a clearly defined formula for predicting neutron release per fission in fuel that has multiple isotopes. For example if you have X% U235 and X% U238 and you know you have a certain mass, or X%U235 X%U238 X%P239 ect, can you calculate the neutron release with one formula. I know there are dimensions for pure critical masses without a moderator and that you can predict the effect a moderator will have on the criticality with the diffusion equations and the neutron transport equation (roughly speaking, please correct me if I cited the wrong equations I'm not being very specific) but you would never have any pure critical mass in a reactor due to the danger and cost, but mostly danger, so I wondered when designing a reactor if there is one equation for neutron production in fuel with lower percentages of U235 and fuel with both U235 and U238 together because if you knew the neutron production per mass at variable percentages, you could predict criticality safely in a moderator instead of a critical mass of a pure element where there is no moderator. Sorry if I was unclear in my phrasing of my question, i typed this very fast, please reply for more information if you do not understand what I'm asking but think you could help. Any information helps, again sorry if I posted this in the wrong place.