Depletion Isotope XS Update throughout Calculation

User Support
1

Hi everyone!
I am new to the OpenMC and planning to conduct depletion calculations but I have a few questions that scratch my mind. Namely,
First of all, the user guide states that “The depletion interface relies on OpenMC to perform the transport simulation and obtain reaction rates and other important information.”
My first question is why transport simulation is used instead of Monte Carlo simulation.

Secondly,
As described in the Theory and methodology of OpenMC, The depletion interface relies on OpenMC to perform the transport simulation and obtain reaction rates and other important information.
Does that mean at the end of each time step OpenMC generates isotopes by governing the transmutation and decay equation and for the following time step OpenMC uses new cross-sections corresponding to isotopes generated in the previous time step? If it is, not necessary to update materials OpenMC already does it. Am I correct?

I am looking forward to hearing from you.
Best regards.

2

Monte Carlo simulation is the method used to solve the transport equation, so sometimes it’s referred to as the “transport simulation”

Yes, OpenMC will solve the transmutation (Bateman) equations and update the material compositions before the next step, so you don’t need to make any manual changes.

3

I meant by the transport equation like Sn Pn or MOC but as I see, you already got it and gave well-explained answers, thanks for helping really appreciated it. have a nice day.
Best regards.