@srichr221 I’ve been doing OpenMC vs FISPACT comparisons for a paper that is going to be submitted shortly along with @eepeterson. In particular, we’ve looked at predicted material compositions after irradiation/decay as well as the resulting gamma spectra. The short answer is that given equivalent nuclear data, the two codes will produce predictions that are very close to one another. There are a few subtle differences — for example, FISPACT uses an approximate gamma spectrum when no discrete lines are available in the underlying ENDF evaluation, whereas OpenMC will try to use a continuous spectrum from the ENDF evaluation if available.
Another subtle difference is that generally you provide FISPACT with multigroup fluxes, and it has to make some assumption about the within-group flux in order to calculation reaction rates. With OpenMC, it’s easy to generate the exact microscopic cross sections that are used to calculate reaction rates. @eepeterson may have more to say on the impact of that.
Keep in mind that FISPACT has quite a few features that are not yet replicated in OpenMC (e.g., pathway and uncertainty analysis), so depending on your needs OpenMC may or may not be a suitable alternative. If there is a feature you really need that is not present in OpenMC though, do let us know!