Continuous-energy Monte Carlo is a high-fidelity tool for reactor-physics analysis, but its cost limits its use in workflows requiring repeated transport-depletion calculations. In this context, lumped fission products provide a model-reduction strategy by representing the collective neutronic effect of the fission product inventory through a reduced set of pseudo-nuclides. This thesis develops a methodology for generating and applying burnup-dependent multigroup lumped fission product cross sections in fast-spectrum Monte Carlo depletion. The method is implemented in OpenMC and is based on parent-specific LFPs. Effec tive fission yields are obtained through auxiliary depletion calculations in which heavy metal transmutation and decay routes are suppressed, leaving fission as the only outgoing reaction channel of the selected parent.