We have added functionality for running depletion simulations independently of neutron transport in OpenMC, an open source Monte Carlo particle transport code with an internal depletion module. Transport-independent depletion uses pre-computed static multigroup cross sections and fluxes to calculate reaction rates for OpenMC's depletion matrix solver. This accelerates the depletion calculation, but removes the spatial coupling between depletion and neutron transport. We used a simple PWR pincell to validate the method against the existing transport-coupled depletion method. Nuclide concentration errors roughly scale with depletion time step size and are inversely proportional to the amount of the nuclide present in a depletable material. The magnitude of concentration error depends on the nuclide of interest. Concentration errors for low-abundance nuclides at longer (30-day) time steps exhibit large negative initial concentration the becomes more positive with time due to overestimation of nuclide production stemming from the lack of spatial coupling to neutron transport. For ten 3-day time steps, fission product concentration errors are all under 3%. Actinide concentration errors range from 10-15% for Am and Cm, 5-7% for Pu and Np, and 2% and less for U. Surprisingly, the numbers are similar for 30-day time steps. These results demonstrate the potential of this new method with moderate accuracy and extraordinary time savings for low and medium fidelity simulations. Concentration error characterization on larger models remains an open area of interest.