Monte Carlo codes are the most accurate way to solve the neutronics in a reactor core but can be computationally expensive, especially for when feedback effects are considered or for transient calculations. In this paper, we use the fission matrix (FM) method to perform static and transient calculations with point-kinetics equations and a quasi-static model for an adiabatic transient with feedback. This was applied to a three-dimensional model of the Transient Reactor Test Facility (TREAT) experimental reactor using the Monte Carlo code Serpent for reference calculations and to generate fission matrix databases (FMDBs). In previous works, FMDBs were generated with uniform fuel temperature profiles. Here, we analyze the use of FMDBs with nonuniform temperature profiles, for static and transient calculations. For static calculations, comparison between Serpent and the FM method using nonuniform and uniform FMDBs showed maximum differences in multiplication factors of 57.0 and 77.9 pcm, respectively. For the fission source distribution, comparisons showed maximum root-mean-square differences of 1.10% and 4.89% for nonuniform and uniform FMDBs, respectively. Similar results were obtained when using homogenized databases. Therefore, using nonuniform FMDBs produces a better approximation than uniform databases. For transient calculations, comparisons between both database sets showed differences of 0.9% and −1.8% for the peak and final total power.