Analysis of nuclear reactor fuel composition and irradiation characteristics plays a crucial role in reactor design, licensing, safety analysis, and decommissioning. Through in-depth analysis of fuel composition and irradiation processes, it is possible to assess reactor performance, predict fuel lifespan and behavior, and optimize fuel management strategies. To conduct accurate and comprehensive analyses, a robust and reliable database is essential. This database should provide a wide range of key parameters that are vital for fuel analysis, including decay constants, decay branching ratios, decay radiation spectra, decay energies, neutron reaction cross sections, and fission product yields. These parameters are necessary for modeling and simulating the behavior of nuclear materials under different irradiation conditions. In this study, the focus was on developing an advanced methodology for processing refined depletion databases. Several evaluated nuclear databases, including ENDF/B-Ⅷ.0, ENDF/B-Ⅶ.1, JEFF-3.3, and JENDL-5, were utilized as sources of nuclear data. The researchers developed a specialized database generation program called ALG (automatic library generator) to facilitate the creation of a high-quality depletion nuclear database. The resulting database was designed using the HDF5 format, which offers efficient storage and retrieval of large amounts of data. Comparative analyses were conducted to assess the differences in key parameters of critical nuclides among the different evaluated nuclear databases. Three validation cases were conducted, including the pure decay of ²³⁷Np, the neutron activation of UO₂, and the pulse fission case of ²³⁵U. In the first case, which involved the decay of ²³⁷Np, excellent results were achieved, with only slight differences observed for short-lived nuclides when compared to the ORIGEN databases. Significant differences were observed in the UO₂ activation cases, primarily due to variations in neutron cross sections resulting from different neutron weight spectra. These discrepancies highlight the importance of considering neutron-induced reactions when analyzing activation processes, as the neutron spectrum has a significant impact on the results. For the calculation of the decay γ spectrum in the ²³⁵U pulse fission case, three newly processed depletion databases demonstrated excellent agreement with experimental measurements. This alignment between the calculated and experimental data validates the accuracy and reliability of the new processed databases. This evaluation helped identify variations in data, allowing for a better understanding of the uncertainties and limitations associated with different databases. By incorporating the most up-to-date and accurate data into the depletion subroutine DEPTH of the Monte Carlo code RMC, the performance and reliability of the simulation tool were significantly enhanced. The updated database now includes additional information such as decay radiation spectra, energy-dependent fission product yields, and subgroup neutron reaction cross-section parameters. These improvements lay a solid foundation for conducting more refined research on nuclear fuel cycles and supporting advanced reactor design and analysis.