The calculation of the spent-fuel composition is the baseline problem of the analysis of the nuclear and radiological safety of objects with spent fuel assemblies. The use of fission product libraries formed on the basis of evaluated nuclear data with an enlarged number of energy points is especially important for highprecision calculations of the nuclide composition of fuel irradiated in a fast spectrum. The results of an analysis of the nuclide composition of mixed oxide and uranium-plutonium-zirconium metal fuel with burnup ~10% h.a. in a fast spectrum, which were obtained in high-precision calculations based on data from different fission product yield libraries, are presented. The use of new types of fuel in fast reactors necessitates more accurate and reliable prediction of the characteristics of spent fuel for the analysis of the nuclear and radiological safety of objects containing spent nuclear fuel. The accuracy of high-precision calculations largely depends on the reliability of the neutron-physical data libraries used. In addition, the libraries of the transport cross sections, fission product yields, and decay and the small-group library of the cross sections of transmutation transformations must be formed on a unified computational platform with matched physical models and methods of preparation, making it possible to validate the correctness and accuracy of the data [1]. To compile libraries of fission product yields, it is necessary to take account of the factors associated with different physical models of the representation of the initial data, for example, models of triple and double fission, the criteria for determining the main disintegration isotopes taking account of the excited states of the nuclides, and others [2, 3]. The evaluated nuclear data with an expanded energy grid of the yield of fission products for energies above 1 MeV and intermediate energy are needed primarily for the formation of fission yield libraries for systems with a fast spectrum [4‐6]. More precise determination of the composition of spent fuel with the reckoned maximum number of isotopes is important for problems associated with its reprocessing and the fabrication of recycled fuel. Higher computational accuracy of the nuclide composition of fuel during burnup is also needed in connection with the use of individual isotopes, for example, 137 Cs, 106 Ru, and 154 Eu, as indicators for evaluating the neutron-physical characteristics of spent fuel. The nuclide composition of spent fuel in a fast spectrum, calculated using the codes MONTEBURNS‐MCNP5‐ ORIGEN2 and different fission product yield libraries, are analyzed in the present work. Comparative results are presented for mixed uranium-plutonium oxide and uranium-plutonium-zirconium metal fuels with burnup ~10% h.a. [7]. Nuclear Data on Fission Product Yields. In evaluated nuclear data files, the fission product yield is conventionally presented for three points ‐ thermal (E = 2.53·10 ‐8 MeV), average point of the fission spectrum (E = 0.4 or 0.5 MeV), and