Abstract: In order to meet the need of Chinese Demonstration Fast Reactor and solve the problem that former calculated cross sections being smaller than the real values, it is necessary to develop a new method of calculating pseudo-fission products data, thus providing a foundation for generating complete neutron data of pseudo-fission products. The cross sections, angular distributions and double differential cross sections of pseudo-fission products were calculated by concentration-weighted summation. In the process of selecting nuclides, contribution method was proposed. The averaged fission yields and the absorption cross sections (reaction channel MT=27) were used to calculate the contributions of fission products to reactor, which quantifies the process of selecting nuclides and improves the accuracy. Finally, a complete pseudo-fission products data set for ²³⁵U was generated, using CENDL_NP library as main data source and TENDL library as supplementary. By comparing these data with former calculated results, the superiority and practicability of above method are validated.