Abstract: Because the elastic scattering cross sections of medium-mass nuclides have very strong elastic scattering resonance phenomenon in the fast neutron region, which will lead to the self-shielding effect of energy scale and make distortions of the neutron flux distribution to zigzagging. In order to accurately describe the distribution of neutron flux in the energy scale, the deterministic one-step method needs to divide a large number of energy intervals for description, and the number of energy groups is generally thousands. Deterministic one-step transport calculation has large memory and low computational efficiency with thousands of groups. In the traditional calculation of the whole-core non-uniform transport calculation for the thousands of groups, the section generation software was mainly used to prepare the self-shielding cross section by using the ultra-fine groups, and then the broad-group cross section was prepared offline by the transport calculation of the equivalent model of cells or components. Finally, the deterministic one-step program was used to calculate the whole-core non-uniform transport. The broad-group cross sections prepared off-line cannot reflect the characteristics of materials of different reactor locations and bring the approximations. In order to solve the problem that the thousand-group energy structure of fast reactors leads to the high computational cost of the deterministic one-step method, an online energy group condensation method with fine group energy spectrum based on SHARK was developed. For the fuel cells of fast reactors, it can keep the composition of fuel and other materials unchanged and search the radius of the fuel to make the escape cross section conserved. It has the capability of online equivalent model establishment, online calculation of fine group energy spectrum and broad group cross sections condensation. For non-fuel cells, the broad group cross section was condensed using the fine group energy spectrum calculated by the one-dimensionalal constant volume model of practical problems. Based on the MOX1000 material of the benchmark for neutronic analysis of sodium-cooled fast reactor cores, a two-dimensional 5×5 rectangular single-assembly problem and a two-dimensional 3×3 multi-assembly problem were constructed and calculated. The numerical results show that compared with the calculation results of the group in 1 968, the acceleration ratio of the online energy group condensation method in the two-dimensional 5×5 rectangular single-assembly problem is 8 to 18 and that in the two-dimensional 3×3 multi-assembly problem is 12 to 17, which can significantly improve the calculation efficiency. In terms of calculation accuracy, the relative deviation of pin power distribution is less than 1%, and the maximum deviation of eigenvalue is about 360 pcm. It is of great significance to deterministic one-step method for fast reactor transport calculation.