Abstract: High flux reactors, as versatile irradiation facilities, have significant applications in various fields such as neutron scattering research, isotope production, and material irradiation testing. However, due to their unique design, different regions of the core exhibit markedly different energy spectrum characteristics. At the same time, two-dimensional full-core-scale high-order transport calculations require a large amount of memory. Therefore, due to current hardware limitations, it is not feasible to perform full-core-scale high-order transport calculations based on ultra-fine group structures. To address these issues, this paper proposes a one-step calculation scheme with online energy group condensation suitable for high flux reactors. In this paper, firstly, an ultra-fine group structure of 2165 groups was selected to complete resonance calculations on fine geometric modeling. Secondly, combined with the low-order scattering source term approximation, only a few source iteration calculations were performed to obtain the low-order approximate flux. Then, the low-order approximate flux and the ultra-fine group total cross-sections were used to calculate the approximate high-order flux and complete the multi-group high-order cross-section condensation. Finally, the high-order approximate flux and multi-group high-order cross-sections were used for multi-group high-order calculations to obtain eigenvalue information and flux distribution information. The method based on two high flux reactor designs was verified and the sensitivity of the selected multi-group energy structure to this calculation scheme was analyzed. The calculation results show that under the 172-group structure, the eigenvalue deviations obtained by using this calculation scheme for the two core designs are both below 136 pcm. At the same time, it can be seen that in this test problem, the fast/thermal neutron flux distribution and flux density peaks are insensitive to the energy group structure, and the relative errors of the fast/thermal neutron flux density peaks for both core designs are below 1.5%. The calculation accuracy of the irradiation channel flux shape is related to the energy group structure and the properties of the materials themselves. In this test, the relative errors of most group fluxes are below 5%. Based on this, it can be seen that the calculation scheme proposed in this paper can effectively solve the memory limitation problem in two-dimensional full-core-scale high-order transport calculations and is suitable for neutron numerical simulation calculations in high flux reactors with high computational accuracy.