Compared to second- and third-generation nuclear power systems, the Generation IV fission and future fusion reactors have higher service temperatures and irradiation doses, as well as harsher corrosive conditions and complex alternating loads. The structural materials for advanced reactors need to be researched and developed further. The oxide dispersion strengthened (ODS) steel has excellent high-temperature performance and irradiation resistance and is considered a promising structural material for advanced nuclear power systems. To reveal the effect of aging on microstructure and mechanical properties of ODS steel at near-service temperatures, the evolution of carbide M₂₃C₆ and nano-oxide particles (NPs) as well as the changes in the mechanical properties of 9Cr ODS steel after aging at 700^oC for varying durations were studied using SEM, TEM, and tensile testing. M₂₃C₆ rapidly precipitated along grain boundaries, gradually aggregated, and grew during early aging (≤ 200 h). While the NPs showed no noticeable change. During the midstages of aging (200 and 1000 h), NPs and carbides grew stably. In the later stages of aging (2000 and 3000 h), carbide particles grew to the micron scale. The average size and number density of the NPs tended to be stable. Compared to the initial 9Cr ODS steel, the growth rate of the average size was 19.7%, and the reduction rate of the number density was 27.1%. Dislocation cells and recovered subgrains appeared within some grains because of the pinning effect of NPs on continuous proliferation dislocations. The tensile strength rapidly decreased at the initial stages of aging. In the intermediate and later stages of aging, although the average size of the NPs increased and the number density decreased, its pinning effect was still prominent. Continuous proliferation dislocations were observed in the matrix, so the tensile strength remained stable. Furthermore, the tensile elongation was low during aging time of 1000 and 2000 h.