To evaluate the Pu partitioning behavior in a feed solution with a high concentration of Pu during the acid split process, a countercurrent experiment was carried out with a dissolver solution derived from irradiated fast reactor fuel. The experimental flowsheet was optimized to prevent Pu polymerization and third phase formation in the Pu partitioning section, and these phenomena were not observed during the operation of the centrifugal contactors. In the Pu partitioning section, some of the U and almost all the Pu was stripped with 0.3 mol dm(-3) HNO3 at 16 degrees C, and the Pu content in the U/Pu product was 37.8%. The Pu content in the U/Pu product was 1.51 times higher than that in the feed solution. To further increase the Pu content in the U/Pu product, it might be necessary to strip Pu at a lower temperature. Concerning the decontamination performance, Cs-137 was decontaminated well in the product.