The recovery of neptunium values from the Purex Process for reprocessing irradiated nuclear reactor fuel is improved by adding a rate-accelerating material to increase the rate of oxidation of the neptunium in the nitric acid solution in the first extraction column. The rate-accelerating material is formed by adding 1-nitropropane to a sodium hydroxide solution to form the aci-form of nitropropane, then adding sodium nitrite and acidifying the solution by the addition of nitric acid.
An improved Purex wet recovery process including the step of extracting and separating uranium and plutonium simultaneously from the fission products in the presence of nitric acid and nitrous acid by using a multistage extractor unit having an extracting section and a washing section is provided for separating and recovering neptunium simultaneously with uranium and plutonium contained in spent nuclear fuel. The improved method comprises the steps of maintaining the nitrous acid concentration in said extracting section at a level suited for effecting oxidation of neptunium from (V) to (VI) valence, while lowering the nitrous acid concentration in said washing section so as to suppress reduction of neptunium from (VI) to (V) valence, and maintaining the nitric acid concentration in said washing section at a high level.
