Journal of Radioanalytical and Nuclear Chemistry

, Volume 296, Issue 2, pp 1127-1131

First online:

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New approaches to reprocessing of oxide nuclear fuel

  • B. F. MyasoedovAffiliated withVernadsky Institute of Geochemistry and Analytical Chemistry, Russian Academy of Sciences
  • , Yu. M. KulyakoAffiliated withVernadsky Institute of Geochemistry and Analytical Chemistry, Russian Academy of Sciences Email author 


Dissolution of UO2, U3O8, and solid solutions of actinides in UO2 in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL−1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.


Spent oxide nuclear fuel Dissolution Ferric nitrate solution Fission products