S. A. Perevalov

Properties of Host Matrices with Incorporated U and Pu Oxides, Prepared by Melting of a Zircon-Containing Heterogeneous Mixture (by Virtue of Exo Effect of Burning Metallic Fuel)

Host matrices with incorporated U and Pu oxides are obtained by melting of a zircon-containing heterogeneous mixture by virtue of exo effect of burning metallic fuel and are characterized by chemical analysis, spectrophotometric and radiometric methods, luminescence, X-ray microanalysis, and atomic emission ICP analysis. The material balance with respect to the incorporated radionuclides is preserved. The radionuclide distribution throughout the bulk of the matrix is nearly uniform. Metallic inclusions based on V, Fe, Si, and Mn, but containing no U and Pu, are found in the matrix. The investigated matrices are quite stable even under hydrothermal conditions (250°C, ∼30 atm): the leachability of U and Pu was determined to be 0.1-0.2 and 0.03 ppm, respectively, and that of Zr, Mn, and Fe, > 0.06 ppm.

New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO2, U3O8, and solid solutions of actinides in UO2, including those containing Cs, Sr, and Tc, in weakly acidic (pH 0.9–1.4) aqueous solutions of Fe(III) nitrate or chloride was studied. Complete dissolution of the oxides is attained at a molar ratio of Fe(III) nitrate or chloride to uranium of 1.6 or 2.0, respectively. In the process, actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). At 60°C, actinide oxides dissolve in these media faster than at room temperature. In the solutions obtained, U(VI) and Pu(III) are stable both at room temperature and at elevated temperatures (60°C), and also at high U concentrations (up to 300 mg ml−1) typical of process solutions (6–8 M HNO3, ∼60–80°C). After the oxide fuel dissolution, U and Pu are recovered from the solution by peroxide precipitation. In so doing, the content of Fe, Tc, Cs, and Sr in the precipitate does not exceed ∼0.05 wt %. From the solution after the U and Pu separation, the fission lanthanides, Tc, Cs, and Sr can be recovered by precipitation of Fe hydroxides in the presence of ferrocyanide ions and can be immobilized in appropriate matrices suitable for long-term and environmentally safe storage.

Use of microwave radiation for preparing uranium oxides from its compounds

The formation of uranium oxides by thermal decomposition of uranyl diaquadihydroxylaminate monohydrate, ammonium diuranate, ammonium tricarbonatouranylate, and uranium peroxide under the action of microwave (MW) radiation was studied. Uranium dioxide is formed by decomposition of these compounds in a reducing atmosphere at the MW radiation power of 600 W and treatment time of 5–10 min. In air, under the same conditions, U3O8 is formed. Under the action of MW radiation, substandard ceramic pellets of UO2 fuel can be readily converted in air to powdered U3O8. The use of MW radiation for thermal decomposition of uranium compounds allows the power and time consumption to be considerably reduced relative to the process with electrical resistance furnaces. A quick method for gravimetric testing of the composition of uranium oxides (UO2 or U3O8) using MW radiation was suggested.

Preparation of Np, Pu, and U dioxides in nitric acid solutions in the presence of hydrazine hydrate

Heating of nitric acid solutions of Np and Pu (∼90°C) in the presence of hydrazine hydrate (HH) leads to the formation of their hydrated dioxides in solution, transforming into crystalline dioxides at 300°C. Thermolysis of a mixed solution of U, Np, and Pu nitrates under the same conditions initially yields hydrated (U,Np,Pu)O2·nH2O, which on heating in air to ∼300°C transforms into a crystalline solid solution of (U,Np,Pu)O2. This method for stabilization of U dioxide in the presence of Pu in an oxidizing atmosphere can be used for preparing (U,Pu)O2 solid solutions of variable composition. This procedure shows doubtless prospects as a simple, efficient, and relatively low-temperature method for the production of MOX fuel for fast reactors.

Behavior of fission products in the course of dissolution of simulated spent nuclear fuel in iron nitrate solutions and of recovery of uranium and plutonium from the resulting solutions

Experiments aimed to examine the spent nuclear fuel dissolution in iron(III) nitrate solutions and to elucidate the behavior of fission products in the process were performed with simulated fuel corresponding to spent nuclear fuel of a WWER-1000 reactor. In Fe(III) nitrate solutions, U is quantitatively transferred from the fuel together with Cs, Sr, Ba, Y, La, and Ce, whereas Mo, Tc, and Ru remain in the insoluble precipitate and do not pass into the solution, and Nd, Zr, and Pd pass into the solution to approximately 50%. The recovery of U or jointly U + Pu from the solution after the dissolution of oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.

Preparation of uranium oxides in nitric acid solutions by the reaction of uranyl nitrate with hydrazine hydrate

UO2·nH2O formed by thermal denitration of uranyl nitrate in solutions under the action of hydrazine hydrate can be converted in air to UO3 at 440°C and to U3O8 at 570–800°C, and also to UO2 in an inert or reducing atmosphere at 280–800°C. After the precipitation of hydrated uranium dioxide, evaporation of the mother liquor at 90°C in an air stream allows not only evaporation of water, but also complete breakdown and removal of hydrazine hydrate and NH4NO3. The use of microwave radiation considerably reduces the time required for complete thermal denitration of uranyl nitrate in aqueous solution to uranium dioxide, compared to common convective heating.

Sorption of plutonium in various oxidation states from aqueous solutions on Taunit carbon nanomaterial

Sorption of Pu from weakly acidic and weakly alkaline solutions on Taunit carbon nanomaterial was studied. Under these conditions, both polymeric Pu(IV) and ionic Pu(V, VI) species are recovered from freshly prepared solutions. Also, Pu is efficiently sorbed from simulated groundwater after more than 10 months of storage. The Pu sorption in all the forms by carbon nanotubes is rapid and almost quantitative (95 ± 5%) at the sorbent-to-solution ratio of 1 : 80 g ml−1. Plutonium preliminarily sorbed on Taunit can be efficiently immobilized in a magnesium potassium phosphate ceramic whose physicochemical properties meet the requirements of prolonged environmentally safe storage of long-lived radionuclides.

Sorption of Pu(IV) in the polymeric colloidal form on rock typical of Mayak Production Association area

Sorption of colloids of polymeric Pu from simulated groundwater on a rock typical of Mayak Production Association area was studied. In 20 days, polymeric Pu with the particle size exceeding 220 nm is 99% sorbed by the rock with the distribution coefficient Kd = 1880. Desorption performed for 5 days allows no more than 40% of the sorbed Pu to be transferred into the solution, even with such strong complexing agents as 0.05 M hydroxyethylidenediphosphonic acid in 0.1 M HNO3 and 0.1 M Tamm solution.

Factors governing the efficiency of dissolution of UO2 ceramic pellets in aqueous solutions of iron nitrate

Dissolution of ceramic UO2 in aqueous Fe(NO3)3 solutions at different temperatures under the conditions of limited contact with air and in the autoclave mode was studied. In the course of UO2 dissolution at 60–90°C, the U/Fe molar ratio appears to be ∼1, whereas at room temperature (25°C) this value is ∼0.5. By varying the acidity of Fe nitrate solutions at these temperatures, it is possible to increase the U/Fe molar ratio to ∼4 and to obtain uranyl nitrate solutions with simultaneous removal of Fe from the solution in the form of a precipitate of the basic salt, or to perform quantitative dissolution of UO2 under the conditions excluding the formation of such precipitate. In the course of dissolution of ceramic UO2 in Fe(NO3)3 solutions, the appearance or absence of Fe(II) ions, the formation or absence of the precipitate of the Fe basic salt, and variation of solution pH are interrelated and are determined by the process temperature.

Carbon nanotubes: Potential uses in radionuclide concentration

The review of literature data related to the preparation, properties, and application of carbon nanotubes for sorption recovery of elements is given. Experimental data on the application of Taunit carbon nanofor radionuclide preconcentration from different solutions, as well as of Taunit-based solid-phase extractants for recovery of actinides and rare-earth elements from nitric acid solutions are presented.