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The possibility of producing UO2 powder from UO3 using microwave radiation (power 300 W) in the presence of organic compounds with amino groups: carbohydrazide (CH), acetohydroxamic (AHA), and aminoacetic (glycine) acids has been investigated. It was found that in these processes in an oxygen-free atmosphere, UO2 is produced from UO3, and in an air, U3O8. It was shown that under the action of microwave radiation in the presence of CH and AHA, powdered UO2 are synthesized from UO3. The physical properties of the powders obtained (bulk density with tapping 2.6–2.7 g/cm3, specific surface area up to 3.2 m2/g, moisture content less than 0.1 wt %) meet the requirements for ceramic-quality powders in the production of fuel pellets.

Studies have been carried out in order to develop new ways to separate Am and Cm in nitrate solutions. It was shown that Am(VI) produced by oxidation of Am(III) by sodium bismuthate in 0.1 and 3.0 M HNO3 solutions is extracted with 30% tri-n-butyl phosphate solution in Isopar M diluent. Using this extractive agent in a mixture with a synergic additive of 0.1 M of trioctylphosphine oxide with 0.1 M HClO4 makes it possible to extract into the organic phase up to 90% of the starting amount of Am with not more than 3% of Cm contained in solution. It was found that, upon introduction of Na4XeO6·8H2O into a 0.1 M HNO3 solution containing Am(III) and NaBiO3, the solution becomes alkaline (рН ~ 10) and Am(III) is oxidized to Am(IV) to give a stable complex of composition Am(IV)·XeO6. As a result, Am remains in solution. Sodium bismuthate present in solution is hydrolyzed to give the solid phase Bi2O5 by coprecipitation of hydrolyzed Cm(III). Thus, the redox separation of Am(IV) from Cm(III) in the solution formed as a result of the interaction Am(III) with a 0.1 M HNO3 solution with sodium bismuthate and perxenate it contains is a simpler and more effective way, compared with the developed extractive method.

Denitration of nitric acid uranium solutions under the action of UHF radiation in ambient and reducing atmosphere in the presence of organic reductants containing amino groups (carbohydrazide, acetohydroxamic acid, aminoacetic acid, hydrazine nitrate) and without them to obtain a mixture of uranium oxides was studied. The conditions of thermal transformation of the initially formed mixture of uranium oxides into uranium dioxide powder under the action of UHF radiation were determined. The characteristics of UO2 powders meet the requirements of TU (Technical Specification) 95414–2005 to ceramic-grade powders.

It was established that heating to 90 °C of nitrate solutions of U, Np and Pu in the presence of hydrazine hydrate results in the formation of hydrated dioxides of these elements. On ignition under inert or reducing conditions in the temperature range of 280–800 °C hydrated uranium dioxide transmogrify into crystalline UO2. On ignition in air atmosphere UO2·nH2O turns into UO3 at 440 °C and into U3O8 at 570–800 °C. It was shown that thermolysis of the solution containing a mixture of uranium, neptunium and plutonium nitrates at 90 °C in the presence of hydrazine hydrate allows one to prepare hydrated dioxides (U, Np, Pu)O2·nH2O which on heating to ~300 °C transmogrify into crystalline product of UO2, NpO2 and PuO2 solid solution. The technique of preparation of solid solutions of U and Pu dioxides is very promising as simple and effective method of production of MOX-fuel for.

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.

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.

Microwave-assisted dissolution of ceramic uranium dioxide in tri-n-butyl phosphate (TBP)–HNO3 complex was investigated. The research on dissolution of ceramic uranium dioxide in TBP–HNO3 inclusion complex under microwave heating showed the efficiency of the use of this method. Nitric acid present in the inclusion complex participates both dissolution of UO2, and oxidation of U(IV)–U(VI), the resulting UO2(NO3)2 extracted with tri-n-butyl phosphate. Dissolution rate depends on both temperature of microwave dissolution process, and concentration of nitric acid present in the inclusion complex. The most intensive dissolution process is when the concentration of nitric acid ≥2 mol/L and the temperature of 120 °C. From the experimental data obtained by two kinetic models activation energies were calculated. At the average activation energy of UO2 dissolution in TBP–HNO3 complex equal 70 kJ/mol, and reaction order is close to one, i.e. the reaction takes place in an area close to kinetic.

Polymeric Pu(IV) at concentrations from 10−5 to 10−8 M in solutions with pH ∼8, i.e., under the conditions close to those in natural waters, does not disproportionate, and its polymeric species are stable. If Pu(VI) is present in solution under these conditions simultaneously with polymeric Pu(IV), it is incorporated in polymeric Pu(IV) species. Without polymeric Pu(IV), Pu(VI) under similar conditions undergoes hydrolysis to form intrinsic polymeric species in solution.

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 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.

Polymeric Pu(IV) in aqueous solutions in the pH range 0.5–3 disproportionates with time to form Pu(III) and Pu(VI). The arising Pu(III) is bound by hydroxyl groups of polymeric Pu(IV) and does not exhibit intrinsic absorption bands in the spectrum of a solution of polymeric Pu(IV). However, after ultrafiltration of the solution through a filter with a pore size of ∼3 nm Pu(III) is clearly identified in the filtrate by its absorption maxima. Pu(VI) occurs in the solution in the ionic state and is not bound by hydroxy groups of polymeric Pu (IV). Therefore, Pu(VI) is identified in the solution absorption spectrum both before ultrafiltration and after it. Thus, storage of solutions of polymeric Pu(IV) with pH 0.5–3 is accompanied by formation of Pu(III) and Pu(VI) ions.

The sorption of Pu(IV), polymeric Pu(IV), Pu(V) and Pu(VI) from the 0.1 M NaClO4 solution onto multiwalled carbon nanotubes was investigated. The kinetic study of the sorption process have shown that the polymeric Pu(IV) has the highest sorption rate, while decrease of sorption rate for plutonium aqua-ions in the order Pu(VI) > Pu(IV) > Pu(V) was found. Strong dependence of sorption kinetics of ionic plutonium species on pH was shown, in contrast to polymeric species, that were shown to quantitatively sorb (99%) in the wide pH range (pH 2–10). Two different sorption mechanisms for ionic and polymeric plutonium species were proposed: on the bases of sorption isotherms chemisorptions of plutonium aqua-ions onto carbon nanotubes and through intermolecular interaction for the polymeric plutonium species was defined. Distribution coefficients of plutonium in various oxidation states were found to increase with pH, showing the highest values for polymeric plutonium sorption (K d = 2.4 × 105 mL g−1 at pH = 6).

After storage of Pu(IV) hydroxide for more than 4 months, ∼90% of this compound polymerizes, the remainder (∼ 10%) being weakly polymerized Pu(IV). In 0.01 M NaCl solutions (pH ∼4–10) being in equilibrium with mixed or polymeric Pu(OH)4 (decantates), plutonium is mainly in the form of highly polymerized colloidal particles of molecular weight exceeding 100 kDa. Therefore, the Pu concentration in the solutions prepared by decantation or centrifugation of decanted solutions can range from 10−4 to 10−7 M. The content of weakly polymerized Pu in solutions varies from 10−7 to 10−9 M and depends on pH of the solution in the range 4–6. This dependence is virtually absent at pH 6–10.

Sorption of Np(VII), Np(VI), and Np(V) from 1 M NaOH by complexing fibrous “filled” sorbents was examined. POLIORGS 33-n and 34-n sorbents containing amidoxime and hydrazidine groups efficiently recover Np in the highest oxidation states and exhibit good kinetic properties. During sorption, Np(VII) is reduced with the sorbent to Np(VI) having a higher, under the actual conditions, distribution coefficient. The distribution coefficients, ml g−1, were estimated at 4.6 × 103 for Np(VII), 1.4 × 104 for Np(VI), and 3.6 × 102 for Np(V).

The kinetics of U(VI) accumulation in the phase of U(IV) hydroxide and of Np(V) in the phase of neptunium(IV) hydroxide, and also the solubility of the formed mixed-valence U(IV)-U(IV) and Np(IV)-Np(V) hydroxides in simulated groundwater (SGW, pH 8.5) and 0.1 M NaClO4 (pH 6.9) solutions was studied. It was found that the structure of the mixed U(IV–VI) hydroxide obtained by both oxidation of U(IV) hydroxide with atmospheric oxygen and alkaline precipitation from aqueous solution containing simultaneously U(IV) and U(VI) did not affect its solubility at the U(VI) content in the system exceeding 16%. The solubility of mixed-valence U(IV–VI) hydroxides in SGW and 0.1 M NaClO4 is (3.6±1.9) × 10−4 and (4.3 ± 1.7) × 10−4 M, respectively. The mixed Np(IV–V) hydroxide containing from 8 to 90% Np(V) has a peculiar structure controlling its properties. The solubility of the mixed-valence Np(IV–V) hydroxide in SGW [(6.5 ± 1.5) × 10−6 M] and 0.1 M NaClO4 [(6.1±2.4) × 10−6 M] is virtually equal. Its solubility is about three orders of magnitude as high as that of pure Np(OH)4 (10−9–10−8 M), but considerably smaller than that of NpO2(OH) (∼7 × 10−4 M). The solubility is independent of the preparation procedure [oxidation of Np(OH)4 with atmospheric oxygen or precipitation from Np(IV) + Np(V) solutions]. The solubility of the mixed-valence Np hydroxide does not increase and even somewhat decreases [to (1.4±0.7) × 10−6 M] in the course of prolonged storage (for more than a year).

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.

The pyrochlore-type matrices containing up to 10% wt. of U, Np, Pu and Am oxides were produced by self-propagating high-temperature synthesis (SHS) proposed as a method of preparation of the ceramic mineral-like matrices for immobilization of actinides. It was shown that basically the matrices consist of a phase of yttrium titanate having pyrochlore structure with actinides incorporated isomorphically. Stability of these matrices relative to hydrothermal leaching of the actinides at 90 °C was studied: the leaching rate does not exceed 1 × 10−7 g/(cm2 day).

Pu-bearing zirconolite and pyrochlore based ceramics were prepared by melting under oxidizing and reducing conditions at 1550 °C. 239Pu content in the samples ranged between ∼10 and ∼50 wt.%. Phase composition of the ceramics and Pu partitioning were studied using X-ray diffraction (XRD) and scanning electron microscopy with energy dispersive system (SEM/EDS). Major phases in the samples were found to be the target zirconolite and pyrochlore as well as a cubic fluorite structure oxide. Normally the Pu content in the Pu host phases was 10–12 wt.%. This corresponds to the Pu content recommended for matrices for immobilization of excess weapons plutonium. At higher Pu content (up to 50 wt.%) additional phases, such as a PuO2-based cubic fluorite-structured solid solution, perovskite, and rutile were found.

Extraction of californium/III/ with di-/−2-ethylhexyl/ phosphoric acid /D2EHPA/ in heptane from pyrophosphate media is almost quantitative between pH 4 and 5. From tripolyphosphate media, however, two to three extractions are needed in the pH range of 3–5 to isolate Cf3+ completely. Reextraction experiments show that 1M H2SO4 can back-extract Cf3+ completely while two to three reextractions with 5M HNO3 can only separate californium/III/. Reverse phase partition chromatography experiments were performed to recover 300 μg of californium/III/. From slope analysis of the extraction data the composition of the extracted species has been found to be Cf/H2P2O7/A.HA and Cf/H4P3O10/.A2.2HA from pyrophosphate and tripolyphosphate solutions, respectively, where D2EHPA is abbreviated as /HA/2.