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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. [ABSTRACT FROM AUTHOR]

A procedure was developed for preparing powders of solid solutions of uranium dioxide with 3 or 10 wt % Ce (as Am surrogate) from nitric acid solutions using microwave radiation. The powders obtained consist of particle aggregates of size no larger than 400 µm; the fraction of particles of size smaller than 25 µm does not exceed 1 wt %. The tap density of the powders is 2.3–2.5 g cm−3, and their specific surface area is 2.2–2.5 m2 g−1. The powder characteristics meet the requirements to powders of ceramic quality for nuclear fuel fabrication. The method developed can be used for producing mixed U-Am oxides on a unit for spent nuclear fuel reprocessing at the Pilot Demonstration Power Engineering Complex with the aim of Am transmutation in the BREST-OD-300 reactor. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Low-temperature mineral-like magnesium potassium phosphate (MPP) compounds were synthesized in the course of immobilization of nitric acid solutions containing cesium, strontium, sodium, ammonium, lanthanum, and iron as simulated radioactive waste (RW). The phase composition and structure of the compounds and the distribution of the RW components were studied. The mechanical strength (15 ± 3 MPa), heat resistance (up to 450°C), and radiation resistance (absorbed dose 1 MGy) of the compounds were evaluated in accordance with the existing regulations. The MPP compound exhibits high hydrolytic durability: The differential leach rate of 239Pu and 152Eu on the 28th day, measured in accordance with GOST (State Standard) R 52 126-2003, is 2.1 × 10-6 and 1.4 × 10-4 g cm-2 day-1, respectively. Introduction of wollastonite into the compound decreases the radionuclide leach rate by a factor of up to 5. The MPP compound shows promise for industrial solidification of liquid RW, including high-level highly saline multicomponent actinidecontaining waste. [ABSTRACT FROM AUTHOR]

Trivalent transplutonium (TPE) and rare earth (REE) elements are extracted to more than 80% with 30% TBP in Isopar M from solutions containing 0.06-0.5 M HNO3 and a salting-out agent, NH4NO3, in a concentration of =6 M. The elements are stripped from the organic phase with 0.1 M HNO3. The Am(III)/Eu separation factors vary from 1.8 to 2, which can be used for their extraction separation. The Cm/Am(III) separation factors in 0.06-3 M HNO3 are in the range from 1.1 to 1.2; therefore, to separate these elements, higher oxidation states of Am, Am(VI) and Am(V), should be used. The effect of various factors on the stability of Am(VI) was examined, and the conditions of the existence of Am(VI) and Am(V) in =0.1 M HNO3 solutions containing ~8 M NH4NO3 were determined. Curium is extracted with 30% TBP in Isopar M virtually completely, whereas americium only partially (=30%) passes into the organic phase in the form of Am(III). In the process, high degree of separation of Cm from Am(V) remaining in the aqueous phase is reached (=99.9%). [ABSTRACT FROM AUTHOR]

Dissolution of UO, UO, and solid solutions of actinides in UO, 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) typical of process solutions (6–8 M HNO, ∼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. [ABSTRACT FROM AUTHOR]