S. E. Vinokurov

Phase composition, structure, and hydrolytic durability of glasses in the Na2O-Al2O3-(Fe2O3)-P2O5 system at replacement of Al2O3 by Fe2O3

Samples of sodium aluminum iron phosphate glasses of the composition (mol %) 40 Na2O, (20 − x) Al2O3, x Fe2O3, 40 P2O5 (series I) and 35 Na2O, (20 − x) Al2O3, x Fe2O3, 45 P2O5 (series II) were synthesized. The phase composition and structure of the samples obtained were determined by X-ray diffraction and IR spectroscopy. At equimolar replacement of Al2O3 by Fe2O3, the structure of the quenched glasses of series I does not change appreciably, in contrast to glasses of series II. Annealing of the glasses leads to their partial devitrification with segregation of crystalline aluminum iron phosphate phases. Glasses of series I with up to 10 mol % Al2O3 replaced by Fe2O3 exhibit the highest hydrolytic durability: The leach rates of Na, Al, Fe, and P from the samples are within (4–10) × 10−8 g cm−2 day−1, meeting the requirements of GOST (State Standard) R 50 926-96. Thus, glasses with approximately equal molar concentrations of Al2O3 and Fe2O3 are the most resistant to crystallization and hydrolysis.

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.

Recovery of rare earth elements, uranium, and thorium from monazite concentrate by supercritical fluid extraction

Quantitative recovery of rare earth elements (REEs), Th, and U by supercritical fluid extraction (SCFE) with carbon dioxide containing adducts of TBP and HDEHP with HNO3 directly from monazite concentrate (MC) powder is impossible and requires the conversion of the constituent elements into more soluble compounds. Microwave (MW) radiation can be efficiently used for MC pretreatment by sintering with Na2CO3 in the presence of coal. The resulting product consists of two phases. One of them contains REEs (∼50%) recoverable by supercritical carbon dioxide (SC-CO2) containing adducts of TBP or HDEHP with HNO3. The second phase is a solid solution of CeO2 with Th and U oxides and remaining amount of REEs. It is resistant to SCFE. Conditions were determined for quantitative dissolution of this phase in a mixture of 4 M HCl with 0.05 M HF. The use of HDEHP under the SCFE conditions allows quantitative recovery of Th and U from the hydrochloric acid solution. In the process, REEs remain in the aqueous phase and are thus separated from Th and U. A possible flowsheet was suggested for the recovery REEs from MC using SCFE with their simultaneous separation from Th and U.

Preparation of powdered uranium oxides by microwave heating of substandard ceramic pellets of oxide nuclear fuel

Microwave (MW) heating of substandard ceramic UO2 pellets in air allows their rapid conversion into powdered U3O8, from which UO2 can be obtained again in a reducing atmosphere. Comparative analysis of the physicochemical and technological properties of the U3O8 and UO2 powders obtained under the action of MW radiation with the industrial (standard) powders demonstrated their suitability for fabricating fuel pellets. The power consumption for MW heating appears to be lower by an order of magnitude than the power consumption for performing similar operations with electric resistance furnaces.

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.

Magnesium Potassium Phosphate Compound for Radioactive Waste Immobilization: Phase Composition, Structure, and Physicochemical and Hydrolytic Durability

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°С), 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.

Chemical-technological and mineralogical-geochemical aspects of the radioactive waste management

This paper considers various matrices that are able to incorporate components of radioactive wastes (RAW) of different origin. It is noted that attempts to develop the single phase crystalline matrix to immobilize all RAW components failed. The only single phase matrix brought to the industrial application is glass, which is able to accumulate practically all RAW components but in limited concentrations. Prospects are related with some types of ceramics for immobilization of narrow fractions of RAW or individual radionuclides (for instance, minor actinides), as well as some types of low-temperature matrices (iron-phosphate, magnesium–potassium–phosphate, and geopolymers). Approaches to choosing the technology of waste form synthesis are considered. Perspectives of application of both high-temperature (cold-crucible induction melting, self-propagating high-temperature synthesis) methods and modified cementation technologies are demonstrated. It is noted that the final isolation of RAW from the biosphere suggests their disposal in underground repositories. The most difficult technical problem is the disposal of RAW containing long-lived radionuclides. It is shown that the quantitative assessment of repository safety with allowance for their characteristics and all possible processes and phenomena is required to substantiate the safe disposal of long-lived radionuclides.

Nuclear fuel cycle and its impact on the environment

In this paper, we consider the present-day situation and outlooks of the development of nuclear power generation in Russia and other countries. It was noted that the implementation of the concept of a closed nuclear-fuel cycle accepted in Russia relies on the solution of the problem of the disposal of spent nuclear fuel (SNF) and radioactive waste (RAW). This paper presents the main results of investigations focused on the development of radiation-safe methods of manufacturing nuclear fuel elements, including mixed uranium–plutonium oxide fuel for fast-neutron reactors; creation of low waste-production technologies of SNF processing and RAW disposal; and the analysis of fundamental features of the behavior and speciation of radionuclides in environmental objects for the development of efficient methods of radioecological monitoring and remediation of radionuclide-contaminated areas.

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.

Association of radionuclides with components of fulvic acids isolated from soils

The distribution of Pu and Am between specific and nonspecific components of various groups of the organic matter of soil was studied. For chernozem and soddy podzolic soil, 63 and 53% of Pu and only 1 and 8% of Am, respectively, are bound with fulvic acid purified to remove low-molecular-weight compounds by fractionation on BAU activated charcoal. Larger relative amount of Am, compared to Pu, in the form of low-molecular-weight nonspecific compounds is responsible for higher migration mobility of Am, compared to Pu, in the environment. The sorption of organic complexes of Pu and Am on carbon nanotubes was studied. The degree of Pu sorption only slightly depends on the nature of the organic substance, except solutions of humic acids from which the Pu sorption is appreciably higher. The degree of Am sorption regularly increases with a decrease in the molecular weight of the organic substances.