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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.

Fabrication of ceramic UO2 fuel pellets using microwave radiation was studied. The UO2 powder was prepared by microwave denitration of a nitric acid solution containing 400 g L–1 uranium. The tapped density (2.39 g cm–3) and total specific surface area (2.70 m2 g–1) of the powder obtained met the requirements to the powder for nuclear fuel fabrication (TU (Technical Specification) 95 414–2005: Uranium Dioxide Powder of Ceramic Grade with the Uranium-235 Isotope Content Lower than 5.0%). Pellets were pressed from the UO2 powder under varied conditions including pressure, its application mode, pressing time, and presence of binder. The pressed pellets were sintered at 1650°С for 2 h in an Ar + 10 vol % H2 atmosphere under the action of microwave radiation. The density of the samples obtained, 10.40 ± 0.02 g cm–3, meets the requirements to ceramic fuel pellets used in thermal reactors.

The use of microwave heating to produce solid solutions of actinide oxides in the processes of thermal denitration of model nitric acid solutions formed in reducing stripping schemes at the final stages of the PUREX process, containing U and Th (Pu simulator) and unreacted reducing agents (hydrazine, etc.), is proposed. As a result, after quantitative distilling of water vapor, acid, and volatile products into the collector, denitration reductive thermolysis of the concentrate (melt) of actinide nitrates containing 100% of a solid solution of oxides U(Th)О2 takes place. The U–Th oxide powders obtained meet the regulatory requirements (TU 95414-2005) for ceramic powders.

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.

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.

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.

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%).

Supercritical fluid extraction (SCFE) using carbon dioxide containing tributyl phosphate (TBP), di-(2-ethylhexyl)phosphoric acid (D2EHPA) and their adducts with HNO3 is applied for extraction of rare earth elements (REE), thorium (Th) and uranium (U) from monazite concentrate (MC) and phosphogypsum (PG). REE extraction from MC and their separation from Th and U are carried out from the product of MC–Na2CO3 baking (MCS), which is obtained under microwave irradiation, after which the phosphates of REE, Th and U present in the MC are converted into their oxides. Up to 50% of REE can be recovered as the adducts of TBP and D2EHPA with HNO3 from the resulting powdered MCS under SCFE conditions, whereas Th and U remain in the solid phase. After a complete dissolution of the MCS residue in the mixture of 4 M HCl and 0.05 M HF, Th and U are quantitatively extracted using supercritical carbon dioxide (SC CO2) containing D2EHPA and thus separated from the REE that remain in an acidic solution. The conditions of quantitative isolation of REE, Th and U from PG are determined. The schemes for complex processing of MC and PG aimed at REE recovery and their separation from Th and U are suggested.

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.

It is demonstrated on real solutions of samples of spent nuclear fuel (SNF) from WWER-1000 reactors (1000-MWel water-cooled water-moderated energy reactors) that weakly acidic solutions of iron(III) nitrate at the molar ratio Fe(III): U ≥ 2.0 dissolve SNF with quantitative transfer of U and Pu into the solution. In the process, Fe partially precipitates in the form of a basic salt precipitate together with a part of the fission products (>90% of Ru, ~90% of Мо, >60% of Tc, and 40% of Zr) already in the step of the fuel dissolution. Cs, Eu, and Am pass into the solution together with U and Pu. With the required conditions followed, U and Pu can be separated from the solution by precipitation of their peroxides or quantitatively extracted from this solution with 30% TBP in Isopar L. The presence of ≥1 M Fe(NO3)3 in the solution considerably increases the distribution ratios of TPE and REE, which allows their recovery from a weakly acidic nitrate solution to be also performed with 30% TBP in a diluent. This process can serve in the future as a basis for the development of a new integrated technology combining the PUREX process with TPE partitioning using a common extractant.

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.

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.

The behavior of uranium and fission product simulators in the processing of model spent nuclear fuel in weakly acidic iron(III) nitrate solutions in the presence of supercritical CO2 containing tributyl phosphate (TBP) has been studied. It has been shown that these conditions provide extraction of uranium from iron oxide solution with simultaneous separation of it from existing fission products. The separation of uranium from the TBP phase is performed via its reextraction with an aqueous solution containing hydrogen peroxide and simultaneous reprecipitation in the form of a peroxide. In this case, a high degree of uranium purification from all the fission products is attained, thus allowing it to be reused in a nuclear fuel cycle.

The solubility of the uranyl nitrate complex with tributyl phosphate (TBP) in liquid and supercritical carbon dioxide (SC-CO2) has been measured as a function of temperature and pressure. This complex is formed upon the dissolution of uranium dioxide in the two-phase system consisting of aqueous iron(III) nitrate and TBP-saturated liquid or supercritical CO2. The kinetics of the dissolution of ceramic UO2 in such a two-phase solvent is studied. It has been shown that uranium can be extracted in the presence of TBP-containing supercritical CO2 via the dissolution of spent oxide nuclear fuel in subacidic iron(III) nitrate solutions with the simultaneous extraction of the uranium complex into the TBP-SC-CO2 phase.

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.

The kinetics of dissolution of ceramic UO2 in aqueous solutions of Fe(III) nitrate in the presence of TBP in supercritical CO2 was studied. Quantitative recovery of U into solutions of Fe nitrate in combination with its extraction from the solution into the fluid occurs under these conditions within ∼2 h, which is by more than an order of magnitude faster than in dissolution in a Fe(III) nitrate solution without TBP under common conditions. The behavior of uranium and simulated fission products (FPs) in reprocessing of simulated spent nuclear fuel (SSNF) in weakly acidic Fe(III) nitrate solutions using supercritical CO2 containing TBP was studied. Under these conditions, SSNF dissolves quantitatively with the simultaneous recovery of the U nitrate complex into the TBP-containing fluid phase. Uranium is recovered from the fluid phase by back extraction with an H2O2 solution with simultaneous precipitation of U in the form of the peroxide UO4·2H2O. In the process, high degree of uranium decontamination from all the FPs is reached, which allows repeated use of U in the nuclear fuel cycle.

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.

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.

Apparent formation constants of Pu(IV) and Th(IV) with two kinds of humic acids were determined in 0.1 M NaClO4at 25 ºC using a solvent extraction method with thenoyltrifluoroacetone in xylene. The acid dissociation constants of humic acids were also measured by potentiometric titration and used as the degree of dissociation for calculating the formation constants. The effect of solution conditions, such as the pH, the initial metal and humic acid concentrations, and the ionic strength, on the formation constants was examined. The obtained data were compared with the ones in the literature.

The sorption properties of solid-phase extractants (SPEs) prepared by impregnation of Taunit carbon nanotubes with adducts of diphenyl(dibutylcarbamoylmethyl)phosphine oxide (CMPO) and tri-n-octylphosphine oxide (TOPO) in HNO3 solutions were studied. The SPEs exhibit high ability to sorb U(VI), Pu(IV), Np(V), Am(III), and Eu(III) from nitric acid solutions, with good kinetic properties. The impregnation conditions and distribution coefficients of the radionuclides in their recovery from 3 M HNO3 were determined. The possibility of preparing SPEs by Taunit impregnation in HNO3 solutions with adducts of tributyl phosphate (TBP) and N,N′-dimethyl-N,N′-dioctylhexylethoxymalonamide (DMDOHEMA) and with Cyphos IL-101 phosphonium ionic liquid was demonstrated.

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.

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.

It was shown that, in contrast to the Purex process using aggressive and environmentally hazardous 8M HNO3 solutions for dissolving spent oxide nuclear fuel (SNF), this fuel can be easily dissolved in aqueous subacid ([H+] ∼0.1 M) solutions of Fe(III) nitrate (chloride) with partial separation of uranium and plutonium from fission products (FP). The low acidity of the solutions obtained (pH ∼1) allows direct application of modern technologies of finishing processing of nuclear fuel by fluoride, carbonate, oxalate, or peroxide precipitation of uranium and plutonium. It was established that U(VI) is isolated from nearly neutral nitric acid solutions as a poorly soluble uranyl hydroxylaminate complex after adding hydroxylamine. It was shown that on thermal decomposition at 200–300°C under ambient atmosphere this compound converts into uranium dioxide. A similar approach was applied to obtain mixed oxide uranium-plutonium fuel (MOX fuel).

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.

Addition of aqueous ammonia and hydroxylamine hydrochloride to a nitric acid solution containing U(VI) and Pu(VI) causes at pH ∼7 precipitation of a mixture of U(VI) hydroxylaminate and Pu(III) hydroxide. The precipitate separated from the mother liquor and dried at 60°C upon further calcination in air at 300°C transforms into a solid solution of PuO2 in UO2 as a result of thermal decomposition of the precipitate with the reduction of U(VI) to U(IV) by the ligand coordinated to the uranyl ion. The developed procedure for preparing mixed U-Pu dioxide can become an alternative to the presently used method for preparing uranium dioxide and mixed uranium-plutonium oxide fuel (MOX fuel) by heating at 800–900°C in a reducing atmosphere.

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 complex formation constants of tetravalent plutonium ion with malonic and succinic acids in aqueous solution were determined by the solvent-extraction method. Also, by taking the known values of the solubility products, the hydrolysis constants and the formation constants, the experimental solubility data of plutonium in the presence of carboxylates were analyzed.

Forms of occurrence of polymeric Pu(IV) in simulated groundwater (SGW) were studied spectrophotometrically and by the method of centrifugal ultrafiltration through filtering inserts permeable to polymeric Pu species with different molecular weights. The dependences of the fractions of definite Pu(IV) forms on the total Pu content in the solution were found. The possibility of formation of Pu(IV) quasipolymeric structures in aqueous solutions was considered in relation to the problem of the transfer of radioactive contaminants with underground water. Equilibrium distribution of Pu(IV) polymers depending on the total Pu(IV) concentration in the solution was analyzed theoretically. From the experimental data obtained, the parameter allowing determination of the weight distribution of the polymeric particles in relation to the total Pu(IV) concentration was theoretically calculated, and their equilibrium distributions depending on the total Pu(IV) concentration were found.

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.

Solubility of UO2, NpO2, and PuO2 powders in simulated groundwater (SGW) was studied by attainment of the equilibrium from below. The solubility was determined in AnO2-SGW heterogeneous systems at different solid: liquid weight ratios and different pH values of the examined solutions both under ambient conditions and in an inert atmosphere. The solubility of UO2, NpO2, and PuO2 depends on the solid: liquid ratio but is independent of pH of the solutions in the range from 6 to 11. The solubilities of UO2 and NpO2 in SGW under ambient conditions are higher that those in an inert atmosphere. The solubility of freshly calcined PuO2 is virtually independent of the atmosphere. Formation of amorphous hydroxides by oxidation of the dioxide surface, whose area is proportional to the weight of the dioxides, increases their solubility.

In the frame of development of improvements of current liquid–liquid extraction technologies for reprocessing spent nuclear fuel, and for development of effective schemes of HLW treatment, novel methods of recovery and separation of uranium, transuranium and rare-earth elements have been proposed. The methods of solvent-free extraction in supercritical CO2, countercurrent chromatography, and sorption by fibrous “filled” sorbents are discussed.

Supercritical carbon dioxide (SC-CO2) containing TBP-HNO3 adduct effectively recovers actinides from solid solutions of their oxides. The efficiency and mechanism of recovery of actinides from oxides are similar both for the adduct dissolved in SC-CO2 and for neat TBP-HNO adduct. SC-CO2 acts as a diluent which, after transportation of the recovered actinide complexes with TBP-HNO3 adduct, can be removed from the system or, if necessary, recycled.

The TBP-HNO3 and N,N′-dimethyl-N,N′-dioctylhexylethoxymalonamide-HNO3 adducts effectively recovery actinides from solid solutions of their oxides in UO2 matrix. The adduct of methyl isobutyl ketone and HNO3 selectively recovers uranium from PuO2-UO2 solid solution, separating uranium from plutonium. The optimal composition of TBP-HNO3 adducts for the recovery of actinides from the dioxide solid solutions was determined. Uranium is recovered with the adduct more rapidly at lower temperatures and smaller volume of the organic phase, as compared to the dissolution of the oxide fuel in aqueous nitric acid. This approach to the processing of spent nuclear fuel can substantially decrease the volume of highly toxic radioactive aqueous and organic wastes arising with the existing technologies.

The separation of U, Pu, and Am recovered from MOX fuel with the adduct of HNO3 with N,N′-dimethyl-N,N′-dioctylhexylethoxymalonamide by countercurrent chromatography (CCC) was studied. Solutions of N,N′-dimethyl-N,N′-dibutyldodecylethoxymalonamide in dodecane were used as stationary phase. The separation of U, Pu, and Am was carried in both isocratic and stepwise elution modes. The better separation of actinides and their higher radionuclidic purity are reached with stepwise elution. The first eluate fraction contained only Am (100%). The second eluate fraction contained U (100%) and Pu (0.7%). The third eluate fraction contained 99.3% of Pu.

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).

The article deals with the use of countercurrent chromatography (CCC), support-free partition chromatography, for separation of U and Pu from the organic extract obtained directly by the dissolution of MOX fuel in supercritical CO2 containing the complex TBP · nHNO3. White spirit solutions with various TBP concentrations were used as a stationary phase. The effects of the compositions of the stationary and mobile phases on the U/Pu partition efficiency are studied. The CCC method allows the separation of U and Pu under conditions of a TBP concentration gradient in the stationary phase and also of an HNO3 concentration gradient in the mobile phase. Chromatographic separation first gives the Pu fraction containing 98.9% of Pu and 0.07% of U, and then the U fraction (99.93% of U and 1.1% of Pu). The separation time is 50 min.

The Np(V) distribution coefficients between kaolinite and solutions of MgCl2 and CaCl2 were determined experimentally at various concentrations of the electrolytes and Np. The Np sorption decreases with increasing concentration of the supporting electrolyte. The sorption is completely reversible. The sorption equilibrium is attained in approximately one week after the start of the sorption-desorption experiments. The constants of NpO2+-Mg2+ and NpO2+-Ca2+ binary ion exchange on kaolinite were determined by fitting the experimental results with an ion-exchange equation for the restricted sorption capacity: \(\log K_{NpO_2^ + - Mg^{2 + } } = 1.26 \pm 0.08\) and \(\log K_{NpO_2^ + - Ca^{2 + } } = 0.96 \pm 0.10\). These constants describe well the experimental data at low Np concentrations (≤1×10−6 M). The ion-exchange capacity of kaolinite, calculated from the experimental data on Np sorption from solutions (3.03×10−4 g-equiv kg−1 MgCl2), somewhat differs from that in CaCl2 solutions (2.15×10−4 g-equiv kg−1).

The Np(V) distribution coefficients between montmorillonite and solutions of MgCl2 and CaCl2 were determined experimentally. The Np sorption decreases with increasing concentration of the supporting electrolyte. The sorption is very fast within the first several hours after the start of the experiment and then decelerates, reaching the equilibrium in approximately two weeks. The constants of binary ion exchange NpO 2 + -Mg2+ and NpO 2 + -Ca2+ on montmorillonite determined by fitting the experimental results are $$\log K_{NpO_2^ + - Mg^{2 + } } = 0.083 \pm 0.03$$ and $$\log K_{NpO_2^ + - Ca^{2 + } } = 0.106 \pm 0.014$$ .

The separation of Am(III) and Cm(III) by countercurrent chromatography (CCC) was achieved using the liquid phase systems "diamide–hydrogenated tetrapropylene (TPH)–HNO3" following the optimisation of: (i) the compositions of the mobile (HNO3 concentration) and the stationary (nature of the diamide and its concentration in TPH) phases, (ii) column length and (iii) operating parameters of the CCC (flow rate of the mobile phase and rotation speed of the coil column). The following diamide extractants have been studied: (i) N,N´-dimethyl-N,N´-dibutyltetradecylmalonamide (DMDBTDMA), (ii) N,N´-dimethyl-N,N´-dioctylhexyl-ethoxymalonamide (DMDOHEMA) and (iii) N,N´-dimethyl-N,N´-dibutyldodecylethoxymalonamide (DMDBDDEMA). It is shown that these diamides can be used for the separation of Am(III) and Cm(III) by CCC. Increasing the column length leads to an increase of the stationary phase retention on the column while improving the Am/Cm separation. Increasing the speed of rotation of the centrifuge from 660 to 950 rpm also results in increasing the stationary phase retention but does not influence the resolution of the Am/Cm separation. Decreasing the flow rate of the mobile phase from 1.0 to 0.5 mL/min leads to a better resolution of Am and Cm separation. The best Am/Cm separation was achieved with systems based on DMDBDDEMA and DMDOHEMA in TPH using a two-layer coil column and an isocratic elution mode. The application of CCC makes it possible to separate the elements within 100 min: the Cm fraction contains 99.5% of Cm(III) and 0.6% of Am(III) inventories and the Am fraction contains 99.4% of Am(III) and 0.5% of Cm(III).

Summary The reactions between Am or Eu present in 0.1 M nitric acid solutions, alone or both together, with 0.25 M K3Fe(CN)6, were studied at room temperature. When Am was the only trivalent metal ion in solution, precipitation of AmFe(CN)6 occurs and the residual Am concentration is about equal to 0.95 mM. However, when Am initial concentration is less than the above specified value for AmFe(CN)6 solubility, Am residual concentration measured is lower that its concentration in the initial solutions. The solubility of EuFe(CN)6 was found to be equal to 30 mM. However, when precipitate formation occurs, the Eu residual concentration after phase separation is about in average 8 mM. When Am and Eu were simultaneously present at the same concentration in solution, the solubility of Am differs little from that measured with Am alone. For initial Am concentrations below 1 mM, the solubility of Am is higher than that observed in the absence of Eu. For initial concentrations of about 8 mM, the Am solubility is lower (about 0.7 mM) than that observed for Am alone. When the initial Eu concentration is constant at 29 mM, the Am precipitation efficiency is much higher than observed in the absence of Eu, for Am concentrations between 30 and 0.5 mM. The residual Am concentratixon in solution thus drops considerably in the presence of Eu (29 mM), and is about 0.04 mM for initial Am concentrations below 4 mM.

Interaction of solid diphenyl(dibutylcarbamoylmethyl)phosphine oxide (Ph2Bu2) with perchloric and hydrochloric acid solutions was studied. On contact with the acids, this agent forms a liquid substance (“liquid reagent,” LR) exhibiting extractive power. The conditions of LR formation and extraction of Am(III) from HCl solutions with LR in the absence of organic diluent were studied. The solubility of Ph2Bu2 in 3 M HClO4 is 2.5 mg l-1. The composition of the LR formed in HClO4 solutions is 2Ph2Bu2·HClO4·nH2O. Perchloric acid is quantitatively backwashed from the LR phase with water. Formation of an Nd(III) complex with LR in HClO4 solutions is accompanied by cementation (solidification) of the compound. The resulting complex has the 3 : 1 [Ph2Bu2 : Nd(III)] composition and mp ∼120°C. According to the IR spectrum, the coordination of Ph2Bu2 with the HClO4 molecule is monodentate, and with the Nd(III) ion (in the absence of organic diluent) it is bidentate.

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.

Extraction of Am(III) from nitric acid solutions with a liquid complex of diphenyl(dibutylcarbamoylmethyl)phosphine oxide (Ph2Bu2) with nitric acid as influenced by concentration of Al, Ca, Na, Ni, Co, Cr, Fe, Zr, and Mo nitrates was studied. Nonextractable (Al, Ca, Na) and poorly extractable (Ni, Co, Cr) nitrates at their widely varied concentration do not interfere with the exhaustive extraction of americium from 3 M HNO3. These salts facilitate both formation of the extraction-active liquid phase on contact of Ph2Bu2 with 1 M HNO3 and exhaustive extraction of americium from the acid. Extraction of Fe(III) at its widely varied concentration insignificantly interferes with extraction of americium, but this effect increases with increasing acidity and phase contact time. Americium is extracted with a powdered reagent from 0.1 M HNO3 when liquid extractant is not formed in the visible amounts. In extraction of americium from weakly acidic solutions in the presence of salts the solid reagent, apparently, partially transforms into the liquid extractant, and two different extraction mechanisms are operative. It was shown that the neat liquid complex of Ph2Bu2 with HNO3 can be used for exhaustive extraction of americium from high-level saline wastes from plants for processing nuclear fuel.

Ceramics based on zirconolite and pyrochlore were prepared by cold pressing-sintering and fusion under the oxidative and weakly reductive conditions. Upon fusion in a weakly reductive medium the content of the perovskite-type phase in the samples increases, especially in plutonium-containing samples, probably due to partial reduction of Ti(IV) to Ti(III) and Pu(IV) to Pu(III). Ceramics based on pyrochlore contain impurity phases of brannerite and cubic solid solution based on uraninite.