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Using multidimensional particle-in-cell simulations, we show that an electron beam with a huge space charge can be accelerated to high energies by irradiating the edge of a solid density target with an intense femtosecond laser pulse at parallel incidence. The process of energy gain of each electron is divided into two parts: the transverse laser field and the longitudinal field of the excited surface plasma wave (SPW). It is shown that the longitudinal field of the SPW dominates the acceleration of the major part of electrons. This process leads to generation of a highly collimated electron beam with a huge space charge.

We present the results of the aberration correction of laser radiation wavefront using a dynamic method for determining the reference wavefront. The method, which is based on the processing of synchronously obtained data on the near- and far-field zones, significantly improves the focusing quality with active wavefront correction, especially under conditions of dynamic aberrations. An increase in the Strehl number S from 0.7 to 0.86 is demonstrated when a beam 18 cm in diameter is focused by an F/2.5 parabolic mirror.

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 results of the operation of a wavefront correction system based on a deformable bimorph mirror of the PEARL subpetawatt laser facility are presented. An improvement in the quality of focusing of laser radiation, which led to an increase in the Strehl ratio from 0.3 to 0.6, is demonstrated. The features of the compensation for phase distortions of the wavefront in the case of a low pulse repetition rate, as well as the correct allowance for the noise of the CCD camera when calculating the Strehl ratio are investigated.

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.

We present the results of studying experimentally the expansion of laser plasma in a strong external magnetic field (with a magnetic flux density of 13.5 T) at various sizes of the region of plasma formation on the surface of a solid-state target. It is shown that when the size of the plasma formation region is smaller than the classical plasma braking radius, a nearly identical topology of plasma flows is observed, which is characterized by the formation of a thin plasma sheet directed along the external magnetic field. If the width of the plasma formation region is comparable with the classical plasma braking radius, an additional plasma sheet starts to be formed.

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.

Adaptive system for wavefront correction based on 240-mm bimorph deformable mirror and Shack-Hartmann wavefront sensor is presented. The dynamic characteristics of the deformable mirror and the performance of the wavefront correction in various operating modes of the PEARL facility as well as the features of phase distortion compensation in a single-shot generation regime are studied. An improvement in the quality of focusing that led to an increase in the Strehl ratio to 0.6 is demonstrated.

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.

The object of the study was the gas distribution networks of various pressure categories that supply gas to municipal and industrial consumers from gas supply points, for example, gas distribution stations (GDS), during operation process the gas networks are exposed to a large range of loads and impacts, which, in combination with a long period of operation, leads to a decrease in the level of reliability. Studies have shown that the reduction of the emergency hazard for the used pressure categories in the gas distribution network in comparison with the higher category is 1.2 – 2 times, while at the same time there is decrease in the reduced costs to 40 percent or more.

The object of the research is an isothermal tank container for storage and transportation of liquefied natural gases, which requires special operating conditions and is related to a technological facility of increased danger. The purpose of the study is to substantiate the type and thickness of the insulating material to reduce the losses of liquefied natural gas during storage and transportation. Based on the results of the analysis, effective insulation materials were selected for use in cryogenic tanks for isothermal storage of LNG, the optimal thickness of the insulation material was determined, which provides the required level of losses for gas evaporation.

The design of ellipsoidal plasma mirrors (EPMs) for the PEARL laser facility is presented. The EPMs achieve a magnification of 0.32 in focal spot size, and the corresponding increase in focused intensity is expected to be about 8. Designing and implementing such focusing optics for short-pulse (

Today, natural gas, with certain advantages over other fossil fuels, is one of the main energy resources for various industries and housing and communal services. As a rule, when choosing a gas transportation method for different categories of consumers, several options are considered depending on the mutual removal of the source and the consumer – construction of a gas pipeline or the creation of a liquefied gas infrastructure. Analysis of energy supply problems in some constituent entities of the Russian Federation showed that a reasonable choice of technological parameters of a gas distribution system based on liquefied natural gas (LNG) is an urgent task that requires preliminary modeling. The paper proposes a solution to a particular problem of supplying consumers with liquefied natural gas, a logistic model for the optimal functioning of the system for the complex: LNG plant – a consumer. Based on the results of calculations for the conditions of the Republic of Karelia, recommendations are proposed for the optimal location of the source (LNG plant) and optimization of the transport component.

The article presents the prospects of using biogas fuel. Mathematical model of processes proceeding at the gas purification in amine installation unit has been developed. A software package has been proposed that allows to determine the quantitative yield of methane, design and thermal insulation characteristics of a biogas reactor by anthropogenic load. We determine the main parameters and energy efficiency of the biogas plant for the conditions of the Saratov, efficiency of energy-saving measures at the biogas fuel usage is shown. The practical significance of the work lies in the use of the developed model in the design of anaerobic reactors on polygon of residential solid waste (RSW), small settlements and microdistricts distant from other sources of energy supply.

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.

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.

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.

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.

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).

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.

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.

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).

Leach resistance of neptunium-doped garnet- and murataite-based ceramics with calculated compositions (in wt.%: 38 Fe2O3, 21 Gd2O3, 16 NpO2, 15 ZrO2, 10 CaO and 55 TiO2, 10 Mn2O3, 10 CaO, 10 NpO2, 5 ZrO2, 5 Al2O3, 5 Fe2O3 respectively was studied. The samples were prepared from pre-activated oxide mixtures by sintering/melting in platinum ampules at 1450 °C. The murataite ceramic is composed of predominant murataite structure phases with five- (5C) and eight-fold (8C) elementary fluorite unit cell and minor rutile (≤5% of total bulk). The garnet-based sample is a polyphase ceramic and consists of major garnet phase, interstitial Fe2O3 and Ca-Fe-phase, and rare grains of (Ca,Zr,Np)O2-phase (cubic solid solution) included in the larger garnet grains. The edge of the garnet grains is enriched with Np as compared with their rim by approximately 1.5−2 times. Np leach-rates were measured using an ANS_16.1 test. Np leach rate from the murataite-based ceramic reduced significantly from 2.44×10-6 g/(cm2 day) within the first 2 hours of leaching to 3.41×10-9 g/(cm2 day) at 672 hours of leaching. At the same time for the garnet-based ceramic corresponding values were 1.49×10-5 g/(cm2 day) and 1.77×10-6 g/(cm2 day), respectively. Higher leach rates at early periods are due to the dissolution of trace phases and Np release from defect surfaces. Leach rate values at later periods of leaching are characteristic of the “equilibrium” leach rate of the element (Np). Much higher Np leaching from the garnet ceramic may be due to 1) higher concentration of Np at the edge of the garnet grains with a relatively low chemical durability; 2) leaching of Np traces from highly soluble minor Np-Zr-rich phase.

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).

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.

Heavy cavalry formation, introduced by the Sarmatian tribes during the first centuries C.E., can be rightfully considered their most ingenious invention, which subsequently had a great influence on the formation of medieval knighthood (Cardini 1981; Russian translation Cardini 1987). This cavalry was later named "cataphracts" (from the Greek kataphraktos, mailed). The term is borrowed from classical tradition and, in general, it reflects the fact that that the issues of Sarmatians' defensive armament are studied more extensively than others in modern literature. In the present article, less-studied elements of Sarmatian military techniques and tactics are explored: the way of using the main weapon of the cataphracts—a lance [contus] and their horse-riding posture in a battle. We have used the two most informative groups of sources—written sources and drawings. The chronological framework for the article is the first and second centuries.

Methods of synthesis of (het)aroylpyruvic acids and their acyclic derivatives (esters, amides and hydrazides), and their reactions with various C-, N-, O-, S-nucleophiles are described. Problems of tautomerism and biological activity of (het)aroylpyruvates and products of their transformation are briefly considered. The bibliography includes 191 references.

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.

Pentafluorobenzoylpyruvic ester reacts with dinucleophiles — hydrazine, phenylhydrazine, hydroxylamine hydrochloride, 1,2-ethylenediamine, 1,2-phenylenediamine, 2-aminophenol — to form heterocyclic compounds. Reactions between bis(ethyl pentafluorobenzoylpyruvate)copper (II) chelate and these amine hydrochlorides also produce heterocycles. The analytical data of these compounds — elemental analysis, IR-, 1 H and 19 F NMR spectral data — are reported.

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).

Solid-phase extraction of plutonium in different individual and mixed oxidation states from simulated groundwater (pH 8.5) was studied. The extraction of plutonium species was carried out in a dynamic mode using DIAPAK C16 cartridges modified by N-benzoylphenylhydroxylamine (BPHA). It was shown that the extent of recovery depends on the oxidation state of plutonium. The extraction of Pu(IV) was at the level of 98–99% regardless of the volume and flow-rate of the sample solution. Pu(V) was extracted by 90–95% and 75–80% from 10- and 100-mL aliquots of the samples, respectively, whereas the extraction of Pu(VI) did not exceed 45–50%. An equimolar mixture of Pu(IV), Pu(V), and Pu(VI) was extracted by 74%. The distribution coefficients (Kd) and kinetic exchange capacities (S) of plutonium in various oxidation states were measured. It was found that during the sorption process, Pu(V) was reduced to Pu(IV) by 80–90% after an hour-long contact with the solid phase. Pu(VI) is reduced to Pu(V) by 34% and to Pu(IV) by 55%. In the case of mixed-valent solution of plutonium, only Pu(V) and Pu(IV) were found in the effluents.

Pentafluorobenzoylpyruvic ester reacts with dinucleophiles — hydrazine, phenylhydrazine, hydroxylamine hydrochloride, 1,2-ethylenediamine, 1,2-phenylenediamine, 2-aminophenol — to form heterocyclic compounds. Reactions between bis(ethyl pentafluorobenzoylpyruvate)copper (II) chelate and these amine hydrochlorides also produce heterocycles. The analytical data of these compounds — elemental analysis, IR-, 1 H and 19 F NMR spectral data — are reported.

Methods of synthesis of (het)aroylpyruvic acids and their acyclic derivatives (esters, amides and hydrazides), and their reactions with various C-, N-, O-, S-nucleophiles are described. Problems of tautomerism and biological activity of (het)aroylpyruvates and products of their transformation are briefly considered. The bibliography includes 191 references.

Actinide wastes involve actinide or rare earth–actinide fractions of high level waste (HLW), Pu-contaminated materials, including incinerator ashes, excess weapons plutonium, and some wastes formed during plutonium conversion in MOX fuel and nuclear accidents. SIA Radon in cooperation with Vernadsky Institute of Geochemistry, Institute of Geology of Ore Deposits, and D. Mendeleev University of Chemical Technology deals with development and testing of actinide waste forms and preparation methods. Zirconolite, pyrochlore, and murataite are considered as host phases for plutonium and other actinides. Two-phase ceramics based on zirconolite-perovskite, pyrochlore-perovskite, perovskite–cubic zirconia-based solid solution, murataite-perovskite, and zirconolite-murataite assemblages were designed for incorporation of actinide and rare earth–actinide fractions of HLW. Glass-ceramics containing apatite-britholite phases have been proposed for incinerator ash fixation. All these matrices have high chemical durability and radiation stability. The most promising method for production of these waste forms is an inductive melting in a cold crucible. Cold pressing and sintering technology is considered as alternative route. Mechanical activation intensifies ceramization process and reduces sintering temperature. Some new methods such as selfsustaining synthesis and plasma melting are being also examined.Copyright © 2003 by ASME