Abstract: The calculation of the spent-fuel composition is the baseline problem of the analysis of the nuclear and radiological safety of objects with spent fuel assemblies. The use of fission product libraries formed on the basis of evaluated nuclear data with an enlarged number of energy points is especially important for high-precision calculations of the nuclide composition of fuel irradiated in a fast spectrum. The results of an analysis of the nuclide composition of mixed oxide and uranium-plutonium-zirconium metal fuel with burnup ~10% h.a. in a fast spectrum, which were obtained in high-precision calculations based on data from different fission product yield libraries, are presented. PubDate: 2015-04-14

Abstract: The dissolution and extractive reprocessing operations for AM-reactor spent fuel based on uranium carbide in a mixture with AM uranium-molybdenum fuel (9% molybdenum) and in combination with VVER-440 are studied and checked on model and real solutions. An approach combining catalytic oxidation of water-soluble organic compounds by nitric acid with their recomplexing by zirconium and/or molybdenum is proposed. The reprocessing of damaged AMB fuel and the possibility of reprocessing mixed uranium-plutonium fuel for fast reactors using autoclave oxidation instead of additional recomplexing are examined. PubDate: 2015-04-11

Abstract: A three-dimensional CFD code CONV-3D is being developed for mathematical modeling of the fluid dynamics of elements of nuclear power facilities. The code is based on developed algorithms with low artificial diffusion, for which discrete approximations are constructed using finite-volume and staggered grid methods. A regularized nonlinear monotonic operator splitting scheme has been developed to solve advection problems. Richardson’s iteration method with a fast Fourier transform for the Laplace operator is used to solve the pressure equation. Compared with the conventional conjugate gradient method this approach to the solution of elliptic equations with variable coefficients gives multiple acceleration. Quasi-direct numerical modeling is used to model three-dimensional turbulent flows in single-phase streams. The CONV-3D code is completely parallelized and is effective on multiprocessor cluster computers, such as Chebyshev and Lomonosov (MGU). PubDate: 2015-04-11

Abstract: The energy-production losses versus the reserved reactivity excess as a function of the probability of accidental shutdown of the reactor are presented. A formulation of the optimization problem of power control during the period of threat under force-majeure circumstances is examined. The efficacy of optimization and the character of the optimal regimes as a function of the duration of the period of threat are presented. PubDate: 2015-04-11

Abstract: The process of separating high-purity chemical concentrates of uranium, which are an intermediate product in the general scheme of obtaining materials of the nuclear industry, is studied and the parameters of the sorption-desorption concentration of uranium using different regenerating solutions are determined. Variants and combinations of the processes of selective separation of chemical concentrates in the form of ammonium and sodium polyuranates and uranium peroxide from commercial desorbates and re-extracts in application to the reprocessing of raw materials from the Elkonskoe and Streltsovskoe deposits are proposed. It is shown that chemical concentrates of uranium meeting the ASTM C 967-08 standard specifications can be obtained. The results presented can be used to improve the technological schemes of uranium extraction during the reprocessing of ores and concentrates with the final product meeting all applicable regulatory requirements. PubDate: 2015-04-11

Abstract: Carbide nuclear fuel is most attractive for use in fast reactors. A general problem of hydrometallurgical reprocessing of spent carbide fuel is the formation of plutonium complexes with water-soluble organic compounds of an acidic nature, which are formed when such fuel dissolves in nitric acid. The drawbacks of the proposed methods of decomposing organic compounds include the possibility of secondary precipitation and reduction of the extractability of multivalent actinides, long-time limitations in choosing the structural materials of the equipment, and incomplete decomposition of organic compounds. Often, the solutions obtained are unsuitable for direct extractive reprocessing because of the need to remove corrosive reagents or their inorganic residues, which increases the volume of secondary wastes. PubDate: 2015-04-11

Abstract: The release of nitrogen oxides (N2O, NO, NO2) accompanying the dissolution of UN in HNO3 with concentration 3–16 mol/liter and the ratio S:L = 1:7.5 and in 7.2 mol/liter HNO3 with S:L varying from 3 to 30 in atmospheric air or argon is studied. In the indicated HNO3 concentration range N2O is formed mainly as a result of the oxidation of UN and secondarily as a result of the reduction of HNO3. The presence or absence of oxygen in the atmosphere has no effect on the amount of N2O formed. The formation of NO2 is due to the reduction of HNO3. The presence of oxygen increases the amount of NO2 formed as a result of the oxidation of NO. The amount of nitrogen monoxide NO released into atmospheric air is higher than the amount released into an argon atmosphere. The amount of NO formed increases with increasing nitric acid concentration. The release of nitrogen oxides from UN as well as from nitric acid results in considerable dilution of nitride nitrogen, which casts doubt on the possibility of recycling 15N in the case where U15N is used as fuel. PubDate: 2015-04-11

Abstract: A scheme for producing a diluent for high-enrichment uranium in a double cascade is examined. The marketable product, viz., low-enrichment uranium, is obtained in the product flow of the first cascade; the diluent is obtained in the product flow of the second cascade from the waste flow of the first cascade. To reduce the cost of the separative work performed on the production of all products it is proposed that the feed point for the first cascade be displaced in the direction of the product flow relative to the optimum, calculated from the condition of optimization of the cascade for the production of 235U. A computational experiment shows this scheme to be efficient and the particularities associated with the use of natural uranium and depleted wastes as initial materials. PubDate: 2015-04-11

Abstract: Accurate data on the neutron yield from the interaction of α-particles with the nuclei of light elements ranging from lithium to potassium are required for solving the problems of nuclear power technologies: development of analytical means for controlling the technological processes of fabricating and reprocessing nuclear fuel, securing radiological protection for workers, improving the systems for managing and monitoring nuclear materials and radioactive wastes, measuring the burnup fraction of spent nuclear fuel, and others. The uncertainty of this information must be <10% for energies ranging from 4 to 9 MeV of α-particles emitted by naturally occurring and artificial radionuclides. The computational uncertainty of the neutron yield can be reduced on the basis of a combined analysis of (α, n) reactions, measured on α-particle accelerators with tunable energy and on compounds of actinides with light elements, using reliable data on the stopping power of α-particles for elements from hydrogen to californium. The results of such an analysis based on experimental and evaluated data for the light isotopes 6Li, 7Li, 9Be, 10B, 11B, 13C, 14N, 17O, 18O, 21Ne, 22Ne, 19F, 23Na, 25Mg, 26Mg, 27Al, 29Si, 30Si, 31P, 33S, 34S, 35Cl, 37Cl, and 41K in the α-particle energy range from 4 to 9 MeV are presented. PubDate: 2015-02-18

Abstract: A method is proposed for concentrating in different output flows in a square cascade with two additional extractions the components of the multicomponent isotopic mixture being separated. The parameters of the proposed cascade, a square cascade with additional extraction, and a cascade with flow expansion at internal steps on condition that the number of separative elements in them is the same are compared for the separation of a mixture of tungsten isotopes. It is shown that the proposed cascade expands the possibilities of previously known methods of concentrating intermediate-mass isotopes, since several target products can be obtained at the same time in the output flows. PubDate: 2015-02-15

Abstract: The reconstruction of fields and adjustment of the parameters of reactor models are increasingly prioritized for reactor physics because the efficiency and safety of the control of a nuclear plant depend on solving them. The results of reconstruction and parameter adjustments are largely determined by the probability distributions of the computed and measured data. The probability distribution of the computed data in turn depends on the properties of the neutron-physical model and probabilistic characteristics of the parameters of this model. The present work is devoted to the search for and investigations of efficient algorithms for calculating the covariation functions of the neutron flux density. A model of a stationary subcritical reactor with an internal neutron source is studied. Linear perturbation theory, the theorem on the spectral decomposition of operators, and the theory of stochastic processes are the tools used for performing the analysis. PubDate: 2015-02-15

Abstract: Research performed at the Bochvar All-Russia Research Institute for Inorganic Materials (VNIINM) and the All-Russia Research Institute of Experimental Physics (VNIIEF) on the development of the TUK-117 multipurpose transport packaging is presented. The TUK-117 packaging possesses biological protection based on depleted uranium and intended for transporting and storing spent nuclear fuel. The higher capacity of the metallic uranium alloy for absorbing of γ-rays and the alloy’s high mechanical characteristics have made it possible to develop a design meeting IAEA specifications. The technology for fabricating safety-engineered articles from uranium and its alloys by pouring the alloy into a prepared mold made of corrosion-resistant steel is patented. PubDate: 2015-02-15

Abstract: A reactimeter was developed for the IBR-2 periodic-pulse reactor. The reactor kinetics was described by difference equations, which relate the reactor’s parameters, corresponding to the instantaneous and preceding power pulses, and the nonlinear dependences of the energy of the power pulse and its amplitude on the reactivity. The fact that the regulated parameter of the reactor is the relative deviation of the amplitude of the power pulse from its base (prescribed average) value was taken into account. In studying transient processes, a statistically optimal filter was used to suppress the reactivity noise due to the reactor’s construction and principle of operation. The best location for inserting the filter in the block diagram of the reactimeter was chosen and the optimal smoothing coefficient determined. PubDate: 2015-02-15

Abstract: The electric and thermal conductivity of the aluminum alloys AMG-2 and SAV-1, which are used to fabricate the fuel elements of the fuel assemblies of a nuclear reactor, was investigated in the range 290–490 K. The Debye temperature was calculated and the contribution of the electronic and lattice components to the thermal conductivity of the alloys was determined. It is shown on the basis of an analysis of the mechanisms of electron and phonon scattering and the experimental data that one possible reason for the observed changes of the thermal conductivity in aluminum alloys are the scattering of electrons by phonons and the deviation of the Lorenz number from the theoretical values as a result of inelastic scattering of the electrons. PubDate: 2015-02-15

Abstract: The results of an investigation of the beam dynamics in a high-power proton accelerator-driver intended for operation as part of a subcritical electronuclear facility are presented. The initial normally conducting part of the accelerator at current 10 mA with current-passage ratio >95% has been developed. A channel with current-passage ratio 100% has been developed for the superconducting part of the accelerator. Modeling of the accelerating cavities for all velocity ranges has been done. High electrodynamic performance has been attained as a result of optimization. PubDate: 2015-02-15

Abstract: Heat-exchanger models in the SOKRAT-BN code which are used to calculate problems with boiling sodium in channels with different geometry are presented. The results of modeling of different experiments on boiling of liquid-metal coolant are presented. Good agreement is obtained between the SOKRAT-BN calculations and experiments performed with stationary and nonstationary boiling of sodium. It is shown that the thermohydraulic processes occurring in reactor facilities during design-basis and beyond design basis accidents can be calculated correctly using the thermohydraulic module of the SOKRAT-BN computer code. PubDate: 2015-02-15