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Том 122, № 5 (2017)

Articles

Computational and Experimental Analysis of Pu, Np, Am, Cm Reaction Rates on BFS Critical Benches

Andrianova O., Dulin V., Mikhailova I., Andrianov A.

Аннотация

A computational and experimental analysis is presented of a systematized and reconsidered series of experimental studies of the absolute fission rate of Pu, Np, Am, and Cm performed on the critical benches BFS-1, -2 in the period from 1990 to 2013. Twenty seven critical configurations are examined – models of the reactor core with different types of fuel and coolant. Previously obtained experimental values are re-examined according to more accurate methods of processing the results taking account of residual deformations of chambers and with the introduction of corrections for the efficiency of fragment detection (acts of fission in chambers). The computational models of the assemblies are supplemented by models for evaluating the ratio of the fission rates using non-analogical modeling algorithms. The matched set of experimental data and computational models can be used in diverse applied and fundamental problems.

Atomic Energy. 2017;122(5):291-298
pages 291-298 views

Article

Molten Salt Reactor for 99Mo Production

Feinberg O., Tataurov A.

Аннотация

99Mo production in a molten salt reactor with fuel salt based on metal fl uorides (LiF–BeF2–UF4) is examined. The proposed method of production is based on a particularity of the behavior of gaseous fission products in salt melts. Molybdenum, together with some other metals, does not form stable compounds in fluoride melt. At least 50% of fragment molybdenum in a molten salt nuclear-physical facility will be in a gas-aerosol phase above the surface of the molten salt. Neutron-physical and thermohydraulic calculations were performed taking into account the specific features of a molten salt reactor. The optimal configuration of the reactor, the materials used, the fuel salt composition, and the fuel cycle are described.

Atomic Energy. 2017;122(5):299-303
pages 299-303 views

Model Studies of Coolant Flow Hydrodynamics in VVER-1000 In-Reactor Pressure Channel

Krapivtsev V.

Аннотация

The results of the development, construction, and testing of a small-scale model of a pressure channel of coolant motion inside the VVER-1000 vessel from the entry connections to the coolant flow entry into the core are presented. A hydrodynamic bench, a model of in-reactor pressure channel in the VVER-1000, and methods for studying the coolant flow hydrodynamics in the channel model are described. The results of experimental studies and a comparative analysis of the coolant flow velocity distribution along the annular pressure channel from the entry into the channel through the pressure connections to the entry section in the elliptical bottom of the cavity of the model are presented. The experimental results obtained for different flow regimes of the coolant through the pressure connections are presented.

Atomic Energy. 2017;122(5):304-310
pages 304-310 views

HYDRA-IBRAE/LM/V1 Thermohydraulic Code Verification Based on BN-600 Experiments

Klimonov I., Usov E., Dugarov G., Butov A., Kudashov I., Ivanov E., Mosunova N., Strizhov V., Anfimov A., Gorbunov V., Kuznetsov D., Osipov S., Bel’tyukov A.

Аннотация

The results of thermohydraulic code HYDRA-IBRAE/LM/V1 verification on experimental data obtained using the BN-600 reactor in different years are described. The particulars of the computational scheme of the reactor using HYDRA-IBRAE/LM/V1 for modeling the BN-600 operating regime are presented. In preparing the scheme, special attention was devoted to accurate modeling of the reactor cool-down regimes on natural circulation as being most important from the safety validation standpoint. The computational results obtained with the uncertainty of the initial data taken into account are presented for such a cool-down regime.

Atomic Energy. 2017;122(5):311-318
pages 311-318 views

Particulars of Uranium-Plutonium Nitride Fuel Swelling During Low-Temperature Irradiation in Fast Reactor to Burnup 5.5% h.a.

Belyaeva A., Gil’mutdinov I., Zhemkov I., Kryukov F., Kuz’min S., Nikitin O., Ivanov Y., Kostyukov Y.

Аннотация

Electron-microscopic and electron-probe x-ray spectral studies of nitride fuel microstructure in thin sections and fractures of pellets in correlation with the xenon distribution were conducted in order to obtain data on the particulars of porosity formation and mechanisms of swelling of nitride fuel during low-temperature irradiation. The results attest indications of restructuring of nitride fuel irradiated to burn up 5.5% h.a. at temperature below 880°C. The formation of some grain structure in the restructuring process, accompanied by emission of gaseous fission products from the fuel matrix into pores, affects the swelling rate of uranium-plutonium nitride fuel.

Atomic Energy. 2017;122(5):319-325
pages 319-325 views

Effect of Operating Parameters on RBMK-1000 Fuel-Element Cladding Oxidation

Sukhikh A., Sagalov S., Pavlov S.

Аннотация

The results of research on the corrosion state of the fuel elements of sixteen RBMK-1000 fuel assemblies operating at the Leningrad NPP to burnup 6.1–35.7 MW·days/kg are presented. The dependences of the oxide thickness on the outer surface of cladding on the operating time, fuel burnup, and average power release of FA are presented. It is determined that increasing the power output from 0.9 up to 2.1 MW approximately halves the oxidation rate on the sections between the spacing lattices. For close FA operating parameters, larger oxide thickness was recorded on sections of fuel elements beneath the steel spacing lattices than beneath the zirconium lattices. It is shown that in terms of the conditions for its manifestation and external indicators the additional oxidation arising beneath the steel lattices has the characteristic features of shadow corrosion.

Atomic Energy. 2017;122(5):326-332
pages 326-332 views

Neutron Flux Density Effect on Vessel Steel Embrittlement

Nikolaenko V., Krasikov E.

Аннотация

The effect of the neutron flux density on the embrittlement of the reactor vessel materials is analyzed in an investigation of the impact of neutron irradiation on their properties. The evaluation of the dependence of radiation embrittlement of the vessel steel on the neutron flux density is made to a fi rst approximation for conventional virtual material by averaging the radiation embrittlement coefficient, taking account of the chemical composition, AF = 800(P + 0.07Cu). The actual radiation embrittlement coefficient was determined experimentally on the basis of the normative dependence ΔTF = AF(F·10–18)1/3. It was found as a result that the dependence of the radiation embrittlement coefficient of virtual steel on the neutron flux density has the form AF = 17.8 + 9.5/φ, where φ is the neutron flux density in the units 1011 sec–1·cm–2. As the neutron irradiation intensity decreases, embrittlement increases; the highest sensitivity to the neutron flux density is observed in the range (0.1–1)·1011 sec–1·cm–2. This dependence makes it possible to evaluate the impact of the intensity of neutron irradiation of vessel steel on embrittlement.

Atomic Energy. 2017;122(5):333-338
pages 333-338 views

Dependence of the Determined Value of the Elastic Modulus of a Material on the Ratio of the Distance Between the Supports to the Height of the Cross-Section of the Bent Sample

Gol’tsev V., Osintsev A., Plotnikov A.

Аннотация

A method is proposed for determining the elastic modulus of a material from the results of bending tests performed on a prismatic sample using simultaneously the method of digital correlation of images for precise measurement of the displacements of points in the sample. This procedure was tested in three-point bending of steel and graphite samples. The effect of the ratio of the distance between the supports to the height of the cross-section of the sample on the reliability of the determined elastic modulus of the material is evaluated. The values obtained for the modulus of elasticity of steel and graphite agree with the characteristic values for the investigated materials when the ratio of the distance between the supports to the height of the cross-section of the sample is greater than 7.

Atomic Energy. 2017;122(5):339-345
pages 339-345 views

Solid-Phase Conversion of Depleted Uranium Tetrafluoride into Oxides Using Mechanoactivated Quartz with the Addition of Sodium Fluoride

Polenov G., Chizhevskaya S., Zhukov A., Magomedbekov E.

Аннотация

The role of sodium fluoride, added to quartz before the latter is mechanically activated, in the solid-phase conversion of depleted uranium tetrafluoride into uranium oxides without mixing of the components of the mixture is determined.

Atomic Energy. 2017;122(5):346-352
pages 346-352 views

Enrichment of Regenerated Uranium in a Gas Centrifuge Cascade with Simultaneous Dilution of 232,236U by Waste and Low-Enrichment Uranium

Smirnov A., Sulaberidze G., Dudnikov A., Nevinitsa V.

Аннотация

It is shown that regenerated uranium can be enriched in a gas centrifuge cascade with three feed streams (waste uranium, low-enrichment uranium, regenerated uranium) and simultaneous dilution of the isotopes 232,234,236U. Computational experiments were performed for different 235U content in the low-enrichment feed. The chosen combination of diluents makes it possible simultaneously to lower the separative work expenditure and reduce the consumption of native raw material with respect to not only previously used multi-stream cascades but also a standard cascade for enrichment of native uranium.

Atomic Energy. 2017;122(5):353-357
pages 353-357 views

Use of Radioactive Beams of 7Be Nuclei to Study the Wear Resistance of Materials

Ogloblin A., Unezhev V., Novikov V., Malamut T.

Аннотация

This work is devoted to the use of beams of radioactive nuclei for applied purposes – to study the wearresistance of materials used in different industries. The crux of the method is that the radioactive nuclei obtained in the nuclear reactions 1H(7Li, 7Be) and 1H(10B, 7Be) are implanted in the experimental sample. After implantation of the nuclei, the sample undergoes wear tests. The γ-radiation of the implanted nuclei is high-energy; the spectrum is comparatively simple and does not produce appreciable radiation damage in the test samples. The method has been successfully tested in determining the wear-resistance of machine parts and mechanisms.

Atomic Energy. 2017;122(5):358-363
pages 358-363 views

Scientific and Technical Communications

Space-Time Characteristics of Relativistic Electrons Moderated in Matter

Akat’ev V., Metelkin E., Savinov A.
Atomic Energy. 2017;122(5):364-368
pages 364-368 views

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