Vol 122, No 5 (2017)

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.

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.

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.

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.

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.

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.