Atomic Energy最新文献

Experiments were carried out to determine the liquidus temperature of 46.5LiF–11.5NaF–42KF-based melts simulating the fuel salt of a molten-salt burner operating in a Np, Am, Cm transmutation mode without the consumption of plutonium. The liquidus temperature was determined by differential scanning calorimetry; the melt composition was monitored by atomic emission, alpha, and gamma spectrometry. Melts containing PuF₃ and UF₄, as well as NdF₃ and NiF₂ simulating the products of fission and corrosion of structural materials were studied. For the actinide fluoride mole fraction of 5–15 mol% in the LiF–NaF–KF-based fuel salt in the presence of fission and corrosion products, the liquidus temperature of the melts was equal to 485–580 °C.

Contemporary approaches to the safety justification of nuclear power plants are characterized by constantly growing requirements for the volume and complexity of a computational analysis. Machine learning provides the ability to analyze large volumes of calculations. The present paper examines the methodology of a multivariate computational safety analysis based on machine learning. The approach proposed in the work will significantly reduce the time spent on labor-intensive calculations of safety justification. The selected learning model is presented with particular attention paid for its properties optimal for the solution of the considered problem. The ways of solving the problems arising during the preparation of a learning sample and model learning are described; the results of the model application for one of the typical problems including the safety justification of VVER reactors during a postulated design-basis accident using the conservative approach are provided. In conclusion, issues requiring further research, as well as the prospects for the development of the proposed methodology are indicated.

The article considers the advancement of a methodology developed by JSC OKB “Gidropress” for the calculation safety justification of VVER reactors in manoeuvring modes. The main challenge of the methodology in terms of the accident analysis is the selection and justification of initial conditions, which are carried out through expert assessment. To solve the problem, it is proposed to use machine learning for automating expert assessments based on available calculation results. The article proposes methods for constructing elements of a neural network and an algorithm for its learning. The results of the work of these elements and their combinations for the solution to the given problem are analyzed. Conclusions are made about the possibility of advancing the methodology through the development and implementation of a multilayer neural network that takes into account the accident type, manoeuvring algorithm, moment of the campaign, specifics of a particular project, and other factors important for the safety justification.

The present study aimed to select a method for extracting and conditioning sorbents located in the basements of the Fukushima Daiichi NPP (Japan) after the accident elimination. FSUE “RADON” conducted tests of two schemes for extracting and hydraulically unloading sorbents using a pump: directly into a filter trap or a dispenser with pulp overflow into the basement, followed by the transportation of thickened sorbents from the dispenser to the filter trap. The filter trap filling time, which is controlled by the passage of sorbent particles through the inspection window of the flange at the outlet of the flow, does not exceed 1.5 h at a filling degree of more than 95%. The recommended post-treatment mode includes rinsing with water to a chloride ion concentration of less than 30 mg/dm³, dehydration with compressed air, and holding for draining the remaining liquid with periodic pumping.

The paper considers the reduction of rare earth metal (REM) oxides and uranium dioxide with lithium produced during the electrolysis of LiCl–Li₂O melt with the formation of intermetallics and palladium. At a cathode potential of 0.6–0.8 V relative to (E_{{mathrm{Li}^{+}}/{mathrm{Li}^{0}}}), intermetallic compounds of CePd₃, NdPd₃, and UPd₄ compositions are formed. The formation current for REM intermetallic compounds is significantly greater than that for uranium. Therefore, when they are co-present in samples, REM intermetallics are formed first, followed by intermetallic compounds of uranium in the presence of palladium unbound by REM alloys. This is due to the significantly greater solubility of neodymium and cerium oxides in the salt melt compared to uranium dioxide. At a cathode potential close to or equal to the potential of liquid lithium, intermetallics with palladium, lanthanides, and uranium Ln₃Pd₄, LnPd, UPd₃ are formed. In this case, an important role is played by the ability of lithium and palladium to form alloys that are liquid at 650 °C.

An analysis of the consequences for an accident caused by the blockage of the core flow area is of particular importance for the safety of metal-cooled reactors. A blockage may be caused by the release of slag accumulated in the primary circuit due to the violation of the coolant technology. In order to identify the self-protection margins of BREST-OD-300 and BR-1200 reactors, a comparative analysis of the consequences was carried out for an accident caused by the partial blockage of the core flow area. The velocity and temperature of the coolant in the core, as well as the temperature of the fuel and fuel element claddings during the blockage of the fuel assembly inlet was calculated using the SVIR 3D thermal-hydraulic code. Both reactors were demonstrated having high self-protection against the accident caused by the blockage of the most flow area at the core inlet. Thus, BR-1200 and BREST-OD-300 fuel elements maintain gas integrity until the inlet flow areas of 12 and 40 fuel assemblies is blocked, respectively.

The paper considers a highly efficient technology of combined electrochemical and ultrasonic decontamination of metals to reduce the volume of waste disposed during the decommissioning of nuclear and/or radiation hazardous facilities. The possibility of a further increase in its efficiency by the use of dilute sulfuric acid as an electrolyte, preliminary exposure of the metal in the electrolyte, and variable polarization, as well as by the reduction in the distance from the treated surface to the emitter is experimentally demonstrated. Laboratory tests with real radioactively contaminated samples of steel from the Leningrad NPP, as well as pilot industrial tests on batches (~100 kg) of metal waste at the Angarsk Electrolysis Chemical Combine JSC were conducted. The obtained results prove high efficiency and productivity of combined decontamination.

The article reviews the development of neutron generators with a neutron target located at the end of a thin needle ion-guide applicator. This design allows a source of fast neutrons to be inserted directly into the volume of the irradiated object. Generators of this type are used for research of small-sized nuclear reactors. Another area of their use is intracavitary brachytherapy. In this case, the irradiation target is delivered directly to the patient’s tumor through the body’s natural cavities. The article reviews a TGI94 neutron generator for the inspection of small-sized space reactors and an NG-20 generator for radiation therapy of cancer patients.

The paper presents the results of the verification of the TARUSA‑9 software, which uses an exact analytical solution to the system of linear differential equations of mass transfer, for measuring the activity of fission products in the water coolant of a pressurized water reactor. A comparison of the calculation and measurement results was performed for three typical reactor operating modes: operation at a variable power with the disabled reactor water purification system, at a constant power with variable purification water consumption, and at a variable power and purification water consumption. The activity of 14 fission product isotopes in 39 reactor water samples was compared. The discrepancy between the results of calculations and measurements does not exceed the measurement error in more than 90% of cases. For the entire set of comparison points, the average relative difference between calculations and measurements is equal to 0 at a confidence probability of 0.95, standard deviation of 1%, and confidence interval of ± 2%. For individual nuclides, these values vary from −10 to 7%, from 0 to 12%, and from −23 to 24%, respectively.

The article presents the results of an experimental study connected with the features of a coolant flow at the inlet to the fuel assembly (FA) in the cartridge core of a RITM nuclear icebreaker reactor. The study is focused at the non-uniformity of the axial flow velocity at the FA inlet. To this end, a series of experiments was carried out using a scale model including structural elements of the inlet section between the orifice plate and the unit attaching fuel rods to the diffuser, as well as the section of the fuel rod bundle up to the second spacer grid. In addition, the propagation of a non-uniform axial flow along the length of the fuel rod bundle was assessed during the experiments. The studies were carried out using the pneumometric method in characteristic cross-sections along the length of the model. Features of the coolant flow are visualized by cartograms of the axial velocity for the working medium flow. The results of experiments were used for hydraulic profiling of the FA inlet section. The obtained experimental database can be used to validate the LOGOS CFD software and clarify the methods for thermal-hydraulic calculations of cores in a cell approximation.