Atomic Energy最新文献

The present paper provides the assessment of the thermal circuit for a lunar nuclear power plant using lunar soil to cool the cold junction of a thermoelectric generator and utilize unconverted heat. A mathematical description of the main processes and calculations were used to demonstrate energy generation and excess heat utilization schemes based on the thermal conductivity of the soil as of little promise given a wide range of soil thermal conductivity coefficients. Most probably, heat engines based on the difference in the temperature on the Moon surface and deep in the lunar soil are also not capable of functioning for any significant amount of time.

The paper considers the concentration of ^{72, 74}Ge in a square cascade with large stage separation coefficients corresponding to gas centrifuges. Cascades are calculated by varying the sections of partial stage streams with minimal deviation of calculated stage feeding streams from the specified one. The determined random search algorithms rapidly converge the objective function to the minimum. Thus, a single square cascade can effectively concentrate ^{72, 74}Ge isotopes.

Spent extractant in a diluent accumulates of hydrometallurgical fuel reprocessing. Extractant in a diluent classified as liquid radioactive waste, which must be converted into an environmentally safe state. The article presents the results of testing a magnesium potassium phosphate matrix for solidifying a spent tributyl phosphate—dodecane simulant. We examined solidified samples for compressive strength, water stability after 90 days of immersion, resistance to cycling freezing and thawing (30 cycles from −40 to 40 °C), and water resistance in accordance with GOST R 52126–2003. Thus, a magnesium potassium phosphate compound can include up to 20 wt% of aerosil pre-saturated with a tributyl phosphate—dodecane mixture. This corresponds to 17.6 wt% of the extractant in the diluent. Moreover, quality indicators of the studied compound meet the requirements for solidified radioactive waste.

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.