Atomic Energy最新文献

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.

The development of small nuclear power plants (SNPPs) is one of the priority areas of the Rosatom State Corporation. The relevance of SNPP development is determined by the economic feasibility and prospects for their use in hard-to-reach areas. SNPPs are expected to be used as a promising energy source in regions with decentralized power supply. The main core of the reactor No. 1 at the “Akademik Lomonosov” floating thermal power plant has been successfully decommissioned. Core fuel elements developed by the JSC “VNIINM” have passed pre-reactor and reactor tests, as well as post-reactor studies, which, together with preliminary operation results, confirmed their operability and potential for use.

In order to use reduced enrichment fuel and increase the burnup of fuel elements in research reactors, metal-ceramic alloys based on uranium-molybdenum superalloys are being developed. The alloy structure is based on a γ-phase, which also contains ceramic (intermetallic) phases. Having the highest density of uranium, compatibility with aluminum, high radiation resistance, and low molybdenum content, these phases precipitate along grain boundaries, which allows fuel particles to be obtained by crushing. This fits into the existing technology. The use of such fuel will make it possible to manufacture and supply reduced enrichment fuel (fuel elements) with increased uranium density and performance to research reactors without increasing costs.

The paper presents experimental data on the formation of natural air circulation in an open circuit with the length of 22.4 m and a pipe diameter of 120 mm, which is located in a room connected to the atmosphere, in a case of the emergency cooldown of a lead-cooled reactor. Natural circulation without heating was ensured by the difference between room and atmospheric temperatures, while the air heated to the temperature of 170–470 °C by the electric heater with a power of 7–12 kW circulated due to the reduction in its density at the outlet. A model is proposed for calculating the pressure of natural circulation using generalized experimental data obtained in an open-air circuit under the conditions of a quasi-stationary flow, temperature, and static pressure. The generalization error does not exceed 30% with a total error of 20%. Hydraulic losses in the circuit and heat losses through the circuit insulation were determined. Atmospheric turbulence is demonstrated to cause pressure pulsations in the circuit with an amplitude close to the flow velocity head of 8.5–20 Pa in the duct. When the atmospheric air pressure at the inlet to the circuit exceeds that at the outlet, the amplitude increases to 280 Pa.

The article presents the results of a mathematical Monte-Carlo simulation of the passive mode of an active well coincidence counter (AWCC) in the Serpent software environment. The developed model was experimentally tested on different types of neutron sources. Estimates show that the results of the numerical simulation in the Serpent software environment can be used to refine and expand the range of effective mass measurements for ²⁴⁰Pu by adjusting the calibration coefficients.

The article represents an approach used at the IPPE JSC to the computational analysis of a severe accident caused by a complete cessation of the system and emergency power supply at a nuclear power plant with a sodium-cooled fast reactor. The analysis starts with the initial event and ends with the calculation of the radiation dose to the population. The used codes and methods, the development of the accident, and its consequences are described. The COREMELT code simulating thermal-hydraulic and neutron-physical processes in fast reactors was used to determine the scale of the core destruction. The accumulation of fission products in the fuel is calculated by the SKIF 1.0 code; their release to the gas volumes of fuel elements, to the coolant during the depressurization of fuel elements, and then to the gas cavity of the reactor is determined methodically (in the future, it is planned to use the Alpha‑M code). The transfer through the gas system is calculated by the COREMELT gas module. The activity of fission products released into the environment is methodically assessed (in the future, it is planned to use the KUPOL-BR code). The VYBROS-BN code was used to determine the radiation dose to the population. According to the performed calculation analysis, despite the significant destruction of the core, the dose load to the population at the border of the zone will not exceed the standards.

The present article reviews the experience of emergency response after major radiation accidents. It is noted that the experience of emergency response in the agricultural sector must be considered taking into account the effectiveness of protective measures, regulation in the field of radiation safety, methodological and resource support, as well as the perception of emergency response measures by the population and decision-makers at various levels. Lessons from the assessment of the consequences of major radiation accidents and emergency response are highlighted.

The article presents the results of determining the effective fraction of delayed neutrons β_eff for a BFS critical assembly whose fuel is composed of metallic plutonium and depleted uranium nitride and dioxide. The traditional method for determining β_eff by introducing the ²⁵²Cf source into the core is supplemented by the ²⁴⁰Pu source of spontaneous fissions in the fuel composition. The described modification simplifies the analysis, increases the reliability of results, and can be used in power reactors with plutonium of known composition in which it is impossible to use a ²⁵²Cf source.