Atomic Energy最新文献

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.

An increase in power during transient operating modes of reactors can lead to the depressurization of fuel rods due to the stress corrosion cracking mechanism. Expansion of fuel pellets due to increasing power induces tensile stresses in the cladding, which can initiate the formation and growth of corrosion cracks in the presence of aggressive fission products. Therefore, the plotted limit curve is primarily aimed at preventing circumferential stresses in the cladding from exceeding a certain maximum value during abrupt changes in power. The present article describes a system for performing multivariate statistical calculations of a single power increase to plot an envelope curve for the permissible area of fuel rod operation. The system uses the START-3A code to perform fuel rod calculations, which includes the ability to vary model and design parameters. The methodology for the calculated limit curve for a rapid increase in the power of a VVER-1200 reactor is presented with examples.

The article describes the history of creating the first Russian nuclear reactor capable of generating powerful controlled fission pulses. The FKBN‑I reactor was launched in 1964 at NII-1011 (currently FSUE “RFNC—VNIITF named after Academ. E.I. Zababakhin”). Forming the basis of FKBN‑I, pioneering work was carried out on the experimental development of technical solutions and control modes, which were used for the design of aperiodic pulsed reactors. The main neutron-physical characteristics and parameters of first fission pulses are given along with a description of methods for obtaining them. The article also mentions the main developers and researchers involved in the creation of the reactor.

A comparative study of technologies for conditioning reprocessed spent ion-exchange resins was carried out by FSUE “RADON” on laboratory and pilot-industrial scales. According to industrial application criteria, a choice was made in favor of dehydration and impregnation using a polymer binder based on epoxy resins. To implement this technology, a pilot unit was developed and manufactured. Tests of the unit carried out with actual resins demonstrated its operability and the compliance of the resulting polymer compound with regulatory requirements.