Atomic Energy最新文献

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.

The implementation of an investment project for the construction of a nuclear power engineering plant with high-temperature gas-cooled reactors and chemical engineering facilities for the production of hydrogen-containing products according to hydrocarbon reforming method represents a major development task in the field of hydrogen technologies. The product line of a nuclear power engineering plant consists of hydrogen, ammonia, and carbamide. The final product will be determined by the consumer. For effective technical and design solutions, enterprises cooperate to conduct research and development work on fuel, graphite, high-temperature materials, manufacture elements and components of the main reactor plant equipment, and verify calculation codes, at the same time as developing a regulatory framework and concept for the management of spent nuclear fuel and radioactive waste. As well as reviewing the main technical solutions for the project of a four-unit nuclear power engineering plant with high-temperature gas-cooled reactors and chemical engineering facilities for large-scale hydrogen production, the present article provides an estimate of the hydrogen cost price at various prices for natural gas, taking into account the stages of production, liquefaction, transportation, and storage.

The article presents the results of a simulation of the TOSQAN (France) and CSE (USA) experiments on the capture of aerosol fission product simulators by droplets of the spray system using the MAVR-TA code. The calculation results are well-correlated with experimental data. The main mechanisms for capturing aerosol particles by sprayed droplets under conditions of beyond-design-basis accidents are described. Analytical calculations were carried out to study the effect of aerosol particle density on the efficiency of particle removal. According to the performed calculations, the density of aerosol particles has a significant impact on their removal from the atmosphere of the containment.

The current development of lead-cooled reactor plants places the main focus of R&D on the primary circuit of the power plant. However, the high melting point of lead significantly affects the design of the secondary circuit, in particular, necessitating the creation of a special system for heating feedwater at the inlet to steam generators to a temperature that exceeds the melting point of the reactor coolant. The present article substantiates the need for the functional allocation of this system and formulates its main requirements.

The paper considers the development and application of digital twins for nuclear power facilities with the purpose of improving operational safety and cost efficiency. The TERMIT technology for automated simulation is presented to develop comprehensive calculation models of various classes and purposes for full-scale simulation of dynamic processes occurring in nuclear power facilities. The main directions and results of the technology application are provided.

The article describes the basic technical solutions and the composition of units for laser dismantling of spent fuel assemblies and cutting of fuel rods for reactors with a liquid-metal coolant during the processing of spent nuclear fuel. The use of lasers as part of robotic and automated complexes for fuel assembly dismantling and fuel rod fragmentation will reduce the loss of valuable fuel materials, prevent their entrainment, reduce the cost of processing, minimize the contamination of solid radioactive waste with fission products, and optimize primary operations in general.

The study of radionuclide migration is of particular concern in when substantiating safety procedures involved in the deep disposal of radioactive waste in the host rocks of the Yeniseisky site located in Krasnoyarsk Krai. One of the main processes controlling the migration of radionuclides is the sorption by mineral phases of host rocks. The present work aims to obtain quantitative characteristics of ¹³⁷Cs, ²³⁷Np, and ²³⁹Pu sorption by the rocks of the Yeniseisky site under conditions simulating a deep disposal of radioactive waste. The distribution coefficients of ¹³⁷Cs, ²³⁷Np, and ²³⁹Pu were determined depending on the ionic strength and temperature of the solutions after leaching the Mg-K-phosphate compound. Differences in the effect of K⁺ and Mg²⁺ cations on cesium sorption by the rocks were established. At an increase in the temperature, the efficiency of ²³⁹Pu and ²³⁷Np sorption by the rocks increases, while the sorption of ¹³⁷Cs remains almost constant.