Atomic Energy最新文献

A simple numerical example suffices to demonstrate the scale of the change in the potential biological hazard when removing actinides from the spent nuclear fuel of a VVER-1000 reactor and the effect of uranium and transuranium element transmutation. The calculations assumed the complete removal of these actinides from the spent nuclear fuel, which can be considered as an ideal estimate. Without transmutation, the radiation equivalence of spent VVER nuclear fuel with consumed uranium raw materials is achieved only after 100 thousand years of exposure. The calculation shows how the time of achieving the radiation equivalence changes when removing certain components of spent nuclear fuel. In particular, if elements from uranium to curium, as well as caesium and strontium, are excluded from landfilled wastes—and, at the same time, thorium and radium accompanying natural uranium—are transmuted, then radiation equivalence is achieved at the complete combustion of the entire volume of mined uranium raw materials over approximately 150 years.

The paper describes an experimental study of lean hydrogen-air flames in a 1490-mm-long closed vertical channel of a square 138×138 mm section behind a horizontal diaphragm with a diameter of 50 mm. Schlieren images of the flame in a homogeneous hydrogen-air mixture are given along with the pressure dynamics for the mixtures with 6, 8, and 16 vol.% of hydrogen. It is shown that shear instability behind the diaphragm leads to a local instability of the lateral surface of a lean hydrogen-air flame, which causes a short-wave perturbation that spreads over the flame surface and crumples it, thereby enhancing the combustion.

The paper considers the possibility of replacing container fuel elements of thermal reactors with uranium dioxide or mixed oxide uranium-plutonium fuel with dispersion fuel elements of high uranium density, based on a metal fuel with a matrix of zirconium alloys, demonstrated high radiation resistance in reactor tests. Such fuel can ensure an increase in the duration of a fuel element campaign and/or a decrease in enrichment, which will save natural uranium and increase the breeding ratio. In addition, the content of plutonium in the fuel element will increase by more than 2 times, which can greatly simplify the closure of the fuel cycle. The use of dispersion composite fuel for thermal and fast reactors can replace the currently existing approach, based on the use of uranium dioxide or mixed oxide uranium-plutonium fuel pellets.

The article presents the results of investigations into the development of new technologies for the synthesis of polycrystalline thermoluminophores and thermoluminescent dosimeters (TLDs) MgB₄O₇:Dy (TLD-580K), MgB₄O₇:Dy,Na/Li (TLD-580N/L), MgB₄O₇:Gd,Na (TLD-GN), ⁶LiF:Mg,Ti (TLD-600), LiF:Mg,Ti (TLD-100). Thermal luminescence diagrams of TLD samples were obtained and compared with monocrystalline materials. The γ‑sensitivity of materials for neutron detection is considered.

The development of hydrogen energy technologies can be held back by the exceptional explosiveness of hydrogen-containing mixtures. In order to minimize the risks of accidents, it is necessary to adapt regulatory documents and create new safety analysis tools based on both experimental and calculated data. As part of ensuring the safety of hydrogen energy technologies, the Russian Federal Nuclear Center—Zababakhin All-Russia Research Institute of Technical Physics (RFNC—VNIITF) has created a complex of laboratory benches and polygon facilities, which are designed for detailed studies of hydrogen outflow, ignition, and combustion in various modes. High-precision numerical models are verified for calculating selected emergency scenarios. A software tool is also being developed for assessing the explosion safety of various emergency scenarios and providing a quantitative measure of the risks involved in specific technological solutions.

The paper analyses accidents and emergencies that occurred at reactor plants having a lead-bismuth eutectic (LBE) located in nuclear submarines and prototype ground-based facilities. Accidents and emergencies are systematized according to the initiating events and causes, as well as their consequences, and the measures developed for their future prevention. In the course of mastering the reactor technology, important scientific and technical problems, including the LBE technology itself, the corrosion resistance of structural materials, radiation safety under Po-210 emissions, the reliability of steam generators and associated equipment under coolant freezing-melting conditions, are studied with a view to providing their solutions.

The results of measuring BOR-60 parameters during the 108A fuel campaign in 2019 and the slow scram mode in 2020 were used to validate the EUCLID/V1 integrated code. A satisfactory agreement between the calculated and experimental results was obtained. The determined uncertainty interval of the mathematical model for the sodium temperature, sodium mass flow, and integral power is [−21, 24] °C, [−15, 18] %, and [−5, 6] %, respectively.

The relevance of research into the thermophysical properties of lead is associated with the implementation of a BREST-300 reactor project. The enthalpy of solid and liquid lead in the range of 432–1327 K was measured by the mixing method using a massive isothermal calorimeter. The enthalpy of melting was determined. Approximation equations for the temperature dependence of enthalpy in the range of 298.15–1325 K were obtained to determine the isobaric heat capacity. The existence of a heat capacity minimum for liquid lead in the region of 830 K was established. An analysis of existing recommended and experimental data on the caloric properties of lead was performed. The need for new measurements of the enthalpy and heat capacity of a melt at temperatures above 1050 K is substantiated.

Within the modernization of programs for calculating nuclide kinetics of BN-600 and BN-800 Beloyarskaya NPP reactors, programs for calculating nuclide kinetics of the fuel and fission products, nuclides in absorbing rods, as well as impurities in structural materials and the coolant, were developed using various methods including the possibility of an uncertainty analysis. The concentration of 116 nuclides is monitored for absorbing rods. For structural materials, including impurities, as well as for the coolant, 503 and 210 nuclides are monitored, respectively. All nuclide concentrations are determined taking into account the restrictions on impurities established by industrial standards. In order to account the coolant movement in the reactor, an original calculation algorithm was developed whose details are provided in this article.

The use of extended libraries of one-group cross-sections in precision nuclide composition calculations confirms the importance of including high-threshold reactions in the libraries. Particular attention is paid to impurity nuclides, which are increasingly relevant in solving problems of nuclear engineering. The existing ambiguity in the formation of extended libraries with high-threshold reactions leads to markedly different concentrations of individual nuclides, including impurities. The effect of extended libraries on the calculated nuclide composition is presented for steel irradiated with a neutron flux of 2.7 ∙ 10¹⁵ s⁻¹ cm⁻² for an irradiation time ranging from 10 to 1000 days.