Atomic Energy最新文献

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.

Abstract

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.

The paper presents the results of studies connected with the features of coolant hydrodynamics at the outlet of various design heads of a RITM fuel cassette for a nuclear icebreaker and an SNPP. Experiments were carried out using a research bench with an air working medium on a model of the fuel cassette outlet section by pneumometric and contrast injection methods based on the theory of hydrodynamic similarity. The hydrodynamic pattern of the coolant flow is visualized using cartograms of the axial velocity at the outlet of the fuel rod bundle, behind the FA heads of various designs, in front of the letdown pipe, as well as in the upper support plate. The determination of fuel rod bundle areas, where the coolant can flow into the letdown pipe, was carried out by the contrast injection method. The results of the studies establish the influence of the FA head design on the heterogeneity of the axial flow at the outlet of the fuel rod bundle and in front of the upper support plate. The more intensive mixing of the coolant behind the head of the fuel cassette of a nuclear icebreaker as compared to the head of the SNPP fuel cassette is due to the inhomogeneity of the flow leaving the passage holes of the heads.

The article examines a mechanism that significantly influences the operation of radioelectronic equipment when it is located in close proximity to an electron accelerator. In particular, the effect of accumulation of an opposite sign charge and the appearance of micro-breakdown effects are considered in terms of this mechanism. A calculation model explaining these phenomena is described. The model is based on measuring the path of positrons and electrons in substances, as well as the characteristic dependence of the transmission coefficient for electrons and positrons on the path length.

The paper presents a justification, setting, and results of experiments on the determination of the liquidus temperature for a molten salt comprised of 73LiF–27BeF₂ with plutonium trifluoride, considered as the fuel circuit salt of a liquid-salt reactor for burning Np, Am, Cm. The liquidus temperature of the melt was determined using a differential scanning calorimetry method; the plutonium concentration in the samples was controlled by recording the γ‑radiation of plutonium isotopes. According to the results of the studies, the liquidus temperature comprised 600–680 °C across a range of plutonium trifluoride concentrations from 0.1–3 mol.%. According to the performed calculations, in order to transmute 250 kg/year of Np, Am, Cm in an MSR‑T reactor, this restriction on the concentration of actinide fluorides in the fuel salt at its operating temperature in the reactor of 650–750 °C requires 550–660 kg/year of plutonium.

The RITM-200N reactor represents a modification of a RITM-200 icebreaker reactor, which is designed on the basis of proven ship reactor construction technologies. The reactor plant is being developed for a ground-based SNPP power unit to be constructed on the territory of the Republic of Sakha (Yakutia). The article reflects the main results of developing the technical design of the RITM-200N reactor and completing a part of R&D in support of technical solutions.

Abstract

A total number of 18 experimental fuel assemblies (FAs), containing mixed nitride fuel rods with claddings of various geometry, were irradiated in an BN-600 core. By the end of 2022, JSC SSC NIIAR had completed the post-reactor studies of 16 ETVS BN-600 nitride fuel rods, 4 fuel rods with a gas sublayer irradiated in the composition of OU‑1 and OU‑2 collapsible irradiation devices, and 3 OU‑4 BOR-60 fuel rods with a liquid metal sublayer. Post-reactor studies of nitride fuel rods were preceded by pre-test calculations of their stress-strain and temperature states, taking into account the actual irradiation parameters. The results of the studies were compared with the calculated data. Fuel radiation swelling represents one of the most important factors that determine the degree of thermomechanical interaction between the fuel and the cladding, thus limiting the performance of nitride fuel rods. The present article summarizes and analyzes the currently obtained data on the swelling and its rate for mixed uranium-plutonium nitride fuel in fuel rods with a gas sublayer.