Atomic Energy最新文献

The development of the JSC “INM” (Zarechny, Russian Federation) as one of the leading centers with a research nuclear facility requires to determine the areas of activity after 2040. The article examines two promising options for the development of the JSC “INM” reactor experimental base. The implementation of each option will promote for the development of existing activity areas. In turn, the construction of a new reactor at the site of the Ural Nuclear Center will significantly expand the research potential of both the organization itself and the region as a whole. In addition, the article presents the main design solutions adopted as a basis for the projects on the significant modernization of the existing IVV-2M reactor and development of the new one.

Periodic pulsed reactors are unique research neutron sources used for experiments in solid state physics, nuclear physics, and radiation research. A reactor of this type represents a dynamic system with multiple feedback loops. Longstanding operation of the BR-2M reactor and its predecessors IBR‑2 and IBR-30 observe unstable pulse dynamics under certain conditions. Several physical feedback phenomena that can lead to unstable dynamics have been discovered and described. In this regard, a software package is being developed to simulate the operation of the reactor taking into account known physical processes, design parameters, and materials of the core in order to study the characteristics of an operating periodic pulsed reactor and design an advanced neutron source. The paper presents a description of the model and results of calculating reactor dynamics.

The paper presents a method developed for calculating a squared-off cascade with several types of stages along the feeding stream. The method minimizes the total feeding stream of the cascade. Calculations of germanium tetrafluoride separation in cascades with different numbers of sections have been performed. Using the example of ⁷⁰Ge concentration, the optimal three-section cascade is demonstrated as effectively approximating the conical profile of R-cascade streams.

Fuel rods containing a mixed uranium–plutonium nitride (MNUP) fuel were subjected to irradiation in BOR-60 and BN-600 reactors to assess their performance and potential use in BREST-OD-300 and BN-1200M reactors. Post-irradiation examinations revealed that their deformation behavior differs from that of fuel rods with oxide fuel. Nitride fuel rods demonstrated increased axial elongation and cladding ovalization under identical irradiation conditions within the same assembly. Elongation and ovalization studies were carried out for 12 fuel rods with cold-worked ChS68-ID cladding, 38 rods with cold-worked EK164-ID cladding, and 69 rods with EP823-Sh cladding. The fuel rods were irradiated in 17 experimental fuel assemblies, covering a range of maximum fuel burnup from 3.1–9.1% heavy atoms and displacement damage from 26–108 displacements per atom (dpa). The experimental data suggest that axial elongation is predominantly attributable to axial forces that emerge from thermomechanical interactions between the fuel and cladding. These interactions are attributed to the random displacement of pellets and/or their fragments from an axisymmetric position.

This study investigates evacuation schemes for the collector of a condensation-evaporation unit used in a separation plant for treating fluorine-containing gases at pressures of up to 1.3 kPa. The results demonstrate that the efficiency of chemical absorbers depends on process conditions. The scheme employing alkaline absorbers exhibits over fivefold greater efficiency compared to the standard system using activated aluminum oxide. The operability of the pump is shown to correlate with the fluoride ion content in the vacuum oil. When alkaline absorbers are used, the accumulation rate of fluoride ions in the oil is 0.61 g/day, which is nearly 50 times lower than with activated aluminum oxide. Furthermore, the modified mercerized wood used as a chemical absorber exhibits a technological advantage over conventional lime-based absorbers. Figures: 3; Tables: 5; References: 7.

The MATADOR software designed for subchannel thermohydraulic calculations of fuel assemblies with different coolants is being developed at OKB Gidropress JSC. Validation of the program with experimental data on key thermohydraulic parameters in the reactor core is of paramount importance, as it confirms the reliability of the modeling of physical processes and phenomena. The article presents the results of the validation and verification of the MATADOR subchannel program based on a series of experiments with supercritical water cooling of rod bundles conducted in China and at the NACIE-UP test facility in Italy, which has a 19-rod assembly cooled with lead-bismuth coolant. The modeling of rod bundles with wire wrapped spacers was performed using a geometric model. Analysis of the computational studies confirmed the ability of the software to reproduce measurements with acceptable accuracy for engineering practice.

A model of an electrodynamic Mössbauer spectrometer has been developed by the NRC “Kurchatov Institute”—PNPI (Gatchina, Russian Federation). The model represents a prototype using the neutron beam of the PIK reactor, which is generated using either constant or periodic activation of a γ-source, and isotope sources for structural and physicochemical studies of steel and alloys, minerals, catalysts, nanostructures, and biological objects. The article presents the physical characteristics of the model, as well as the sample preparation methodology and measurement modes with sample cooling to liquid nitrogen temperature. The experiments performed on a crystal hydrate with initially divalent iron admixed with Fe³⁺ have demonstrated the capabilities of the spectrometer for non-destructive testing of chemical reactions. The potential of the spectrometer in nanotechnology is illustrated by the results of studying magnetic graphene oxide composites and nanodiamonds grafted with europium ions.

This study presents a comparative analysis of the performance of helical and straight-tube steam-generating channels heated by liquid metals. In the straight-tube model, an unsteady heat transfer crisis was observed. Under specific operating conditions, dead zones were documented in both the economizer and superheater sections. The unsteady heat transfer crisis can be eliminated by installing throttling devices at the tube inlet, while the dead zones can only be eliminated by intensifying heat transfer. One approach to enhancing heat transfer is the use of helical tubes. In the helical steam-generating channel, no thermohydraulic instability was detected during longitudinal flow of the coolant along the heat exchange tubes. The findings indicate that helical steam-generating tubes exhibit a substantially higher efficiency when operated with transverse flow than with longitudinal flow.

One effective way to restore public confidence in nuclear power following severe accidents is to develop reactors that possess a high degree of inherent self-protection and passive safety features. Examples of such reactors include fast reactors cooled by heavy liquid metals, such as lead-bismuth and lead. These reactors lack the potential energy of compression and the chemical energy accumulated in the coolant which, under certain highly unlikely combinations of initiating events, could cause a release of radionuclides requiring the evacuation of the population. Due to the lack of many safety systems, nuclear power plants with heavy liquid metal coolants may be more competitive by their very nature. Lead-bismuth and lead-based coolants, like all others, have their own advantages and disadvantages. This article presents a comparative analysis of the characteristics of these coolants, their impact on the safety, reliability, and operational characteristics of reactors. It concludes that nuclear reactors with these coolants could find wide application in nuclear power after gaining operational experience with the first prototypes.

This article presents the engineering and physical design of a molten salt reactor (MSR) with circulating fuel based on LiF–BeF₂ salt mixtures, intended for the transmutation of neptunium (Np), americium (Am), and curium (Cm). Neutronic calculations demonstrate the feasibility of constructing a high-power MSR with a transmutation rate of at least 250 kg/year of Np, Am, and Cm. The proposed reactor layout suggests that a modular loop configuration is optimal for enhancing transmutation efficiency. In this design, the reactor, equipment, and primary circuit pipelines are arranged within a compact protective housing, which serves to minimize the volume of fuel salt outside the reactor core. Figure: 6, Tables: 4, References: 3.