Physics of Atomic Nuclei最新文献

The created plasma device PLM-M is used to test the heat-shielding lining of the in-vessel components of a thermonuclear reactor with steady-state plasma having parameters similar to the near-wall and divertor plasma of a tokamak. The PLM-M is a linear magnetic trap with a multi-cusp magnetic field. A special feature of the device is the many hours of steady-state operation with magnetized plasma of high parameters. Tungsten modules manufactured using ITER divertor technology have been tested with high-heat loads from ~1 to ~5 MW/m². There were no significant macroscopic surface changes, large-scale cracks or splits on the surface, or significant erosion of the tungsten layout during plasma tests. For additional load on the tungsten during plasma tests, a laser with a power simulating the level of ELMs was used. Traces of arcs were detected on the plasma-facing tungsten surface of the model during plasma tests. Additional tests of cooled tungsten modules of in-vessel components are planned in order to estimate their erosion in the ITER and develop recommendations for in-vessel component design for the FNS.

In this article, on the basis of world experience in the field of handling tritium-containing media and on the basis of our own experimental data, a scheme of the tritium cycle system for a tokamak with reactor technologies is proposed. When calculating the characteristics of the tokamak tritium cycle system, it was assumed that the total volume of tritium in the facility is 20 g, and 0.1 g of tritium is consumed per pulse. The proposed scheme provides a complete recycling of tritium and consists of a subsystem for the sorption storage of deuterium and tritium on intermetallic compounds based on zirconium; a subsystem for the chemical purification of the exhaust gas mixture from the toroidal chamber of the tokamak trapping nitrogen, oxygen and carbon ions; a subsystem for the membrane separation of hydrogen isotopes from the gas mixtures; subsystems for the separation of hydrogen isotopes by thermal diffusion; subsystems for the purification of waste process streams by the method of phase isotope exchange, which provides for the preliminary catalytic oxidation of hydrogen isotopes to water; and a subsystem for the concentration of tritium-containing water by the method of chemical isotope exchange in the “water–hydrogen” system. Requirements are provided and analytical and radiometric equipment is proposed to ensure control of the necessary parameters and reliable protection of personnel, the public, and the environment from the effects of tritium.

The key features of armour material joining technology for ITER enhanced heat-flux first wall panels, supplied by the Russian Federation, are considered. To achieve uniform heating of the brazing zone, ensure the required properties of the precipitation-hardening heat-sink material (CuCrZr), and maintain the integrity of the bimetallic joints of the multilayer structure, it was necessary to refine and optimize the induction heating parameters. The developed vacuum induction furnace, special equipment, and jigs made it possible to achieve high repeatability of the quality of brazed joints. The vacuum induction brazing technology for beryllium armour has been qualified according to the rules of the ITER Organization, and a full-scale prototype has been manufactured and tested. The first successful experimental work on brazing of tungsten armour using this method was carried out.

Cryogenic fuel targets (CFTs) open up the possibility for the practical implementation of in-line technologies in inertial fusion energy (IFE) systems for the production of clean fuel and electric and thermal energy. This article presents the results of a new cycle of research in the field of constructing a repeatable layering module (LM) for CFTs with high energy yield. The aim of the work is the mass production of CFTs, which is of particular scientific interest to the IFE community. The LM construction principle is based on the free-standing target (FST) method, i.e., the formation of a solid fuel layer inside free-standing and line-moving targets, proposed and developed at the Lebedev Physical Institute (LPI). This technology formed the basis of the LPI project for the development of a specialized LM for the low-cost in-line production of CFTs and their repeatable noncontact delivery to the focus of high-power laser facilities operating in a pulse-periodic mode.

The divertor is one of the most energy-loaded elements of the ITER experimental tokamak. During the ITER operation, the divertor experiences both stationary thermal loads and rapid thermal impacts owing to transient processes in the plasma. The potentially most harmful transients during normal operation are edge-localized modes (ELMs). To mitigate the thermal impacts caused by ELMs, a current approach involves reducing the energy content of individual ELMs by increasing their frequency (up to 30–60 Hz). Because of the high pulse repetition rate, ~10⁸ ELM events are expected during the foreseen lifetime of the divertor components. Such a large number of pulses can lead to thermocyclic fatigue of the divertor material, the formation of a microcracks network, and melting along the edges of cracks as a result of failure of heat conduction. At the Budker Institute of Nuclear Physics, an experimental stand to study the performance of plasma-facing materials under the influence of a large (≥10⁷) number of ELM-like thermal impacts is being developed. To simulate the thermal impact on the material surface, it is planned to use a pulsed electron beam. In the present article, the prototype of an electron beam source for materials science research and the results of the beam characterization experiments are described. In experiments on electron beam generation, a beam current of 10 A at an accelerating voltage of 19 kV was achieved. The beam pulse duration of 1 ms at frequencies up to 10 Hz was demonstrated. Using imaging diagnostics based on luminescent ceramics, the beam current distribution was measured. The achieved beam parameters correspond to a specific power of 1.27 GW/m², which meets the requirements for materials science applications in the interests of fusion-class facilities.

This paper describes the control system for a dispersion interferometer based on a CO₂ laser developed by the Budker Institute of Nuclear Physics (Novosibirsk, Russia) for the Globus-M2 tokamak (Ioffe Institute, St. Petersburg, Russia). The basis of this system is the programmable controller ADAM-5510E/TCP with a set of analog and digital input/output modules and specially developed software that allows monitoring and control of the operation and parameters of interferometer elements. During the year of the dispersion interferometer operation as a standard diagnostic at the Globus-M2 tokamak, the control system showed good levels of reliability and sustainability.

Since the 1990s, comprehensive studies of the interaction of powerful plasma streams, electron beams, and laser radiation with solid bodies have been carried out at the Budker Institute of Nuclear Physics under loads expected at the first wall of a fusion reactor. First, the experiments have been performed using the GOL-3 facility, then the BETA bench. At the moment, new experimental facilities are being set up, including the Plasma Station at the under-construction SKIF synchrotron radiation source, and a new-generation open-trap GDMT reactor is being designed. The properties of near-surface plasma and its dynamics (density, temperature, velocity, charge composition) have been investigated using the developed diagnostic complex of the GOL-3 multimirror trap. Surface modification of various materials (tungsten, graphite, ceramics) under different pulse thermal loads has been studied. The use of the electron beam on BETA bench allowed in situ observation of surface modification processes during thermal shock, which is unavailable for plasma impact facilities. Comparison with the data of other researchers has been performed and their consistency with each other has been shown. Thresholds for cracking and melting of tungsten, explosive (brittle) fracture of graphite have been obtained. Synchrotron radiation has been used to study stresses in materials, with which stresses during pulsed irradiation have been measured with high temporal resolution. Experimental work is supplemented by theoretical and computational studies. A model of tungsten fracture is proposed according to which stresses leading to tungsten cracking occur at the cooling stage. For the first time in the world, the appearance of cracks significantly later than the thermal load has been experimentally observed using the BETA bench.

The results of the experimental determination of the spallation strength of polymer materials are presented using samples made of acrylic glass and polystyrene as examples. The study was conducted in the energy density range of 35–1500 J/cm². Despite the similar physical and technical parameters of the examined polymers (acrylic glass, polystyrene, epoxy resins), the nature of their failure under pulsed volumetric energy release is different [1, 2]. It has been shown that, in polystyrene, cracks originate directly from the energy release zone, while in acrylic glass, a transparent undamaged area is observed next to the energy release zone, beyond which cracks form. The spallation strength of polystyrene was determined to be σ = 0.35 GPa, which is similar to the spallation strength of acrylic glass. This is interesting, as the spatial localization of the failure zones in samples made from these two materials differs significantly.

In the design of the DEMO fusion reactor, as well as the facility of a fusion neutron source FNS, as one of the options, it is assumed that a three-loop cooling system will use a blanket of liquid metals. Liquid lithium from the primary loop transfers heat in intermediate heat exchangers to liquid sodium of the second loop, which in turn transfers heat to the water of the third loop in the steam generator. The possibility is considered of using a “neutral” coolant in the second loop—a eutectic lead–bismuth alloy, which does not chemically interact with either liquid lithium or water in the third loop, which makes it possible to exclude contact of liquid lithium or sodium with water. A preliminary calculation of the intermediate liquid lithium–eutectic alloy heat exchanger has been performed, and it is shown that the use of this alloy will allow for the cooling of the fusion reactor with the same number of intermediate heat exchangers and without a significant increase in their size.

The influence of pulsed plasma irradiation on a boron carbide (B₄C) coating using the QSPA-T setup was investigated. The duration of the rectangular plasma pulses was 0.5 ms, with intervals between pulses of 5 to 10 min. The maximum power density in the middle of the plasma flow reached 1 GW/m². The thickness of the coating at various surface locations ranged from 20 to 40 μm. The modification of the surface layers and the transformation of the coating at elevated temperatures during the pulsed plasma irradiation process over four consecutive series of pulses are described. It was shown that the boron carbide coating withstood a complete cycle of tests under irradiation with 100 plasma pulses at a maximum power density of 1 GW/m². No significant damage was detected on the surface of the coating, and the layer remaining after the tests maintained the crystalline structure of B₄C throughout the irradiation zone to a depth of at least 2 μm.