The interaction of peripheral D–Li-plasma with the protective ceramic elements of electrical porcelain, EP (xAl2O3⋅ySiO2), on the limiter of the T-11M tokamak has been studied. The calculated theoretical estimate of the maximum possible lithium capture is 6 mg (5 μg/discharge). The authors’ experimental method, based on the acid leaching with formic acid of the products of the interaction of EP with elemental lithium from the surface of dismantled protective ceramic elements, allowed us to obtain a practical estimate of Li-sorption equal to 3 mg (2.5 μg/discharge) ( 50
The SMOLA facility was built at the Institute of Nuclear Physics, Siberian Branch of the Russian Academy of Sciences (INP SB RAS) to experimentally test the concept of helical confinement. This method is a development of the multiple‑bounce confinement scheme. A stream of rotating plasma flows out of the open trap through a magnetic field with helical symmetry that is periodically modulated along the field line. Confinement efficiency depends strongly on the ratio of the ion mean free path to the period of the magnetic‑field modulation. Experiments on the SMOLA facility confirmed improved plasma confinement both in the regime dominated by Coulomb collisions and at reduced collision frequency. Simulation of plasma flow regimes for next‑generation facilities requires a further reduction of the dimensionless Coulomb collisionality. For this purpose a SMOLA* installation is being constructed at INP SB RAS. The paper discusses the new magnetic configuration with mirror‑symmetric helical end‑plugs, an auxiliary ion‑cyclotron resonance (ICR) heating system, methods for generating plasma rotation, and features of the diagnostic suite.
During the work on integrating the diagnostics into the ITER ports, a problem arose with providing the required radiation protection. It was proposed that a large amount of a neutron-absorbing material, boron carbide, be placed in the vacuum chamber in the port plug. The chemical composition and the thermal and vacuum properties of various B4C ceramics from several manufacturers were studied. Sintered ceramics in the form of blocks with holes were selected. A specification for the production of ceramics for ITER ports was agreed upon. Experiments on irradiation of ceramics with neutrons at the VITA accelerator were carried out. The degradation of mechanical properties after prolonged fast neutron irradiation was investigated. Serial production of sintered ceramics was started. The ITER Organization has approved the End of Manufacturing reports on the serial batches of ceramics.
Tokamak T‑15MD gradually reaches the design parameters. At the moment, an important task is to obtain stable discharges with diverter configurations and a plasma current of at least 500 kA with a pulse duration at the quasi‑stationary stage of the order of 1 s. To effectively solve this problem, it is necessary to support the experiments with numerical modeling both in the interval between the experimental campaigns and directly during the campaign. The use of numerical analysis of experiments in the 2025 T‑15MD campaign made it possible to optimize plasma discharge scenarios and work out an active feedback system in relation to the vertical and horizontal displacements of the plasma column. Stable discharges were obtained with a diverter and a plasma current of 500 kA and a duration at the quasi‑stationary stage of the order of 1 s. The paper summarizes the experimental experience and numerical calculations obtained during the spring campaign of 2025 and provides recommendations for obtaining one of the discharge scenarios with a diverter on the T‑15MD tokamak unit.
The article describes a method for controlling the geometric parameters of the cooling system channels of the diagnostic shielding modules of the equatorial port no. 11 of the ITER nuclear installation made using counter deep drilling technology. The technique is based on the use of an angular reflector moving through the test channel and a laser tracker. The results of the qualification of the methodology on a full-size test product modeling the deepest channels in the diagnostic shielding module are presented.
This study investigates the behavior of an Al–AlN–TiB2 ceramic composite during heating and provides a preliminary assessment of its compatibility with hydrogen plasma. The study included an analysis of the atomic flux from the surface during heating, sample mass loss, and deuterium retention during plasma irradiation. It was found that free aluminum diffuses to the surface at approximately 600°C and evaporates from the material at temperatures above 900°C. Regular surface breakdowns and microparticle emission were observed during exposure to deuterium plasma. Deuterium retention in the material was relatively low and varied little with fluence at low irradiation temperatures.
The classic method of evaluating the plasma energy content Wdia and the poloidal beta βp is to perform and analyze diamagnetic measurements. This article discusses a preliminary version of the T-15MD tokamak (R = 1.48 m, a = 0.67 m, B0 = 2.0 T, Ip = 2 MA) diamagnetic measurement system. The results of modeling the signals of the diamagnetic measurement system are analyzed, and the algorithm for obtaining the plasma-related toroidal magnetic flux from experimental data is presented. The composition, design, and arrangement of the primary sensors of diamagnetic loops and compensation coils proposed for installation on the T-15MD tokamak are considered.
A heavy-ion beam injector is being developed at the accelerator facility of the Kurchatov complex of theoretical and experimental physics at the National Research Centre “Kurchatov Institute” according to the thematic plan. The injector consists of a laser-plasma ion source and a linear accelerator. Ion beams are extracted from laser-produced plasmas created by irradiating solid targets with high intensity CO2 laser pulses and matched into a linear accelerator with a low-energy beam transport system. The plan envisages the use of the laser plasma ion source with already commissioned linear accelerators I-3 and I-4. This work describes physical principles of the production of laser pulses with stabilized temporal and spatial properties, and the design of the repetition rate CO2 laser system FOKUS, which is currently in operation at the accelerator facility of the Kurchatov complex of theoretical and experimental physics, and can be successfully used with other accelerators.
During accidents at tokamak-like thermonuclear facilities (TNF) initiated by the loss of coolant from the cooling system of the first wall or divertor into the vacuum vessel (VV), steam ejected into the VV reacts exothermically with hot metal surfaces and dust. This is accompanied by the formation of hydrogen and produces a risk of explosion of the hydrogen-air mixture when hydrogen escapes outside the VV or when air enters the VV. Safety analysis of TNF requires numerical estimates of the amount and rate of hydrogen production, the composition of the gas atmosphere, and the possibility of hydrogen stratification. For these purposes, IBRAE RAS is developing the integral code PHAETON and the CFD code CABARET-T. The paper presents a description of interfaces for performing coupled calculations using these codes and provides the results of test calculations for a simplified model of the ITER VV.
Experimental studies involving multiple ( 104 cycles) exposures of ELM-like pulses revealed key erosion patterns in AlN–TiB2 composite ceramics. The primary degradation mechanism was found to be damage to the aluminum-containing phase, while significant erosion of refractory TiB2 occurred only after substantial removal of the low-melting-point component consisting of free aluminum and aluminum nitride. Although the aluminum-containing phase demonstrates relatively low thermal shock resistance, it plays a crucial role by filling pores and enhancing both thermal conductivity and mechanical strength of the material. For practical implementation of TiB2-based materials, two key challenges must be addressed: optimization of additive materials and refinement of manufacturing processes to reduce material porosity. The use of aluminum-containing materials in plasma-facing components of deuterium–tritium fusion devices may lead to accumulation of the long-lived radioactive 26Al isotope. Nevertheless, the results obtained demonstrate the potential of composite materials combining refractory and low-melting-point constituents.
Creating large scale magnet systems for thermonuclear fusion devices requires new approaches for development of superconducting cables. Owing to a number of specific properties of high-temperature superconductors, such superconducting cables cannot be based on the well-developed template of low-temperature cables. In this paper, we describe the concept of so called subcables—prefabricated stabilized stacks of HTS tapes. The process of development is described, as well as experimental properties of different samples of subcables: the paper shows different designs of subcables as well as the means for their production; critical current measurements at liquid nitrogen temperature for different bending diameters are described. The paper discusses the results and gives conclusions on future usage of subcables for high-current cables.
The article presents the results of the analysis of the density profile peaking factor and the inverse density gradient scale length (IDGSL) after switching-on the on-axis electron cyclotron resonance heating (ECRH) in the T-10 tokamak. Two series of experiments with different chord-averaged densities in the ohmic regime before the ECRH switching-on are considered. In each series, the total plasma current was changed from shot to shot at a fixed chord-averaged density and the ECRH power. In this case, the ECRH power was several times higher than the ohmic heating power. It was shown that at the ECRH stage, the density profile peaking factor and the IDGSL become almost the same in each series of experiments. This circumstance indicates the formation of the “stiff” IDGSL with high-power on-axis ECRH. Such property of stiffness for the IDGSL requires its consideration in transport models, when we analyzing the density evolution.
The history of assessments of the lifetime of ITER divertor components under intense plasma loads is considered, from the conceptual-design stage in the 1990s to the present. The properties of various materials considered as candidates for the first wall are compared. Mechanisms of material erosion under plasma exposure are examined. The main methods for studying material resistance to intense plasma fluxes are described, with references to experimental results in this area. Problems arising from ITER’s transition to an all-tungsten first wall are outlined.
The problem of monitoring the impurity influx into the plasma is examined within the framework of a new baseline of ITER operation with a tungsten first wall. Replacing beryllium with tungsten increases a resistance of the wall to thermal loads and erosion but creates a risk of plasma contamination by a heavy impurity (tungsten) and complicates the pumping-out of light impurities. To solve these problems in ITER, the periodic wall boronization is proposed. A special attention is paid to the analysis of capabilities of H-alpha and Visible Spectroscopy (HA VS) diagnostic to perform monitoring the impurities, primarily boron and tungsten, in the edge plasma layer (SOL) under the new ITER operational baseline. A list of spectral lines for recording the radiation from boron atoms and ions, as well as tungsten, is proposed. Furthermore, potential issues, such as the divertor stray light and the overlapping of spectral lines from different elements, are discussed. Using the numerical modeling (BM1D2V, SDTrimSP, SOLPS, and WallDYN3D codes), flux densities of the boron and tungsten atoms from the first wall due to sputtering by the deuterium atoms and ions were estimated for two characteristic divertor operating regimes. It is shown that the protective boron layer effectively shields the tungsten wall from an erosion but leads to the significant boron influx into the plasma. For radiation intensity calculation on HA VS diagnostic cameras the Raysect and Cherab libraries were used. The calculations of radiation intensity on the selected spectral lines of boron demonstrate the feasibility of their recording by the HA VS system and emphasize the importance of taking into account the wall reflectivity for the accurate signal interpretation.
This article considers approaches to fusion safety regulation accepted or suggested worldwide. Substantial differences in approaches used in individual countries are demonstrated. Mild regulatory regimes may be considered an advantage in a “fusion race”, especially if private investment is involved. At the same time, substantial uncertainties remain that hinder safety justification for fusion facilities. Failure to adequately address these uncertainties may have adverse implications for fusion development worldwide. A number of approaches currently suggested internationally may be recommended for implementation into Russian regulatory practices, including: gradual licensing of fusion facility with license conditions limiting allowed facility parameters at each stage, wider use of graded approach and consideration of a broader range of safety justification methods.
The processing procedure of data of MHD diagnostics for tokamaks with circular and vertically elongated plasma cross sections is presented. The utilized processing of the signals of MHD diagnostics based on the cross-S-transform reduces the fraction of incoherent components in the signals and also provides the spatial spectrum of coherent components and their time–frequency distribution.
At the SRC RF TRINITI (Moscow, Troitsk), a plasma gun with a new gas injection system is being developed as part of the task of creating a neutron source. In the new injection system, gas is injected into the working volume using fast-acting valves in the direction from the outer to the inner electrode. This establishes the gas distribution with the increased density near the central electrode. A Laval nozzle is used to achieve azimuthal uniformity of the flow. The objective of this work has been to study the plasma-forming gas injection into the interelectrode gap of the new plasma gun. In the course of the study, a numerical model of the injection system has been developed. The hydrodynamic equation system is solved using the finite volume method on a moving mesh, which made it possible to take into account the motion of the gas valve stem. The gas-dynamic variables on the next time layer are calculated using an implicit scheme. The gas pressure distribution in the interelectrode gap is measured with pressure transducers, and the gas flow parameters are estimated. Based on the results obtained, the numerical model is validated. The developed numerical model adequately describes the behavior of the gas flow in the interelectrode gap.
At present, work is underway to create a complex of a tokamak with reactor technologies (TRT). An atomic beam is considered as one of the main tools of additional heating of tokamak plasma. Three twin deuterium injectors are proposed for the TRT. The complex injector system requires an appropriate infrastructure. The paper proposes solutions for maintaining the injectors in the TRT experimental hall, engineering of power supply systems, cryogenic systems, cooling systems, pumping facilities, and others.
Using liquid lithium to protect the first wall of tokamaks is a common idea implemented in many installations. In 2022, a lithium droplet injector (AO NIKIE) was developed to conduct real-time lithization during the discharge of T-11M tokamak, achieving excellent results. One of the drawbacks of this injector was the limited amount of lithium in the feed tank, which required its removal from the vacuum chamber of the tokamak and its depressurization for refueling. In 2024, a lithium injector with an external lithium refueling system without depressurizing the tokamak was developed. This paper presents the results of testing such an injector in the plasma discharges of the T-11M tokamak. Its performance was demonstrated experimentally; ten refuelings were performed without depressurizing the vacuum chamber of the tokamak. The main operating parameters of the injector were determined: the lithium microdroplet velocity ranged from 1.5 to 7 m/s, and the injected lithium flow varied from 10 to 70 mg/s. The effect of lithium droplet injection on the discharge parameters was studied. Calculations (modeling) of the distribution of ionized and neutral components of lithium in the plasma column were carried out.
Development of thermonuclear devices requires constant experimental iterations both at the level of constructional materials and at the level of whole systems and their elements. Building testing facilities able to perform superconducting magnet systems of a thermonuclear reactor test is an important task of high priority, since providing certain experimental options can shift the whole concept of thermonuclear energy development. Powering systems of such facilities are the key elements that define the specialization of the facility as a whole. This article describes the structural parts and development process of 40 kA HTS current leads that are under construction for superconducting coils of various sizes intended for use in testing facilities.
