Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m
−2
in the steady state of DT discharges, increasing to ~0.6–3.5 GW m
−2
under disruptions and ...ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.
•Sizing of major DEMO tokamak components.•Identification of relation between aspect ratio and magnetic field.•Identification of most compact DEMO design point.•Discards the choice of DEMO as a ...spherical tokamak.•Discards the choice of DEMO with a high aspect ratio and a high magnetic field.
The design point that had been chosen for EU DEMO in 2016 is reviewed here and a modification is proposed with a lower aspect ratio. Previously the same aspect ratio, A, was chosen for EU DEMO as in major tokamak experiments including ITER (A = 3.1), and, to rely on mature technology, a peak magnetic field no greater than 13 T was considered. Here we do not consider these limitations recognizing the recent commissioning of JT60-SA with A = 2.5 and the successful recent operation of a model coil at a field of >20 T.
EU DEMO must have a burning plasma and meet performance requirements relevant to a fusion power plant - at present, 2 GW fusion power and 2 h pulse length. The better plasma confinement at higher magnetic field allows reaching this condition in a smaller plasma. Thus, increasing the magnetic field appears as an obvious strategy to reduce the machine size. We confirmed though previous observations that the choice of a high magnetic field is associated with a large aspect ratio, mainly to generate space for the larger TF coils. In practice the magnetic field strength on DEMO-size TF coils is limited to ∼12 T by the high electromagnetic loads. Also, the extreme heat flux on the divertor increases further with the magnetic field. Hence the magnetic field on the plasma axis is limited in EU DEMO to ∼5.4 T, its aspect ratio to approximately 3.
The limiting factor to lowering the aspect ratio is the space on the inboard side. This is primarily driven by the requirement to integrate the central solenoid to drive the plasma current inductively. Our literature review suggests that non-inductive plasma scenarios, as considered in most power plant studies in literature, are optimistic and not sufficiently supported by experimental results. Also, the space required for the superconducting toroidal field coils, the tritium breeding blanket, and the neutron shield is substantial. For a DEMO device the space on the inboard side becomes insufficient for aspect ratios below ∼2.6. We therefore conclude the aspect ratio of EU DEMO should be chosen within the range ∼2.6 - ∼3.0 trading-off lower magnetic field and lower divertor heat loads against machine compactness.
Filaments or blobs are well known to strongly contribute to particle and energy losses both in L- and H-mode, making them an important plasma characteristic to investigate. They are plasma structures ...narrowly localized across a magnetic field and stretched along magnetic field lines. In toroidal devices, their development is observed to take place in the peripheral plasma. Filament characteristics have been studied extensively over the years using various diagnostic techniques. One such diagnostic is the Doppler backscattering (DBS) method employed at the spherical tokamak Globus-M/M2. It has been observed that the DBS signal reacts to the backscattering from filaments. However, the DBS data have proven difficult to analyze, which is why modelling was undertaken using the code IPF-FD3D to understand what kind of information can be extrapolated from the signals. A circular filament was thoroughly investigated in slab geometry with a variety of characteristics studied. Apart from that, the motion of the filaments in the poloidal and radial directions was analyzed. Additionally, other shapes of filaments were presented in this work. Modelling for the real geometry of the Globus-M/M2 tokamak was performed.
The diagnostic complex of the Globus-M2 spherical tokamak (
R
= 36 cm,
a
= 24 cm), the only operating tokamak in Russia with a divertor plasma configuration, which operates in the range of ...subthermonuclear temperatures (
T
e
to 1.6 keV,
T
i
to 4.5 keV) and densities (
n
e
to 2 × 10
20
m
–3
), is described. The Globus‑M2 tokamak is the unique scientific facility, which is a part of the Federal Center for Collective Use of the Ioffe Institute, Russian Academy of Sciences “Materials Science and Diagnostics in Advanced Technologies.” This allows third parties to perform their research using it. The work contains a list of all diagnostics currently available on the tokamak. The description of the diagnostics is structured in such a way that the reader gets an idea of their capabilities for measuring plasma parameters with an emphasis on the limits and accuracy of the measured values, and also spatial and time resolution. At the same time, many technical details are omitted in order to save space; references are given to papers with a more detailed description of individual diagnostics.
First experiments with nonaxisymmetric magnetic perturbations, toroidal mode number n=2, produced by newly installed in-vessel saddle coils in the ASDEX Upgrade tokamak show significant reduction of ...plasma energy loss and peak divertor power load associated with type-I edge localized modes (ELMs) in high-confinement mode plasmas. ELM mitigation is observed above an edge density threshold and is obtained both with magnetic perturbations that are resonant and not resonant with the edge safety factor profile. Compared with unperturbed type-I ELMy reference plasmas, plasmas with mitigated ELMs show similar confinement, similar plasma density, and lower tungsten impurity concentration.
Towards a compact spherical tokamak fusion pilot plant Costley, A E
Philosophical transactions of the Royal Society of London. Series A: Mathematical, physical, and engineering sciences,
03/2019, Letnik:
377, Številka:
2141
Journal Article
Recenzirano
Odprti dostop
The question of size of a tokamak fusion reactor is central to current fusion research especially with the large device, ITER, under construction and even larger DEMO reactors under initial ...engineering design. In this paper, the question of size is addressed initially from a physics perspective. It is shown that in addition to size, field and plasma shape are important too, and shape can be a significant factor. For a spherical tokamak (ST), the elongated shape leads to significant reductions in major radius and/or field for comparable fusion performance. Further, it is shown that when the density limit is taken into account, the relationship between fusion power and fusion gain is almost independent of size, implying that relatively small, high performance reactors should be possible. In order to realize a small, high performance fusion module based on the ST, feasible solutions to several key technical challenges must be developed. These are identified and possible design solutions outlined. The results of the physics, technical and engineering studies are integrated using the Tokamak Energy system code, and the results of a scoping study are reviewed. The results indicate that a relatively small ST using high temperature superconductor magnets should be feasible and may provide an alternative, possibly faster, 'small modular' route to fusion power. This article is part of a discussion meeting issue 'Fusion energy using tokamaks: can development be accelerated?'.
Important progress in the development of high-temperature superconductors (HTSC) of the second group made it possible to design the quasi-stationary tokamak with reactor technologies (TRT) with the ...high magnetic field (
B
t0
= 8 T). The high magnetic field will ensure the achievement of plasma fusion regimes in the tokamak with the fusion energy gain
Q
> 1 at the considerably reduced size of the facility (
R
0
= 2.15 m,
a
= 0.57 m), and, consequently, at its reduced cost. TRT will be capable of operating in the quasi-stationary regimes (≥100 s) with hydrogen, helium, and deuterium plasmas (with the densities
n
e
of up to 2 × 10
20
m
–3
) and in the regimes with short (duration Δ
t
< 10 s) deuterium–tritium plasma shots with the fusion energy gain
Q
> 1 limited by the radiation heating of toroidal coils. TRT is being designed as a plasma prototype for both the pure fusion reactor and the fusion neutron source for the hybrid (fusion–fission) reactor. The TRT missions are the development of the key fusion technologies and their integration in one facility. These technologies are as follows: the HTSC electromagnetic system operating at the extremely high magnetic fields; the metal and liquid-metal (lithium) first wall and innovative divertor; the unique advanced systems for the auxiliary plasma heating and non-inductive current drive, including the systems for atomic beam injection with energy of 0.5 MeV and power of several tens of megawatts, the electron cyclotron heating system based on the megawatt-power gyrotrons with a frequency of 230 GHz and a total power of ~10 MW, and the ion cyclotron heating system at frequencies of 60–80 MHz with a power of several megawatts; the tritium fuel cycle; the remote control technologies; the technologies for diagnostics capable of operating under the fusion reactor conditions; the technologies for maintaining quasi-stationary plasma discharges; and the technologies for the tokamak operation in the fusion ignition regime, in which the heating by alpha particles is the dominant heating mechanism at the axis of the plasma column, in the deuterium–tritium experiments limited by the radiation heating of the toroidal coils. The results are presented from the conceptual design of the basic TRT components, as well as the expected characteristics of its operation. It is shown that TRT has a wide window of working parameters suitable for studying the reactor operating regimes. The high magnetic field provides the necessary margins of the pressure, MHD stability, and plasma controllability variation. Implementation of the advanced divertor and first wall concepts, including those using the liquid-metal technologies, will provide the optimum choice of design options in order to reliably control the heat and particle fluxes under the reactor conditions. The advanced systems for the auxiliary heating and current drive will make it possible to implement both the pulsed and stationary regimes of the reactor operation. Calculations of the TRT discharge scenarios show that, for the DT mixture with equal content of components, the long discharges (with duration exceeding 100 s) can be realized with a neutron flux of more than 0.5 MW/m
2
onto the wall, as well as the stationary discharges with a flux of approximately 0.2 MW/m
2
. Thus, TRT can be a real prototype of the fusion neutron source for the hybrid reactor.
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