In recent years, machine-learning (ML) research methods have received increasing attention in the tokamak community. The conventional database (i.e., MDSplus for tokamak) of experimental data has ...been designed for small group consumption and is mainly aimed at simultaneous visualization of a small amount of data. The ML data access patterns fundamentally differ from traditional data access patterns. The typical MDSplus database is increasingly showing its limitations. We developed a new data management system suitable for tokamak ML research based on experimental advanced superconducting tokamak (EAST) data. The data management system is based on MongoDB and hierarchical data format version 5 (HDF5). Currently, the entire data management has more than 3000 channels of data. The system can provide highly reliable concurrent access. The system includes error correction, MDSplus original data conversion, and high-performance sequence data output. Furthermore, some valuable functions are implemented to accelerate ML model training of fusion, such as a bucketing generator, the concatenating buffer, and distributed sequence generation. This data management system is more suitable for fusion ML model research and development than MDSplus, but it cannot replace the MDSplus database. The MDSplus database is still the backend for EAST tokamak data acquisition and storage.
Abstract
Plasma-molecular interactions generate molecular ions which react with the plasma and contribute to detachment through molecular activated recombination (MAR), reducing the ion target flux, ...and molecular activated dissociation (MAD), both of which create excited atoms. Hydrogenic emission from these atoms has been detected experimentally in detached TCV, JET and MAST-U deuterium plasmas. The TCV findings, however, were in disagreement with SOLPS-ITER simulations for deuterium, indicating a molecular ion density (
D
2
+
) that was insufficient to lead to significant hydrogenic emission, which was attributed to underestimates of the molecular charge exchange rate (
D
2
+
D
+
→
D
2
+
+
D
) for deuterium (obtained by rescaling the hydrogen rates by their isotope mass). In this work, we have performed new SOLPS-ITER simulations with the default rate setup and a modified rate setup where ion isotope mass rescaling was disabled. This increased the
D
2
+
content by
>
×
100
. By disabling ion isotope mass rescaling: (1) the total ion sinks are more than doubled due to the inclusion of MAR; (2) the additional MAR causes the ion target flux to roll-over during detachment; (3) the total
D
α
emission in the divertor increases during deep detachment by roughly a factor of four; (4) the neutral atom density in the divertor is doubled due to MAD, leading to a 50% increase in neutral pressure; (5) total hydrogenic power loss is increased by up to 60% due to MAD. These differences result in an improved agreement between the experiment and the simulations in terms of spectroscopic measurements, ion source/sink inferences and the occurrence of an ion target flux roll-over. Extrapolating simplified scalings of divertor molecular densities (TCV & MAST-U) to reactor-relevant densities suggests the underestimation of molecular charge exchange could strongly impact divertor physics (neutral atom density, ions sinks) and hydrogen emission (which has implications for detachment control) in deeply detached conditions, warranting further study.
The effective treatment of the heat and power exhaust is a critical issue in the road map to the realization of the fusion energy. In order to provide possible, reliable, well assessed and on-time ...answers to DEMO, the Divertor Tokamak Test facility (DTT) has been conceived and projected to be carried out and operated within the European strategy in fusion technology. This paper, based on the invited plenary talk at the 31st virtual SOFT Conference 2020, provides an overview of the DTT scientific proposal, which is deeply illustrated in the 2019 DTT Interim Design Report.
Abstract
Multi-spectral imaging of helium atomic emission (HeMSI) has been used to create 2D poloidal maps of T
e
and n
e
in TCV’s divertor. To achieve these measurements, TCV’s MANTIS multispectral ...cameras simultaneously imaged four He I lines (2 singlet and 2 triplet) and a He II line (468nm) from passively present He and He
+
. The images, which were absolutely calibrated and covered the whole divertor region, were inverted through the assumption of toroidal symmetry to create emissivity profiles and, consequently, line-ratio profiles. A collisional-radiative model (CRM) was applied to the line-ratio profiles to produce 2D poloidal maps of T
e
and n
e
. The collisional-radiative modeling was accomplished with the Goto helium CRM code which accounts for electron-impact excitation and deexcitation (EIE), and electron-ion recombination (EIR) with He
+
. The HeMSI T
e
and n
e
measurements were compared with co-local Thomson scattering measurements. The two sets of measurements exhibited good agreement for ionizing plasmas: (5 eV ≤ Te ≤ 60 eV, and 2 × 10
18
m
-3
≤ ne ≤ 3 × 10
19
m
-3
) in the case of majority helium plasmas, and (10 eV ≤ Te ≤ 40 eV, 2 × 10
18
m
-3
≤ ne ≤ 3 × 10
19
m
-3
) in the case of majority deuterium plasmas. However, there were instances where HeMSI measurements diverged from Thomson scattering. When T
e
≤ 10 eV in majority deuterium plasmas, HeMSI deduced inaccurately high values of T
e
. This disagreement cannot be rectified within the CRM’s EIE and EIR framework. Second, on sporadic occasions within the private flux region, HeMSI produced erroneously high measurements of n
e
. Multi-spectral imaging of Helium emission has been demonstrated to produce accurate 2D poloidal maps of T
e
and n
e
within the divertor of a tokamak for plasma conditions relevant to contemporary divertor studies.
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.
Abstract
Microtearing instability is one of the major sources of turbulent transport in high-
β
tokamaks. These modes lead to very localized transport at low-order rational magnetic field lines, and ...we show that flattening of the local electron temperature gradient at these rational surfaces plays an important role in setting the saturated flux level in microtearing turbulence. This process depends crucially on the density of rational surfaces, and thus the system-size, and gives rise to a worse-than-gyro-Bohm transport scaling for system-sizes typical of existing tokamaks and simulations.
Fishtail divertor (FTD), a new divertor concept, is a significant research object of Experimental Advanced Superconducting Tokamak (EAST) tokamak divertor system. In order to meet the requirement of ...sinusoidal current precision and variable frequency, the combination of inductance and capacitance ( LC ) series resonance theory and high-frequency pulsewidth modulation (PWM) technology is put forward. A sine wave power supply with multifrequency points (1-3800 Hz) and current up to 8000 A is developed. The FTD power supply (PS) adopts dc switching power technology, sinusoidal PWM (SPWM) technology, and LC series resonance theory. The operating points of output current include 10-20 Hz at 8000 A, 30-110 Hz at 5000 A, and 1300-3800 Hz at 2400 A. The electromagnetic environment in the device is complex and arguable. The output current accuracy can reach 5% by the current feedback and frequency feedback control strategy. The test results demonstrate that the FTD PS can meet the system requirements, and it has strong reference significance, which can provide reference for the design of other divertor PS.
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.
The plasma diagnostic and control (D&C) system for a future tokamak demonstration fusion reactor (DEMO) will have to provide reliable operation near technical and physics limits, while its front-end ...components will be subject to strong adverse effects within the nuclear and high temperature plasma environment. The ongoing developments for the ITER D&C system represent an important starting point for progressing towards DEMO. Requirements for detailed exploration of physics are however pushing the ITER diagnostic design towards using sophisticated methods and aiming for large spatial coverage and high signal intensities, so that many front-end components have to be mounted in forward positions. In many cases this results in a rapid aging of diagnostic components, so that additional measures like protection shutters, plasma based mirror cleaning or modular approaches for frequent maintenance and exchange are being developed.
Under the even stronger fluences of plasma particles, neutron/gamma and radiation loads on DEMO, durable and reliable signals for plasma control can only be obtained by selecting diagnostic methods with regard to their robustness, and retracting vulnerable front-end components into protected locations. Based on this approach, an initial DEMO D&C concept is presented, which covers all major control issues by signals to be derived from at least two different diagnostic methods (risk mitigation).
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.
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