Plasma Confinement Research Articles

Multi-granularity short circuit fault diagnosis for tokamak power converters under small sample conditions and noise interference

The power supply system of a tokamak device is a crucial subsystem responsible for providing the magnetic field for plasma confinement. The timely application of fault diagnosis techniques to identify abnormal states of the power supply is of crucial importance for the safe and efficient operation of tokamak devices. Intelligent diagnostic methods based on Convolutional Neural Networks (CNN) are widely employed nowadays. However, within the operational environment of the device, there exist multiple physical field couplings, and inevitable noise will impact the data. Moreover, the measurement equipment exhibits instability, resulting in limited availability of data that meets the required criteria. To address these issues, this paper investigates a fault diagnosis approach utilizing an Auxiliary Classifier Generative Adversarial Network (ACGAN) for data augmentation, and proposes a Multi-Granularity Denoising CNN (MGDCNN) that accounts for noise interference, enabling diagnosis of faults at different granularities. In our approach, we first construct a multi-granularity structure of the data based on prior knowledge. Then, a hierarchical structure corresponding to the multi-granularity data and robust to noise is built by introducing an embedded Discrete Wavelet Transform (DWT) into the one-dimensional (1-D) CNN. Finally, the natural stratification of the network is combined with individual granularity branches for fault diagnosis. Experimental results on simulated datasets and public benchmark datasets demonstrate that the proposed model achieves a diagnostic accuracy of 98.81 % for short-circuit faults in power supply converter at the finest granularity. Even with noisy data, the model maintains a diagnostic accuracy of approximately 97 %, indicating superior noise resistance performance in the inference process of the MGDCNN network.

Integrated analysis of plasmas rotation effect on HL-3 hybrid scenario

Abstract The hybrid scenario, which has good confinement and moderate MHD instabilities, is a proposed operation scenario for ITER. In this work, the effect of plasma rotation on the HL-3 hybrid scenario is analyzed with the integrated modeling framework OMFIT. The results show that toroidal rotation has no obvious effect on confinement with a high line averaged density of n bar~7×1019m-3. In this case, the ion temperature only changes from 4.7 keV to 4.4 keV with the rotation decreasing from 105 rad/s to 103 rad/s, which means that the turbulent heat transport is not dominant. While in scenarios characterized by lower densities, such as n bar~4×1019m-3, turbulent transport becomes dominant in determining heat transport. The ion temperature grows from 3.8 keV to 6.1 keV in the core as the rotation increases from 103 rad/s to 105 rad/s. Despite the rise in ion temperature, the effects of rotation on both electron temperature and density are not obvious. Additionally, it is noteworthy that the variation in rotation does not significantly affect the global confinement of plasma in both scenarios with low and high density.

Plasma confinement by an optoelectronic system

Plasma confinement was succeeded by an optoelectronic system with the aid of a vacuum ultraviolet (VUV) light source, called the photoemission-assisted plasma system. The photoemission-assisted plasma was generated by utilizing photoelectrons from the substrate cathode. The photoelectrons were emitted from the substrate by external VUV irradiation via the photoelectric effect and then worked as initial electrons triggering the plasma generation. The photoemission-assisted plasma was confined with bright luminescence in an argon atmosphere by controlling the flow rate and pressure. The plasma confinement survived at up to 6400 Pa, which was much higher than the pressure estimated from the current–voltage characteristics. These results suggested that the area exhibiting luminescence dominated by the γ regime becomes small as the argon flow rate increases; however, the area does not vanish because the VUV-excited photoelectrons are sufficiently supplied. The residual area is dominated by the α regime without luminescence. Thus, the photoemission-assisted plasma seems to be confined on the balance between α and γ regimes. Because the current in the α-regime area is one hundredth in magnitude compared with that in the γ-regime area, the actual current density results in over 40 times with strong luminescence. This confined plasma with certain voltage and current condition may be expected for developing a new plasma reaction system and for application in semiconductor engineering.

Electron ITB formation in EAST high poloidal beta plasmas under dominant electron heating

Plasma confinement and transport in tokamaks play a crucial role in the development of high poloidal beta steady-state operation scenarios. Therefore, it is very important to study the relevant mechanism of internal transport barriers (ITBs), which can help plasmas to obtain better confinement in order to achieve higher fusion gain. This paper mainly introduces the analysis of the characteristics of electron heat transport of discharges with ITB in high βP operation regime in Experimental Advanced Superconducting Tokamak (EAST). Based on the statistical analysis of stable discharges with βP > 1.5, it is found that there is an obvious bifurcation of the normalised electron temperature gradient (ETG) ( R/LTe ) in the range of βP = 2–2.2. Then the discharges of lower βP ( βP < 2, where the value of βP is below the bifurcation threshold) and of higher βP ( βP > 2.2, where the value of βP is above the threshold) were selected for analysis. The diagnostic data provided by Thomson scattering, x-ray crystal spectrometry and charge exchange recombination spectroscopy are used to provide reliable parameter profiles and then combined with the data of external magnetic probe measurements and the polarisation interferometer diagnosis system to fully reconstruct the balance. A relevant plasma current calculation model is used to calculate and analyse the current density profiles and power deposition, and then the transport analysis is carried out. Interestingly, in the higher βP discharges, it is found that the turbulence intensity provided by the CO2 laser collective scattering system gradually decreases and the normalised ETG R/LTe gradually increases. It is also found that the electron heat transport coefficient decreases in the discharge with higher βP and the growth rate of the electron-scale turbulence calculated by transport gyro-Landau fluid (TGLF) is significantly reduced. Meanwhile, similar conclusions are also obtained in the discharges when the βP is further increased.

Effect of toroidal rotation on impurity transport in tokamak improved confinement

The centrifugal force effects from toroidal rotation in improved confinement plasmas are analyzed on high-Z impurities in tokamaks. Tungsten (W) transport simulations are performed using the impurity transport code developed in the integrated code TASK. The geometric factors PA and PB are introduced into the neoclassical transport coefficients to include the effects of the toroidal rotation, which come from poloidal asymmetry in the high-Z impurity profiles. Inward neoclassical particle pinch driven by the main ion density gradient is enhanced by the poloidal asymmetry to be the dominant mechanism for W accumulation in the plasma central region. Simulations with experimental plasma profiles show good agreement with the experimental results and first-principle simulation results in the H-mode. In the hybrid mode and advanced mode, the impurity accumulation is enhanced in the internal transport barrier (ITB) regions. The condition to suppress impurity accumulation is investigated by calculating dependencies on the toroidal rotation velocity and ITB position. The neoclassical transport is sufficiently small with the prospected ITER condition of the Mach number of main ions Mi ∼ 0.1. The impurity transport inside the ITB is strongly influenced by competition between the density peaking effect and the temperature screening effect, and the present simulations show suppression of the impurity accumulation with the outer ITB position to improve the plasma performance, due to the relatively larger temperature gradient of the main ion.

Prospects of negative triangularity tokamak for advanced steady-state confinement of fusion plasmas in MHD stability consideration

The steady-state confinement, beta limit, and divertor heat load are among the most concerned issues for toroidal confinement of fusion plasmas. In this work, we show that the negative triangularity tokamak has promising prospects to address these issues. We first demonstrate that the negative triangularity tokamak generates the filed line rotation transform more effectively. This brings bright prospects for the advanced steady-state tokamak scenario. Given this, the MHD stability and equilibrium confinement of negative triangularity tokamak are investigated. We point out that the negative triangularity configuration with a broad pressure profile is indeed more unstable for low-n magnetohydrodynamic modes than the positive triangularity case so that the H-mode confinement can hardly be achieved in this configuration, where n is the toroidal mode number. Nevertheless, we found that the negative triangularity configuration with high bootstrap current fraction, high poloidal beta, and peaked pressure profiles can achieve higher normalized beta for low-n modes than the positive triangularity case. In a certain parameter domain, the normalized beta can reach about twice the extended Troyon limit, while the same computation indicates that the positive triangularity configuration is indeed constrained by the Troyon limit. This shows that the negative triangularity tokamaks are not only favorable for divertor design to avoid the edge localized modes but also can have promising prospects for advanced steady-state confinement of fusion plasmas in high beta.

Drift wave soliton formation via beat-driven zonal flow and implication on plasma confinement

In this work, gyrokinetic theory of drift waves (DWs) self-regulation via the beat-driven zonal flow (ZF) is presented, and finite diamagnetic drift frequency due to plasma nonuniformity is shown to play a dominant role in the ZF beat generation. The obtained nonlinear DW equation is a nonlinear Schrödinger equation, in which the linear dispersiveness, linear growth, nonuniformity of diamagnetic drift frequency, and cubic nonlinearity induced by the feedback of beat-driven ZF to DWs are self-consistently included. The nonlinear DW equation is solved numerically in both uniform and nonuniform plasmas. It is shown that the DW envelope soliton may form due to the balance of linear dispersiveness and nonlinearity and lead to turbulence spreading to linearly stable region. It is further found that though the threshold on the DW amplitude for soliton formation is well within the relevant parameter regimes of realistic tokamak experiments, solitons cannot extend beyond the range bounded by the turning points of the wave packet when plasma nonuniformity is self-consistently accounted for.

Recent results on high-β plasma confinement studies in the Gas Dynamic Trap

This paper is devoted to experimental studies of plasma confinement with high relative pressure ( $\beta$ ) in the Gas Dynamic Trap (BINP, Novosibirsk). In previous high- $\beta$ confinement studies a maximum local $\beta = 0.6$ was achieved in the fast-ion turning point, contributed to by a beam-driven population of fast ions with an anisotropic distribution function. In this study the axial magnetic field profile was modified to bring the turning points closer to one another, which effectively increased the energy density of plasma and pushed the $\beta$ value higher. Experiments were performed for two non-standard magnetic configurations, where the axial fast-ion confinement region length was reduced by 1.5 and 2 times compared with the standard configuration. The average values of $\langle \beta _{\perp } \rangle$ over the plasma central cross-section were found to be 0.1 and 0.18, respectively, for the two configurations, with the latter value significantly exceeding the $\langle \beta _{\perp } \rangle =0.08$ of the standard configuration, in which the previous record was set. Moreover, halving the fast ion confinement region almost doubled the D–D fusion proton flux from the trap centre compared with the standard configuration. The electron temperature in both new magnetic configurations was only slightly smaller than in the standard configuration. In addition, an effect of Alfvén ion–cyclotron instability (AICI) development on the pressure in the turning points is discussed. Presumably, with some decrease in magnetic field an evolving AICI does not result in considerable pressure axial redistribution, so the pressure maximum is in the turning points’ vicinity despite the instability.

Progress of open systems at Budker Institute of Nuclear Physics

This paper is based on a report at the 2nd International Fusion Plasma Conference &amp; 13th International Conference on Open Magnetic Systems for Plasma Confinement (iFPC &amp; OS 2023), August 21–25, 2023, Busan, Korea and provides a brief overview of the status of work at the Budker Institute on the study of hot plasma confinement in open-type magnetic traps with a linear axisymmetric configuration. The main attention is paid to key problems: magnetohydrodynamics (MHD) stability in regimes with extremely high relative pressure, longitudinal electronic thermal conductivity, stability with respect to the development of kinetic modes and transverse transport. This paper provides an overview of the methods we are developing to address these problems, the experimental and theoretical results achieved and plans for future development. The last section of the article provides brief information about the preliminary design of the gas-dynamic multiple-mirror trap device, the development of which has been completed.

RELAP5/Mod3.3 MHD module development and validation: WCLL-TBM mock-up model

Magnetohydrodynamic (MHD) phenomena significantly impact the design of magnetic-confinement fusion reactors, particularly in the context of breeding blankets (BB) utilizing liquid metals (LMs) as working fluids. These effects arise from the interaction between the electro-conductive flowing metal and the magnetic field used for plasma confinement in the reactor chamber. Induced electrical currents in the liquid generate Lorentz forces, thereby altering flow behaviour in comparison to standard hydrodynamic conditions. For example, MHD effects modify velocity distribution and mass transport within ducts, amplify pressure losses, and influence heat transfer mechanisms. Accurate estimation of these impacts is crucial for the effective design of a liquid metal breeding blanket. While computational tools are essential for fusion-related physical analyses, no comprehensive MHD code currently exists for simulating all relevant phenomena in a liquid metal blanket. In this context, models predicting both distributed and concentrated MHD pressure drops have been integrated into the thermal-hydraulic system code RELAP5/Mod3.3. The Verification and Validation (V&V) process compares code results to direct numerical simulations and experimental data. For validation, RELAP5 recreates experimental results of a Water Cooled Lithium Lead test blanket module at magnetic field intensities ranging from Ha=500−3000, confirming the reliability of the newly implemented MHD subroutines for predicting pressure drops within this parameter range.

A control oriented strategy of disruption prediction to avoid the configuration collapse of tokamak reactors

The objective of thermonuclear fusion consists of producing electricity from the coalescence of light nuclei in high temperature plasmas. The most promising route to fusion envisages the confinement of such plasmas with magnetic fields, whose most studied configuration is the tokamak. Disruptions are catastrophic collapses affecting all tokamak devices and one of the main potential showstoppers on the route to a commercial reactor. In this work we report how, deploying innovative analysis methods on thousands of JET experiments covering the isotopic compositions from hydrogen to full tritium and including the major D-T campaign, the nature of the various forms of collapse is investigated in all phases of the discharges. An original approach to proximity detection has been developed, which allows determining both the probability of and the time interval remaining before an incoming disruption, with adaptive, from scratch, real time compatible techniques. The results indicate that physics based prediction and control tools can be developed, to deploy realistic strategies of disruption avoidance and prevention, meeting the requirements of the next generation of devices.

Experimental issues of energy balance in open magnetic trap

The paper presents an overview of experimental results of an investigation of different energy loss channels in the gas dynamic trap (GDT), which is a magnetic mirror plasma confinement device in the Budker Institute of Nuclear Physics. Energy losses along magnetic field lines are considered as well as losses onto radial limiters, which restrict the plasma column radius and provide its magnetohydrodynamic stability via the ‘vortex confinement’ mechanism. The losses along the field lines were measured using a set of pyroelectric bolometers on the plasma absorber and the losses onto the limiters were determined with thermistors from their temperature rise. Additionally, the losses due to charge exchange of fast plasma ions on the residual neutral gas in the GDT were measured using a longitudinal array of pyroelectric bolometers mounted on the wall of the central cell. An attempt was made to draw up the energy balance in the GDT in order to identify the predominant loss channels and reduce those losses in the future.

Direct fusion drive based on centrifugal mirror confinement

A concept for direct fusion drive based on centrifugal mirror (CM) confinement of thermonuclear plasmas (DFD-CM) is described. In centrifugal mirrors, electric and magnetic fields are combined to confine the plasma within a rapidly rotating annulus of burning plasma fixed between two mirror magnets. High energy fusion products leave the reactor core at a rate determined by the velocity of plasma rotation and the strength of the mirrors. Those departing through the jet-side mirror deposit their energy in a “warm” plasma that then expands through a magnetic nozzle to deliver jet power in the 100–1000 kW range. Fusion products departing through the power-side mirror are converted to electricity to power the reactor. Moderate thrusts at attractive specific impulses (10,000+ seconds) are possible. Findings are presented on CM reactor dynamics in propulsion applications, to include new insights into the relationship between mirror and centrifugal components of plasma confinement. Additionally, analysis will be presented on reactor operability limits and characterization of viable configurations based on power density, technology constraints, and the ability to self-power. Ongoing research into the physics of the warm plasma will be discussed, to include description of improved fidelity estimates for fusion energy deposition. Finally, considerations for Alfvén’s frozen-in theorem relative to fusion plasmas and magnetic nozzle performance will be outlined.Viable space commerce and future space exploration will require advanced power and propulsion technologies capable of multi-MW power generation with high specific impulse, moderate thrust levels, and low system specific mass. DFD could make a round trip to Mars in 3 months, and to Saturn, in less than 3 years. A DFD-CM powered mission could deliver an orbiter and 4 atmospheric probes to Neptune in four years. A second-generation Interstellar Probe powered by DFD-CM could reach the edge of the heliosphere (1000 astronomical units) in 25 years.

Lawson Criterion Analysis of D-3He Fusion Reaction

The Lawson criterion constitutes a pivotal condition for the viability of sustained fusion reactions. This study employs numerical methods and data analysis in Microsoft Excel to ascertain the Lawson criterion for D-³He fusion at temperatures ranging from 75 to 150 keV. Furthermore, the investigation examines the influence of bremsstrahlung radiation and evaluates fusion reactivity based on Hively and Bosch-Hale parameters. Hively’s reactivity value falls within the range of 1.121 × 10−22 – 6.750 × 10−22 m3/s, the instability occurs at temperatures < 100 keV. The Bosch-Hale value differs significantly from the Hively reactivity, which is 1.208 × 10−22 – 2.340 × 10−22 m3/s. The bremsstrahlung radiation within the range of 4.819 × 10−20 – 6.802 × 10−20 keV exhibits minimal impact on the reaction. The Lawson criterion value within the scope of this study typically falls within the range of 0.437 × 1021 – 1.452 × 1021 s/m3. This range signifies the necessary combination of confinement time and plasma density to facilitate the generation of clean energy within a temperature range of 75 to 150 keV, thus propelling the fusion process toward the ignition phase.

Reconstruction of poloidal magnetic field profiles in field-reversed configurations with machine learning in laser-driven ion-beam trace probe

The diagnostic of poloidal magnetic field ( ) in field-reversed configuration (FRC), promising for achieving efficient plasma confinement due to its high β, is a huge challenge because is small and reverses around the core region. The laser-driven ion-beam trace probe (LITP) has been proven to diagnose the profile in FRCs recently, whereas the existing iterative reconstruction approach cannot handle the measurement errors well. In this work, the machine learning approach, a fast-growing and powerful technology in automation and control, is applied to reconstruction in FRCs based on LITP principles and it has a better performance than the previous approach. The machine learning approach achieves a more accurate reconstruction of profile when 20% detector errors are considered, 15% fluctuation is introduced and the size of the detector is remarkably reduced. Therefore, machine learning could be a powerful support for LITP diagnosis of the magnetic field in magnetic confinement fusion devices.

Degradation of Stains from Metal Surfaces Using a DBD Plasma Microreactor.

The surface cleaning of metals plays a pivotal role in ensuring their overall performance and functionality. Dielectric barrier discharge (DBD) plasma, due to its unique properties, has been considered to be a good alternative to traditional cleaning methods. The confinement of DBD plasma in microreactors brings additional benefits, including excellent stability at high pressures, enhanced density of reactive species, reduced safety risks, and less gas and energy consumption. In the present work, we demonstrated a DBD plasma-based method for the degradation of stains from metal surfaces in a microreactor. Aluminum plates with capsanthin stains were used to investigate the influence of operational parameters on the decolorization efficiency, including plasma discharge power, plasma processing time, and O2 content in the atmosphere. The results revealed that an increase in plasma discharge power and plasma processing time together with an appropriate amount of O2 in the atmosphere promote the degradation of capsanthin stains. The optimum processing condition was determined to be the following: plasma power of 11.3 W, processing time of 3 min, and Ar/O2 flow rate of 48/2 sccm. The evolution of composition, morphology, bonding configuration, and wettability of aluminum plates with capsanthin and lycopene stains before and after plasma treatment were systematically investigated, indicating DBD plasma can efficiently degrade stains from the surface of metals without damage. On this basis, the DBD plasma cleaning approach was extended to degrade rhodamine B and malachite green stains from different metals, suggesting it has good versatility. Our work provides a simple, efficient, and solvent-free approach for the surface cleaning of metals.

Superconductivity for Nuclear Fusion: Past, Present, and Future

AbstractWithin fusion research and development, there are three main categories of fusion devices: magnetic confinement fusion, inertial confinement fusion, and magneto-inertial confinement. The focus on achieving power production has historically centered around magnetic confinement fusion, employing devices such as tokamaks, stellarators, and spheromaks. The plasma confinement in these machines relies on powerful magnetic fields generated from large, complex electromagnetic systems containing superconducting coils. Superconductivity, or the flow of current without resistance at low temperatures, allows the electromagnets to fulfill the demanding requirements of fusion devices. Analyzing the history of superconducting magnets in the application of fusion energy production provides necessary insight into the current state of the technology and allows for identification of current and future trends in research and development. Throughout its history, fusion research has experienced cyclic periods of depression followed by renewed interest. Breakthroughs in superconducting technologies have played a part in stimulating these periods of renaissance, cementing its role as an enabling technology for fusion. Future trends in research aim to address several challenges in using superconducting magnets in fusion devices, including manufacturing difficulties, irradiation and long-term availability, quench detection and protection, and finally the high cost of the materials and cryogenic cooling. The resolution of these issues is crucial for advancing fusion devices toward practical energy production.

Magnetic flutter effect on validated edge turbulence simulations

Small magnetic fluctuations ( B1/B0∼10−4 ) are intrinsically present in a magnetic confinement plasma due to turbulent currents. While the perpendicular transport of particles and heat is typically dominated by fluctuations of the electric field, the parallel stream of plasma is affected by fluttering magnetic field lines. In particular through electrons, this indirectly impacts the turbulence dynamics. Even in low beta conditions, we find that E × B turbulent transport can be reduced by more than a factor 2 when magnetic flutter is included in our validated edge turbulence simulations of L-mode ASDEX Upgrade. The primary reason for this is the stabilization of drift-Alfvén-waves, which reduces the phase shifts of density and temperature fluctuations with respect to potential fluctuations. This stabilization can be qualitatively explained by linear analytical theory and appreciably reinforced by the flutter nonlinearity. As a secondary effect, the steeper temperature gradients and thus higher η i increase the impact of the ion-temperature-gradient mode on overall turbulent transport. With increasing beta, the stabilizing effect on E × B turbulence increases, balancing the destabilization by induction, until direct electromagnetic perpendicular transport is triggered. We conclude that including flutter is crucial for predictive edge turbulence simulations.

Impact of divertor neutral pressure on confinement degradation of advanced tokamak scenarios at ASDEX Upgrade

Over previous campaigns, an intense experimental program on advanced tokamak (AT) scenarios, has been carried out at the ASDEX Upgrade tokamak with full-tungsten wall. These discharges have been executed shortly after the boronization of the first wall to reduce the density and the impurity influx. The confinement level of such AT discharges was found to vary considerably, even when discharges with similar, if not identical, engineering parameters were carried out. This work investigates the causes of such confinement variations. Among all plasma quantities analyzed, confinement quality of AT scenarios correlates best with divertor neutral pressure, highlighting the key role of edge and scrape-off layer physics in determining global plasma confinement. In particular, it is found that the main cause of confinement degradation is the reduction of pedestal stability, which is in turn caused by the outward shift of the maximum density gradient position typically observed when the divertor neutral pressure increases. Owing to the low density of AT discharges under analysis, the movement of the maximum density gradient position can be caused entirely by changes in deuterium outgassing from the wall, which is strongly influenced by the boron layer deposited on the plasma-facing components and by the deuterium wall inventory. Finally, the predictive capability of confinement quality with the integrated model IMEP [Luda et al., Nucl. Fusion 60, 036023 (2020)] is tested on these discharges and shows promising results.

Final results of the first phase of the PROTO-SPHERA experiment: obtainment of the full current stable screw pinch and first evidences of the jet + torus combined plasma configuration

In astrophysics, the boundary conditions for plasma phenomena are provided by nature and the astronomer faces the problem of understanding them from a variety of observations [Hester J J et al 1996 Astrophys. J. 456 225], on the other hand, in laboratory plasma experiments the electromagnetic boundary conditions become a major problem in the set-up of the machine that produces the plasma, an issue that has to be investigated step by step and to be modified and adapted with great patience, in particular in the case of an innovative plasma confinement experiment. The PROTO-SPHERA machine [Alladio F et al 2006 Nucl. Fusion 46 S613] is a magnetic confinement experiment, that emulates in the laboratory the jet + torus plasma configurations often observed in astrophysics: an inner magnetized jet of plasma centered on the (approximate) axis of symmetry and surrounded by a magnetized plasma torus orthogonal to this jet. The PROTO-SPHERA plasma is simply connected, i.e., no metal current conducting rod is linked to the plasma torus, while instead it is the inner magnetized plasma jet (in the following always called the plasma centerpost) that is linked to the torus. It is mandatory that no spurious plasma current path modifies the optimal shape of the plasma centerpost. Moreover, as the plasma torus is produced and sustained, in absence of any applied inductive electric field, by the inner plasma centerpost through magnetic reconnections [Taylor J B and Turner M F 1989 Nucl. Fusion 29 219], it is required as well that spurious current paths do not surround the torus on its outboard, in order not to lower the efficiency of the magnetic reconnections that maintain the plasma torus at the expense of the plasma centerpost. Boundary conditions have been corrected, up to the point that the first sustainment in steady state has been achieved for the combined plasma.