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Long-wavelength density fluctuations ( k ρ i 0.97, indicating the prevalence of density ‘voids’ at inner radii and density ‘blobs’ at outer radii and outside of the separatrix. The turbulence intensity flux ⟨ v ~ r n ~ 2 ⟩ is calculated to characterize turbulence spreading at the plasma edge. During ECH/NBI power ramps and at counter- I p injected torque, ⟨ v ~ r n ~ 2 ⟩ is directed inward inside the separatrix, which is evidence of inward spreading of turbulence intensity from the edge gradient region caused by the inner propagation of density ‘voids’. Significantly weaker ⟨ v ~ r n ~ 2 ⟩ is observed with co- I p torque. A correlation between co- I p torque, turbulence intensity δ n / n at ρ = 0.97, and increased srape-off layer (SOL) heat flux decay length λ q is found in the torque scan scenario, showing that edge turbulence plays a material role in determining the SOL conditions and heat flux width.

Since the first plasma realized in 2020, a series of key systems on HL-3 (known as HL-2M before) tokamak have been equipped/upgraded, including in-vessel components (the first wall, lower divertor, and toroidal cryogenic/water-cooling/baking/glow discharge systems, etc.), auxiliary heating system of 11 MW, and 28 diagnostic systems (to measure the plasma density electron temperature, radiation, magnetic field, etc.). Magnet field systems were commissioned firstly for divertor plasma discharges. During the 2nd experimental campaign of HL-3 tokamak, several great progresses have been achieved. Firstly, the successful operation with plasma current larger than 1 MA was achieved under a divertor configuration. Secondly, the advanced divertor concept with two distinct snowflake configurations was realized. It is found that the distribution of ion saturation current and heat flux on bottom plate becomes wide due to magnetic surface expansion, demonstrating the advantage of such configuration in the heat flux mitigation. In addition, using the combination of NBI, ECRH and LHCD, the standard sawtoothing high confinement mode of megampere plasma was firstly accessed on the HL-3. The successful commissioning of HL-3 is beneficial for the initial operation of ITER.

A real-time capable core Ion Cyclotron Range of Frequencies (ICRF) heating model on NSTX and WEST is developed. The model is based on two nonlinear regression algorithms, the random forest ensemble of decision trees and the multilayer perceptron neural network. The algorithms are trained on TORIC ICRF spectrum solver simulations of the expected flat-top operation scenarios in NSTX and WEST assuming Maxwellian plasmas. The surrogate models are shown to successfully capture the multi-species core ICRF power absorption predicted by the original model for the high harmonic fast wave and the ion cyclotron minority heating schemes while reducing the computational time by six orders of magnitude. Although these models can be expanded, the achieved regression scoring, computational efficiency and increased model robustness suggest these strategies can be implemented into integrated modeling frameworks for real-time control applications.

Experimental studies of the dynamics of shear flow and turbulence spreading at the edge of tokamak plasmas are reported. Scans of line-averaged density and plasma current are carried out while approaching the Greenwald density limit on the J-TEXT tokamak. In all scans, when the Greenwald fraction f G = n ¯ / n G = n ¯ / ( I p / π a 2 ) increases, a common feature of enhanced turbulence spreading and edge cooling is found. The result suggests that turbulence spreading is a good indicator of edge cooling, indeed better than turbulent particle transport is. The normalized turbulence spreading power increases significantly when the normalized E × B shearing rate decreases. This indicates that turbulence spreading becomes prominent when the shearing rate is weaker than the turbulence scattering rate. The asymmetry between positive/negative (blobs/holes) spreading events, turbulence spreading power and shear flow are discussed. These results elucidate the important effects of interaction between shear flow and turbulence spreading on plasma edge cooling.

The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q ⩾ 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant β phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent 'pedestal-to-divertor' integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T e and radiation which need control. High- β P scenarios with q min > 2, q 95–7.9, β N > 4, β T–3.3% and H 98y2 > 1.5 are sustained with high density ( n ¯ = 7E19 m−3, f G–1) for 6 τ E, improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors. [ABSTRACT FROM AUTHOR]

The heat flux width (λq) is a key parameter determining the heat load at divertor targets. In recent years, drifts have been found to play a remarkable role in the edge plasma transport, while its influence on λq has not been well understood. Investigations of the influence of drifts on λq, systematic simulations using the SOLPS-ITER code are performed in this work. The statistics of the simulation results show that the drift under favorable/unfavorable B T tends to increase the λq in the outer/inner side and decrease the λq in the other side, which is consistent with the experiment observations. At the upstream and the target, the mechanisms of the influence of the drifts on λq are different. The upstream λq (λq,u) is directly affected by the drift-induced convective heat flux, while λq at the target (λq,t) is indirectly influenced through heat conduction (in the high-recycling regime) and the sheath (in the detached regime) due to the change of plasma parameters there. Furthermore, the synergetic effect of geometry and drift under favorable B T leads to an anomalously large λq,t in the inner side at high density. [ABSTRACT FROM AUTHOR]

In this study, the effects of high-energy hydrogen ion irradiation, generated by the MTPF-2 type modern plasma focus device, on the surface and structural properties of graphite were investigated. High-energy protons produced by the plasma focus device were used to irradiate graphite samples in various shots. The radiation-induced changes on the graphite surface were examined using optical microscope and scanning electron microscopy. The modifications on the irradiated samples' surface were clearly observed in the scanning electron microscopyand optical microscope images. The microscopic images of the irradiated samples revealed point sputtering as well as pores on the sample surface, along with point melting, which are dependent on the ion fluence (number of shots). X-ray diffraction analysis was employed to study the structural changes in the graphite caused by the high-energy proton irradiation. The X-ray diffraction spectra of the irradiated samples showed shifts in the peak positions, changes in the peak intensities, and an increase in the crystallite size compared to the X-ray diffractionspectra of the reference sample. Additionally, a Faraday cup detector was used to characterize the hydrogen ion spectrum produced in the plasma focus device. The results indicated that the average energy of the generated ions was ~ 46 keV. [ABSTRACT FROM AUTHOR]

Tokamaks are a very promising option to exploit nuclear fusion as a programmable and safe energy source. A very critical issue for the practical use of tokamaks consists of the power flow required to initiate and sustain the fusion process, in particular in the poloidal field coils. This flow can be managed by introducing a DC energy storage based on supercapacitors. Because such storage may be the most expensive and largest part of the poloidal power supply system, an excessive size would cancel its potential advantages. This paper presents innovative strategies to optimize the DC storage in poloidal power supply systems. The proposed solution involves the sharing of the DC storage between different coil circuits. The study is supported by novel analytical formulas and by a circuital model developed for this application. The obtained results show that this method and the related algorithms can noticeably reduce the overall size of the storage and the power exchange with the grid, providing a practical contribution toward the feasibility and the effectiveness of nuclear fusion systems. [ABSTRACT FROM AUTHOR]

Introduction: This work explores the use of eXplainable artificial intelligence (XAI) to analyze a convolutional neural network (CNN) trained for disruption prediction in tokamak devices and fed with inputs composed of different physical quantities. Methods: This work focuses on a reduced dataset containing disruptions that follow patterns which are distinguishable based on their impact on the electron temperature profile. Our objective is to demonstrate that the CNN, without explicit training for these specific mechanisms, has implicitly learned to differentiate between these two disruption paths. With this purpose, two XAI algorithms have been implemented: occlusion and saliency maps. Results: The main outcome of this paper comes from the temperature profile analysis, which evaluates whether the CNN prioritizes the outer and inner regions. Discussion: The result of this investigation reveals a consistent shift in the CNN's output sensitivity depending on whether the inner or outer part of the temperature profile is perturbed, reflecting the underlying physical phenomena occurring in the plasma. [ABSTRACT FROM AUTHOR]

Predictive design modeling of a Dissipation-Focused Divertor for future operation in DIII-D reveals that increasing the poloidal distance of the pump duct entrance from the target surface along the low-field side divertor baffle increases neutral compression and modifies the spatial distribution of power dissipation. With a divertor pump located mid-leg between the target and the X-point, SOLPS-ITER boundary plasma simulations without drifts predict the formation of a dense neutral cloud near the target with > 30x higher neutral compression in detachment, a more stable detachment front located further from the target, and ∼25% lower outer midplane separatrix density required for detachment onset, compared to a pump located in the scrape-off layer at the target surface. Up to 19 MW of power flowing into the divertors is modeled using the following two numerical implementations for particle pumping: a specified fraction of particles incident on variable wall sections of the plasma grid is removed from the computational domain (so-called albedo pumping), and a pump duct is modeled which includes dynamics of kinetic neutrals in the duct. The simulations show that the detachment front is located between the divertor target and the X-point and is relatively stable near the pump entrance, without a strong dependence on gas puff rate or injected power. The mid-leg pump design spatially separates the two primary functions of a divertor (power handling and particle exhaust), with the majority of power dissipation occurring near the target plate and particle exhaust taking place further upstream. The benefit of enhanced dissipation using mid-leg pumping comes at the cost of a higher outer midplane separatrix density for a given amount of particle injection.

The technique based on quadrupole mass spectrometry of the material ejected by laser-induced ablation – LIA-QMS is proposed as a tool to measure H/D/T content in carbon co-deposits in the first wall and divertor regions of tokamak reactors. Tungsten tiles exposed in Globus-M2 tokamak were taken for validation experiments. Measurement accuracy was determined by comparison with the results obtained by laser-induced desorption – LID-QMS and conventional thermal desorption spectroscopy (TDS). Total amount of deuterium in hydrocarbon deposits was found to be 2.8 × 1017 D/cm2, when measured by – LIA-QMS, 3.0 × 1017 D/cm2 for LID-QMS and 2.9 × 1017 D/cm2 for TDS. The main uncertainty of D surface concentration are caused by inhomogeneity of the deposit thickness over the sample area or by presence of hydrocarbon species in the released gas.

Fusion power plants will require detachment to mitigate sputtering and keep divertor heat fluxes at tolerable levels. Controlling detachment on these devices may require the use of real-time scrape-off-layer modeling to complement the limited set of available diagnostics. In this work, we use the configurable Hermes-3 edge modeling framework to perform time-dependent, fixed-fraction-impurity 1D detachment simulations. Although currently far from real-time, these simulations are used to investigate time-dependent effects and the minimum physics set required for control-relevant modeling. We show that these simulations reproduce the expected rollover of the target ion flux — a typical characteristic of detachment onset. We also perform scans of the input heat flux and impurity concentration and show that the steady-state results closely match the scalings predicted by the 0D time-independent Lengyel–Goedheer model. This allows us to indirectly compare to SOLPS simulations, which find a similar scaling but a lower value for the impurity concentration required for detachment for given upstream conditions. We use this result to suggest a series of improvements for the Hermes simulations, and finally show simulations demonstrating the impact of time-dependence.

Plasma-wall interaction (PWI) is a critical concern in tokamaks because of its significant impact on the lifetimes of plasma-facing materials (PFMs), fuel retention, and plasma performance. In 2020, the Experimental Advanced Superconducting Tokamak (EAST) PFMs were upgraded to predominantly metallic walls. Following the 2021 experimental campaign, the deposition distribution and fuel retention on the surfaces of the PFMs along both the poloidal and toroidal directions were analyzed. The poloidal tests commenced at the high-field side (HFS), proceeded to the lower divertor, then to the low-field side (LFS), and finally to the upper divertor. The toroidal tests were performed at the midplane of the HFS, beginning from port A and ending at port P. The distributions of the deposits in the poloidal and toroidal directions were clearly asymmetrical. Mo and Fe particles sputtered from the first wall and inner stainless steel (SS) components were prone to deposition in the lower divertor region, as evidenced by the fact that the elemental content in the far-SOL region of the inner divertor exhibited Mo and Fe peaks, as did both the near- and far-SOL regions of the outer divertor. In addition, quick re-deposition of W and Fe was observed, as demonstrated by the fact that their contents near the erosion sources were higher than those farther away along the toroidal first wall on the HFS. This tendency was stronger for W than for Fe. Further analysis indicated that the deposits consisted of Li, C, O, W, Mo, Fe, Cu, and Ni. The Li and Fe contents were much higher than those of other metal impurities, with peak values of 8.17 μg/mm2 and 7.78 μg/mm2, respectively. The Li content decreased along the HFS first wall from 8.17 μg/mm2 at P-2 to 1.82 μg/mm2 at P-22, and then increased to 2.72 μg/mm2 at P-32 in the divertor, while the Fe content was higher around the top side and the midplane of the HFS. The deposited Li originated from routine wall conditioning and existed primarily in the form of Li2CO3. Additionally, other metal impurities deposited on the Li2CO3 surfaces exhibited various irregular shapes, often appearing as aggregated and recrystallized small particles. Furthermore, significant deuterium (D) retention on the order of 1021 atoms/m2 was measured on all the analyzed SiC-coated graphite tiles on the HFS. The D content at location P-10 at the midplane was the highest along the poloidal direction of the HFS because it was directly facing the NBI beam.

Real-time monitoring and control of thermal loads on tungsten plasma-facing components is mandatory for ITER to avoid their overheating during plasma discharges. For the Start of Research Operations (SRO) phase, dedicated First Wall Heat Load Control (FWHLC) functions are included in the ITER Plasma Control System (PCS) to prevent the development of excessive plasma heat loads on the inertially cooled first wall (FW). In this work, we propose a FWHLC design with a model-based approach, which features a simplified thermal model for the FW armour. This control-oriented model simulates the thermal response of ITER FW to slow changes in plasma equilibrium and energy, which during ITER operation will be monitored by the real-time wide angle infrared camera system. This allows computationally efficient runs for rapid controller performance assessment but can also assist operation scenario design by pre-empting potential heat load issues. FWHLC is designed to react to measured and predicted FW temperatures overcoming predesigned limits with real-time variations of the plasma shape or auxiliary power references, aiming to reduce thermal loads before a premature plasma termination is required to protect the FW. The new scheme is tested in closed loop simulations, where perturbations to nominal ITER low current scenarios are introduced in the wall thermal model to trigger the response of FWHLC, supporting the effectiveness of the proposed policy.

The estimation of the heat flux density distribution profiles in tokamak devices is a very important research topic for edge plasma physics purposes and also to ensure the safety of the machine. In the radial direction, the heat flux exhibits an exponential decay that could be captured by thermal sensors distributed in the plasma facing components. Radially distributed thermal sensors based on Fiber Bragg grating technology have been embedded in the WEST lower divertor to study the heat flux deposition profiles during plasma operation. The comparison between embedded measurements and a 3D finite element model shows a small decay length (5 – 10 mm) on top of a wider heat flux with a decay length around 30 to 50 mm. A tool using neural network has been developed in order to predict the values of the different parameters describing the deposited heat flux from embedded temperature measurements in steady state. A large span of deposited heat fluxes with maximum heat flux ranging from 1 to 9 MW/m2 and decay length from 5 to 50 mm were characterized using this tool over a database of more than 250 experimental L-mode pulses performed in WEST in attached divertor configuration. The comparison of the predicted heat flux parameters values with macroscopic plasma parameters have revealed the appearance of the narrow component with the increase of the divertor power load (Pdiv) with a threshold dependant of the plasma current (IP).

Global power balance calculations in steady state H mode plasmas varying the distance between separatrices (drsep) and the divertor configuration have been performed in MAST Upgrade. As drsep becomes more negative, more of the power crossing the separatrix goes to the lower divertor. The inner divertor receives a higher fraction of the power exhaust in a Super-X divertor plasma compared to a conventional divertor plasma at similar negative drsep, which is a concern for high power devices employing alternative divertor configurations for power exhaust handling. Global power accounting suggests >30% of the input power is unaccounted for with the power loss channels quantified in this work. Charge exchange and orbit losses from the NBI could account for a large fraction of unaccounted power but it is not possible to precisely determine this without further diagnostic calibration.

The recent Federici et al (2024 Nucl. Fusion 64 036025) article makes the argument that higher magnetic fields cannot reduce the size and cost of a tokamak-based fusion power plant due to: structural considerations of the toroidal field (TF) coils, the required thickness for neutron shielding and the blanket, and challenges with heat exhaust in the divertor. This conclusion is based on a series of assumptions that are design decisions made by the authors, not fundamental limits on physics or engineering. This Comment demonstrates that the conclusions of Federici et al are invalid if one makes different design choices and that its results are therefore not broadly generalizable. [ABSTRACT FROM AUTHOR]

[Introduction] The superconducting tokamak serves as the foundation for steady-state operation. Timely and reliable quench detection is the key to ensuring the secure operation of superconducting magnets, which has the highest level of safety in fusion device operation. [Method] In this paper, an overview of the principles and implementation approaches of quench voltage detection for fusion magnets was provided. It briefly summarized the research progress of quench detection and the quench detection schemes adopted by mainstream tokamak devices and introduced the basic principles and processes of quench voltage detection. Taking the experimental advanced superconducting tokamak (EAST) as an example, this paper analyzed the interference sources and coupling mechanisms of quench voltage detection, proposed a two-stage decoupling compensation scheme, and established an optimization analysis model for primary compensation and a dynamic compensation mechanism for plasma-coupling interferences. [Result] The experimental results indicate that the compensation scheme can achieve a high noise suppression ratio greater than 99.9%, which can effectively improve the signal-to-noise ratio (SNR) and reliability of quench detection. [Conclusion] The established optimization analysis model for primary compensation and the dynamic compensation mechanism for plasma-coupling interferences can be extended to other tokamak devices, thereby providing necessary accumulation for the design of quench detection and the secure operation of magnets in future fusion reactors.

[Introduction] Fusion energy has the characteristics of large energy released through reactions, safe and reliable operation, abundant fuel sources, and low environmental pollution. Therefore, it is expected to become a commercial energy source that can be supplied in the market on a large scale, providing stable energy output and power supply in the future. To popularize the development path of magnetic confined fusion energy in China, the article reviews the discovery and realization path of fusion energy. [Method] The article provided an overview of the early research and development process of magnetic confined fusion energy in China by the method of literature review. The paper took the development of magnetic confined fusion energy as an example to provide a preliminary overview of the construction of typical research devices in China, such as tokamak, stellarator, spherical tokamak, reversed field pinch, magnetic mirror field, linear plasma, and dipole magnetic field devices. [Result] Based on the construction and researches of these devices, China has cultivated a group of scientific and technological talents in the field of magnetic confined fusion research, achieving significant progress. In addition, the article provides an overview of international cooperation in fusion energy research, as well as the International Thermonuclear Experimental Reactor (ITER) project that China is participating in construction. [Conclusion] Although the current research on fusion energy still needs to overcome significant challenges from various aspects, such as fusion plasma physics, fusion reactor materials, and tritium self-sustaining technology, it is believed that the magnetic confined fusion energy in China will turn from a blueprint into a reality in the future thanks to the urgent need for energy structure transformation and the strong support for fusion researches in China.

[Introduction] Plasma disruption poses a significant threat to the tokamak nuclear device during its running and can cause damage to the device. Such damage can be reduced by adopting the disruption mitigation system, which has an action time highly dependent on the real-time running plasma disruption prediction system for predicting the plasma disruption moment. The deep-learning-based neural network has been used to train plasma disruption prediction models, and the real-time running of the deep-learning-based disruption prediction models requires a huge amount of real-time data from multiple diagnostics. [Method] The article proposed a design scheme for a real-time data acquisition system. The real-time data acquisition and transmission system was designed based on the modular structure and divided into the multiple channels acquisition module, ADC converting control and data reading module, data grouping and packing module and data transmission network module. The data transmission network module was developed on the hardware UDP network stack running on the FPGA at a speed of 10 G. This hardware UDP network stack featured a deterministic data transmission process, enabling a very low transmission latency of the system. [Result] The real-time data acquisition system has a sampling rate reaching 2 MSa/s, a data throughput rate exceeding 9.3 Gb/s, and a data transmission latency of less than 10 μs. [Conclusion] This data acquisition system facilitates the fast transmission of diagnostic data streams to disruption prediction models. The high sampling rates enable the system to perform real-time transmission of one-dimensional diagnostics such as radiation and electron temperature, improving the temporal resolution of data. The high data throughput rate can increase the transmission volume of diagnostic data, and the low data transmission latency can reduce the time required for disruption prediction models to obtain diagnostics data.

[Introduction] In recent years, as the installed capacity of the fusion device has been increasing, the power supply system of the device requires high-amplitude and long-pulse power output, which causes a strong power impact on the grid, and even power supply disconnection in severe cases to pose a serious threat to the safety of the fusion device. Additionally, the high-amplitude pulses also result in a lot of redundancy in the power supply system design. [Method] To solve these problems, this paper proposed a novel fusion power supply topology with energy storage, that is, the power supply system was designed with energy storage to mitigate the impact of pulse power on the grid. During the operation of the power supply, the grid provided steady-state power while the energy storage device delivered pulse power, effectively reducing the cost of the power supply. [Result] According to the experience in industrial power supply and distribution, this paper presents the calculation formulas for the selection of key components in different topologies and conducts a cost analysis and comparison for different power supply topologies based on simulation data. [Conclusion] The experimental results show that the new power supply topology not only reduces the power impact on the grid and improves the energy utilization of the power supply system, but also reduces transformer capacity, thus it is cost-efficient.

Plasma disruption in tokamak experiments is a challenging issue that causes damage to the device. Reliable prediction methods are needed, but the lack of full understanding of plasma disruption limits the effectiveness of physics-driven methods. Data-driven methods based on supervised learning are commonly used, and they rely on labelled training data. However, manual labelling of disruption precursors is a time-consuming and challenging task, as some precursors are difficult to accurately identify. The mainstream labelling methods assume that the precursor onset occurs at a fixed time before disruption, which leads to mislabeled samples and suboptimal prediction performance. In this paper, we present disruption prediction methods based on anomaly detection to address these issues, demonstrating good prediction performance on J-TEXT and EAST. By evaluating precursor onset times using different anomaly detection algorithms, it is found that labelling methods can be improved since the onset times of different shots are not necessarily the same. The study optimizes precursor labelling using the onset times inferred by the anomaly detection predictor and test the optimized labels on supervised learning disruption predictors. The results on J-TEXT and EAST show that the models trained on the optimized labels outperform those trained on fixed onset time labels.

Disruption-generated runaway electron (RE) beams represent a potentially severe threat for tokamak plasma-facing components. Application of properly designed 3D fields can act as a mitigation mechanism, as recently investigated in ASDEX Upgrade (AUG) and COMPASS experiments and in the tokamak discharges of RFX-mod. In all of these devices, the dynamics of the disruption are affected by the application of magnetic perturbations (MPs), and the resulting RE beam current and lifetime are significantly reduced. These experiments show, in particular, that the strength of the observed effects strongly depends on the poloidal spectrum of the applied MPs, which has been reconstructed including the plasma response. This paper reports the main findings on RE mitigation from the previously mentioned three devices, highlighting the common physics behind them and their interpretation by using the guiding center code ORBIT. [ABSTRACT FROM AUTHOR]

Extended-MHD modeling of planned Runaway Electron Mitigation Coils (REMC) for SPARC and DIII-D is performed with the NIMROD code. A coil has been designed for each machine, with the two differing in shape and location, but both having n = 1 symmetry (with n the toroidal mode number). Compared to previous modeling efforts, three improvements are made to the simulations boundary conditions. First a resistive wall model is used in place of an ideal wall. Second, the ThinCurr code is used to compute the time-dependent 3D fields used as magnetic boundary conditions for the simulations. Third, the simulation boundary is moved from the first-wall location to the Vacuum Vessel (VV), which extends the boundary past the location of the internal REMC. To remove the 3D coil from the simulation domain, an equivalent set of 3D fields is calculated at the VV boundary that produce approximately the same field distribution at the last closed flux surface assuming vacuum between the two. Each of these three boundary condition improvements leads to an improvement in the predicted performance of the REMC for both machines. The resistive wall alone primarily effects the resonance of the coil with the plasma after the TQ, affecting the q-profile evolution in the SPARC modeling, and allowing the applied spectrum to be modified in response to the plasma in the DIII-D modeling. The movement of the simulation boundary has the most significant effect on the RE confinement overall, including in the early stages, particularly for a DIII-D inner wall limited equilibrium, where the RE loss fraction increases from 90% to > 99%, with SPARC RE losses also occurring much earlier when the boundary is placed at the VV. [ABSTRACT FROM AUTHOR]

In fusion, the validation of turbulent transport models is undertaken with the goals of making basic physics discoveries as well as for development of new predictive models to improve the operation and enhance the performance of existing and future fusion reactors. A fusion industry is just beginning to emerge globally. Like fission, validation in fusion energy research is a vibrant research area, but unlike fusion, a fission industry exists. The fission power industry motivates validation efforts, often performed at universities with small-scale experiments and advanced models and simulations developed in-house. Because fission research spans basic physics and applications, and addresses near-term and long-term industry interests, validation is thriving. This perspective article describes the validation of turbulent transport models in both fusion research and fission research, draws parallels between the validation methods and techniques used in two areas of the fields, and presents an outlook for thriving university fusion and fission research programs underpinned by a virtual cycle of basic and applied research that supports industry needs as well as tackling intellectual grand challenges. [ABSTRACT FROM AUTHOR]

Tokamak is an important device for controlled nuclear fusion research. The plasma electronic density control system (PEDCS) is an important system for controlling the Tokamak discharge process, which should be of high stability, rapidity, and accuracy. Gas seeding systems are widely used in many Tokamak devices to achieve plasma electronic density control. According to the mechanism model analysis for the plasma electronic density object, an adapted single neuron proportion integration differentiation (PID) control algorithm with the radial basis function (RBF) neural network tuning is studied. The principle and the implementation of the intelligent control algorithm are described in detail in this paper. The intelligent controller enables the system to optimize the PID parameters online according to the density state in the discharge process. The experimental results show that the adapted algorithm achieves a good control effect and also improves the control performance. The proposed method provides a useful reference for Tokamak devices and other similar control systems. [ABSTRACT FROM AUTHOR]

Copyright of Southern Energy Construction is the property of Southern Energy Construction Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Southern Energy Construction is the property of Southern Energy Construction Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Southern Energy Construction is the property of Southern Energy Construction Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Southern Energy Construction is the property of Southern Energy Construction Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

This paper describes a new practical numerical model for the calculation of lateral electromagnetic (EM) loads in tokamaks during asymmetric vertical displacement events (AVDEs). The model combines key features of two recently reported trial models while avoiding their drawbacks. Their common basic feature is the superposition of two patterns of halo current: one perfectly symmetric and another perfectly anti-symmetric. This model combines the following features that have not been combined before (a) a helically distorted halo layer wrapping around core plasma, and (b) halo-to-wall interception belts slipping along plasma-facing walls. This combination almost doubles the lateral net forces. An AVDE creates significant lateral net moments. Being relatively modest at VDEs, the lateral moments become a dominant component of EM loads at AVDEs. The model carefully tracks the balance of net EM loads (zero total for the tokamak), as a necessary condition for the consequent numerical simulation of the tokamak’s dynamic response. This balance is needed as well for the development of tokamak monitoring algorithms and simulators. In order to decouple from the current uncertainties in the interpretation and simulation of AVDE physics, the model does not simulate AVDE evolution but uses it as an input assumption based on the existing interpretation and simulation of AVDE physics. This means the model is to be used in a manner of parametric study, at widely varied input assumptions on AVDE evolution and severity. Parametric results will fill a library of ready-for-use waveforms of asymmetric EM loads (distributed and total) at tokamak structures and coils, so that the physics community may point to specific variants for subsequent engineering analysis. This article presents the first practical contribution to this AVDE library.

Tokamaks are a very promising option to exploit nuclear fusion as a programmable and safe energy source. A very critical issue for the practical use of tokamaks consists of the power flow required to initiate and sustain the fusion process, in particular in the poloidal field coils. This flow can be managed by introducing a DC energy storage based on supercapacitors. Because such storage may be the most expensive and largest part of the poloidal power supply system, an excessive size would cancel its potential advantages. This paper presents innovative strategies to optimize the DC storage in poloidal power supply systems. The proposed solution involves the sharing of the DC storage between different coil circuits. The study is supported by novel analytical formulas and by a circuital model developed for this application. The obtained results show that this method and the related algorithms can noticeably reduce the overall size of the storage and the power exchange with the grid, providing a practical contribution toward the feasibility and the effectiveness of nuclear fusion systems.

The first results of investigation of edge localized modes (ELMs) in the Globus-M2 tokamak using the Doppler backscattering method are presented in this paper. Specifically, ELMs that are initiated by sawtooth crashes in the H-mode are discussed. The goal of this paper is study plasma turbulence behaviour during ELMs and to showcase what ELM characteristics can be obtained using Doppler backscattering (DBS). An increase of the poloidal rotation velocity during an ELM burst and a decrease in the inter-ELM periods was observed. The effect of ELMs on the plasma turbulence was investigated and estimated to span around 6 cm inside the separatrix. This is to do with the fact that the sawtooth crashes which are responsible for initiating the ELMs take place in the core plasma. Additional experiments with standard reflectometry indicate that ELMs develop 3 cm inside the separatrix where the pedestal region is believed to be in Globus-M2. The direction of the expansion of the ELMs from the inner plasma region to the edge was determined and the velocity was estimated to be around 8 km s−1. During a single ELM burst a series of filament structures were found in the peripheral DBS channels. In an attempt to understand the processes involved modelling of the reaction of the DBS signals to filaments was done using the BOUT ++ and IPF-FD3D full-wave codes, and the cases for both linear and nonlinear scattering were considered. The results show that the presence of nonlinear scattering during ELMs can lead to an overestimation of the measured velocity values in the region of filament existence near the separatrix. [ABSTRACT FROM AUTHOR]

An analytical formula systematically predicts the observed frequency of pressure relaxation limit cycle oscillations in the vicinity of the transition to high confinement in four tokamaks (JET, ASDEX Upgrade, COMPASS, Globus-M). The experimental dataset spans the widest available range of frequencies, machine sizes and plasma ion species. The machine size dependence is explained by the connection length scale of plasma flows parallel to the magnetic field. The model also explains the observed up-down poloidal current asymmetry and the impact of the plasma ion species mass and charge. [ABSTRACT FROM AUTHOR]

This paper reviews the recent advancements of the algorithm and application to fusion plasma instability study of the fully kinetic Particle-in-Cell (PIC) method. The strengths and limitations of both explicit and implicit PIC methods are described and compared. Additionally, the semi-implicit PIC method and the code ECsim used in our research are introduced. Furthermore, the application of PIC methods in fusion plasma instabilities is delved into. A detailed account of the recent progress achieved in the realm of tokamak plasma simulation through fully kinetic PIC simulations is also provided. Finally the prospective future development and application of PIC methods are discussed as well. [ABSTRACT FROM AUTHOR]

To achieve optimal operation of the polarimetry-based FOCS, the light polarisation state at the input of the sensing fibre part must be close to a linear one. In the case of a FOCS deployed on a tokamak, the Joint European Torus (JET) in the present work, the long fibre optics link between the laser source and the sensing fibre modifies the polarisation in an unpredictable way, making it unclear which source polarisation state is to be set. A method for performing the necessary polarisation adjustment in a systematic way is proposed based on the FOCS analysis. The method requires performing data acquisition at two different input polarisations. Based on these measurements, the optimal laser source polarisation can be found. The method was experimentally verified using laboratory set-up and then successfully demonstrated with the FOCS installed at JET. [ABSTRACT FROM AUTHOR]

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.

Langmuir probe measurements in neutral beam injection (NBI) heated, low-confinement mode plasmas in JET ITER-like wall showed that the current to the divertor targets, Idiv, in helium (He) plasmas was up to 70% lower on the low-field side (LFS) than in otherwise identical deuterium (D) plasmas. The edge plasma density at which the rollover of Idiv occurred i.e. the onset of detachment, was 10% higher in He plasmas on both the LFS and high-field side (HFS). The density of Idiv rollover increases by 25% for He when the NBI power increases 1MW to 5MW. The total radiated power was similar in He and D plasmas for densities below the Idiv rollover. At densities above the Idiv rollover density, the total radiated power and power from within the separatrix are higher in He, reducing the power across the separatrix and subsequently Idiv,LFS. In He plasmas, the peak radiated power was observed within the confined region above the X-point in tomographic reconstructions from bolometry.

Disruption-generated runaway electron (RE) beams represent a potentially severe threat for tokamak plasma-facing components. Application of properly designed 3D fields can act as a mitigation mechanism, as recently investigated in ASDEX Upgrade (AUG) and COMPASS experiments and in the tokamak discharges of RFX-mod. In all of these devices, the dynamics of the disruption are affected by the application of magnetic perturbations (MPs), and the resulting RE beam current and lifetime are significantly reduced. These experiments show, in particular, that the strength of the observed effects strongly depends on the poloidal spectrum of the applied MPs, which has been reconstructed including the plasma response. This paper reports the main findings on RE mitigation from the previously mentioned three devices, highlighting the common physics behind them and their interpretation by using the guiding center code ORBIT.

Introduction: This work explores the use of eXplainable artificial intelligence (XAI) to analyze a convolutional neural network (CNN) trained for disruption prediction in tokamak devices and fed with inputs composed of different physical quantities.Methods: This work focuses on a reduced dataset containing disruptions that follow patterns which are distinguishable based on their impact on the electron temperature profile. Our objective is to demonstrate that the CNN, without explicit training for these specific mechanisms, has implicitly learned to differentiate between these two disruption paths. With this purpose, two XAI algorithms have been implemented: occlusion and saliency maps.Results: The main outcome of this paper comes from the temperature profile analysis, which evaluates whether the CNN prioritizes the outer and inner regions.Discussion: The result of this investigation reveals a consistent shift in the CNN’s output sensitivity depending on whether the inner or outer part of the temperature profile is perturbed, reflecting the underlying physical phenomena occurring in the plasma.

In fusion, the validation of turbulent transport models is undertaken with the goals of making basic physics discoveries as well as for development of new predictive models to improve the operation and enhance the performance of existing and future fusion reactors. A fusion industry is just beginning to emerge globally. Like fission, validation in fusion energy research is a vibrant research area, but unlike fusion, a fission industry exists. The fission power industry motivates validation efforts, often performed at universities with small-scale experiments and advanced models and simulations developed in-house. Because fission research spans basic physics and applications, and addresses near-term and long-term industry interests, validation is thriving. This perspective article describes the validation of turbulent transport models in both fusion research and fission research, draws parallels between the validation methods and techniques used in two areas of the fields, and presents an outlook for thriving university fusion and fission research programs underpinned by a virtual cycle of basic and applied research that supports industry needs as well as tackling intellectual grand challenges.

In this paper the dynamics of high energy charged particles in negative and positive triangularity tokamaks is studied through the simulation of a neutral beam injection on analytical MHD equilibria using the exact orbit code FOCUS. It is observed that the negative triangularity configuration presents fewer losses of non-thermal particles, and an energy deposition more localized than its counterpart.

Abstract Enhancements of edge zonal flows, radial electric fields, and turbulence are observed in electron cyclotron resonance heating-heated plasmas (Zhao et al 2013 Nucl. Fusion 53 083011). In this paper, the effects of sawtooth heat pulses on flows and turbulence are presented. These experiments are performed using multiple Langmuir probe arrays in the edge plasmas of the HL-2A tokamak. The edge zonal flows, radial electric fields, and turbulence are all enhanced by sawteeth. Propagation of the zonal flow and turbulence intensities is also observed. The delay time of the maximal intensity of the electric fields, zonal flows, and turbulence with respect to the sawtooth crashes is estimated as ∼1 ms and comparable to that of the sawtooth-triggered intermediate phases. Not only the zonal flows but also the radial electric fields lag behind the turbulence. Furthermore, the intensities of both the zonal flows and electric fields nearly linearly increase/decrease with the increase/decrease of the turbulence intensity. A double-source predator–prey model analysis suggests that a relatively strong turbulence source may contribute to the dominant zonal flow formation during sawtooth cycles.

BackgroundTokamak plasma disruption generates a runaway current carrying enormous amounts of energy that, if not suppressed, can cause severe damage to equipment.PurposeThis study aims to investigate the effects of injecting a deuterium-argon/neon gas mixture on a runaway current during plasma disruption.MethodsBased on the high plasma current discharge conditions of the HL-2M tokamak device in China, numerical simulations were conducted using a fluid model in the DREAM code. Variations of plasma parameters, such as plasma current (Ip), ohmic current (Iohm), runaway current and the ohmic electric field, with the injected deuterium-argon content and ratio during the disruption process were consistently simulated.ResultsResults show that injecting a deuterium-argon/neon gas mixture suppresses the eventual formation of a platform runaway current, and an optimal content and ratio of the deuterium-argon/neon gas mixture are existed for effective runaway current suppression. Within the range of the pre-disruption plasma current (Ip) discussed in this study, the amounts of neon/argon and deuterium in the gas mixture should be 0.50%~0.70% and 1020~1021 m-3, On fusion-reactor-scale tokamak devices with Ip as high as 10 MA, the amount of the injected gas mixture must reach 1022 m-3, which cannot be achieved under the current massive gas injection (MGI) technique.ConclusionsThe pre-disruption plasma current (Ip) is the key factor that influences a runaway current. The larger Ip is, the larger is the runaway current that is formed and more amount of the gas mixture must be injected. On fusion-reactor-scale tokamak devices with Ip as high as 10 MA, the amount of the injected gas mixture must reach 1022 m-3, which cannot be achieved under the current massive gas injection technique. Injecting a deuterium-argon/neon gas mixture through a shattered pellet would be a more viable approach.

One of the essential systems for the upgrade of Damavand Tokamak is the automatic control of the vacuum vessel pressure profile. This research performs the identification, modeling, and control of the vacuum chamber pressure in Damavand Tokamak. As a first step, the experimental structure is designed and implemented for the application of an identification signal and for the collection of data from the vacuum vessel pressure. Following the creation of the base vacuum, the input voltage to the piezoelectric is changed and applied. At the same time, both the vacuum vessel pressure and applied voltage are measured. Using experimental data, several linear models are identified for control. Then, considering the practical limitations of the control signal (25

This paper reviews the recent advancements of the algorithm and application to fusion plasma instability study of the fully kinetic Particle-in-Cell (PIC) method. The strengths and limitations of both explicit and implicit PIC methods are described and compared. Additionally, the semi-implicit PIC method and the code ECsim used in our research are introduced. Furthermore, the application of PIC methods in fusion plasma instabilities is delved into. A detailed account of the recent progress achieved in the realm of tokamak plasma simulation through fully kinetic PIC simulations is also provided. Finally the prospective future development and application of PIC methods are discussed as well.

The JT-60SA device offers unique conditions before ITER for the study of the interaction of energetic particles with plasma waves. With similar dimensions to JET, e.g., a major radius but with a slightly more elongated plasma volume, JT- 60SA is used as a high-power device where additional heating power (including 10 MW of the 500 keV Neutral Beam Injection) of up to 41 MW and the potential for high non-inductive plasma current operation pave the path for numerous challenges in physics on MHD stability, in particular, when considering the effects of energetic particles. Several operational scenarios with ITER and DEMO-relevant plasma regimes, in terms of non-dimensional plasma parameters, are anticipated. In this work, the stability of Alfvén eigenmodes (AEs) in variants of two of the most relevant operational scenarios with single null is analyzed: a full Ip inductive scenario at high density (1.1 × 1020 m-3 on-axis electron density) and 5.48MA/2.05T toroidal plasma current and magnetic field, and an advanced (hybrid) scenario with an ion energy transport barrier (ITB) and 3.5MA/2.28T toroidal plasma current and magnetic field. The workflow included the CRONOS code to establish the scenario, the ASCOT code to calculate the slowing-down energetic particle distributions for a positive/negative ion source-based neutral beam, and the MISHKA/CASTOR-K suite to calculate the MHD spectra of AEs and the associated drive/damping contributions from the NBI energetic ions, as well as the thermal ion landau damping. The systematic analysis, over a large Fourier space of the toroidal mode number/mode frequency, provides evidence that although a significant fraction of supra-Alfvénic particles stemming from the negative ion source-based neutral beam (500 keV) can, in some cases, drive to AEs in both scenarios, it is not enough to overcome the thermal ion landau damping. In addition, the advanced scenario with ITB is shown to be stable against AEs localized in the vicinity of the barrier as well, offering good prospects of sustainability of the plasma performance and of ITB. Finally, some sensitivity scan results are shown on the influence of fast ion density and q-profile on the AE mode spectra and stability. [ABSTRACT FROM AUTHOR]

The problem considered includes the development and modeling of an adaptive control algorithm for unstable vertical plasma positioning in a vertically elongated tokamak. At each iteration, a new PID controller is automatically synthesized for the evolving plasma model identified using the least squares method. The parameters of the feedback controller were computed based on the desired placement of the poles of the closed-loop control system in the left half-plane of the complex plane. The initial control system model utilized was a robust system synthesized using Quantitative Feedback Theory (QFT). The system was simulated on a real-time digital test bed (https://www.ipu.ru/plasma/about). [ABSTRACT FROM AUTHOR]

Divertor power load is a major challenge towards a burning plasma in a next-step tokamak. Here, the first results of a divertor power load characterisation in tritium plasmas in type-I ELMy H-mode, obtained in the JET deuterium-tritium campaign (DTE2) performed in 2021, are presented. It is demonstrated that both, transient loads due to type-I ELMs as well as the power fall-off length, do not exhibit an explicit ion mass dependence, with remarkably similar values in the tritium plasmas and in the deuterium references. This gives an improved credence to published scaling law predictions, solely based on deuterium plasma experiments. Moreover, the type-I ELM impact on the inner divertor target is studied in deuterium discharges. A slightly increased parallel energy fluence on the inner target with a factor of 1.08 compared to the outer target is observed. This is explained by the smaller major radius of the inner target. [ABSTRACT FROM AUTHOR]

For more than a decade, an unprecedented predict-first activity has been carried in order to predict the fusion power and provide guidance to the second Deuterium–Tritium (D–T) campaign performed at JET in 2021 (DTE2). Such an activity has provided a framework for a broad model validation and development towards the D–T operation. It is shown that it is necessary to go beyond projections using scaling laws in order to obtain detailed physics based predictions. Furthermore, mixing different modelling complexity and promoting an extended interplay between modelling and experiment are essential towards reliable predictions of D–T plasmas. The fusion power obtained in this predict-first activity is in broad agreement with the one finally measured in DTE2. Implications for the prediction of fusion power in future devices, such as ITER, are discussed. [ABSTRACT FROM AUTHOR]