Confinement Degradation Research Articles

Enhanced pedestal transport driven by edge collisionality on Alcator C-Mod and its role in regulating H-mode pedestal gradients

Abstract Experimental measurements of plasma and neutral profiles across the pedestal are used in conjunction with 2D edge modeling to examine pedestal stiffness in Alcator C-Mod H-mode plasmas. Enhanced D α experiments on Alcator C-Mod observed pedestal degradation and loss in confinement below a critical value of net power crossing the separatrix, P net = P net crit ≈ 2.3 MW, in the absence of any external fueling. New analysis of ionization and particle flux profiles reveal saturation of the pedestal electron density, n e ped , despite continuous increases in ionization throughout the pedestal, inversely related to P net . A limit to the pedestal ∇ n e emerges as the particle flux, Γ D , continues to grow, implying increases in the effective particle diffusivity, D eff . This is well-correlated with the separatrix collisionality, ν sep ∗ and a turbulence control parameter, α t , implying a possible transition in type of turbulence. The transition is well correlated with the experimentally observed value of P net crit . SOLPS-ITER modeling is performed for select discharges from the power scan, constrained with experimental electron and neutral densities, measured at the outer midplane. The modeling confirms general growth in D eff , consistent with experimental findings, and additionally suggests even larger growth in χ e at the same P net crit .

Benign Saturation of Ideal Ballooning Instability in a High-Performance Stellarator.

To examine the robustness of the designed 5% β limit for high-performance operation in the W7-X stellarator, we undertake nonlinear magnetohydrodynamic (MHD) simulations of pressure-driven instabilities using the m3d-c1 code. Consistent with linear analyses, ideal ballooning instabilities occur as β exceeds 5% in the standard configuration. Nonetheless, the modes saturate nonlinearly at relatively low levels without triggering large-scale crashes, even though confinement degradation worsens as β increases. In contrast, in an alternative configuration with nearly zero magnetic shear, ideal interchange modes induce a pressure crash at β=1%. These results suggest that the standard W7-X configuration might have a soft β limit that is fairly immune to major MHD events, which enhances the appeal of the stellarator approach to steady-state fusion reactors.

High performance power handling in the absence of an H-mode edge in negative triangularity DIII-D plasmas

Experiments performed during strongly-shaped high-power diverted negative triangularity (NT) experiments in DIII-D achieved detached divertor conditions and a transient-free edge, showcasing the potential for application of NT to a core-edge integrated reactor-like scenario and providing the first characterization of the parametric dependencies for detachment onset. Detached divertor conditions will be required in future devices to mitigate divertor heat fluxes. Access to dissipative divertor conditions was investigated via an increase in upstream density. Detachment onset at the outer strike point was achieved with H-mode level confinement H98−y2∼1 and reactor-relevant normalized pressures βN∼2 . Confinement degradation was observed with deeper detachment, associated with the loss of an electron temperature pedestal. Differences in geometry, radial transport, impact of cross field drifts are discussed to explain differences in access to detachment in NT discharges. Higher normalized densities, with respect to equivalent discharges in positive triangularity, were necessary to achieve detachment, partially explained by the shorter parallel connection length to the targets. The effect of cross-field particle drifts (E×B, B ×∇ B) on access to detachment was demonstrated by the lower upstream density needed to access detachment with ion B ×∇ B drift directed outside of the active divertor (Greenwald fraction fGw∼ 0.9–1.0 vs fGw∼ 1.3). The upstream density at detachment onset was observed to increase linearly with plasma current with ion B ×∇ B drift into the divertor, consistent with the observed narrowing of the scrape-off layer heat flux width λ q . Edge fluid simulations capture separatrix densities needed to achieve detachment in NT plasma and their dependence on drift direction. The ability to reproduce detachment dynamics in NT plasma increases the confidence in future design studies for NT divertors.

Plasmoid drift and first wall heat deposition during ITER H-mode dual-SPIs in JOREK simulations

The heat flux mitigation during the thermal quench (TQ) by the shattered pellet injection (SPI) is one of the major elements of disruption mitigation strategy for ITER. It’s efficiency greatly depends on the SPI and the target plasma parameters, and is ultimately characterised by the heat deposition on to the plasma facing components. To investigate such heat deposition, JOREK simulations of neon-mixed dual-SPIs into ITER baseline H-mode and a ‘degraded H-mode’ with and without good injector synchronization are performed with focus on the first wall heat flux and its energy impact. It is found that low neon fraction SPIs into the baseline H-mode plasmas exhibit strong major radial plasmoid drift as the fragments arrive at the pedestal, accompanied by edge stochasticity. Significant density expulsion and outgoing heat flux occurs as a result, reducing the mitigation efficiency. Such drift motion could be mitigated by injecting higher neon fraction pellets, or by considering the pre-disruption confinement degradation, thus improving the radiation fraction. The radiation heat flux is found to peak in the vicinity of the fragment injection location in the early injection phase, while it relaxes later on due to parallel impurity transport. The overall radiation asymmetry could be significantly mitigated by good synchronization. Time integration of the local heat flux is carried out to provide its energy impact for wall heat damage assessment. For the baseline H-mode case with full pellet injection, melting of the stainless steel armour of the diagnostic port could occur near the injection port, which is acceptable, without any melting of the first wall tungsten tiles. For the degraded H-mode cases with quarter-pellet SPIs, which have 1/4 total volume of a full pellet, the maximum energy impact approaches the tolerable limit of the stainless steel with un-synchronized SPIs, and stays well below such limit for the perfectly synchronized ones.

Degradation of fast-ion confinement depending on the neutral beam power in MHD quiescent LHD plasmas

We investigated the degradation of neutral beam (NB) fast-ion confinement depending on the NB power without magnetohydrodynamics instabilities in the Large Helical Device (LHD). In the LHD deuterium experiment, the neutron emission rate per NB power decreased by up to 20% with increasing injected NBs during a single discharge. Because there were no significant variations in the electron temperature and density, the NB shine-through rate, or the magnetic fluctuation due to the change in NB power, the reduction in the neutron emission rate indicates the degradation of the fast-ion confinement. In this paper, we formulated this degradation depending on the NB power and quantitatively estimated the degraded effective confinement time. In addition, we performed neutron emission rate simulations using the obtained effective confinement time. The simulation and experimental results were in good agreement, suggesting that the degraded effective confinement time is valid.

Experimental Investigations of Quasi-Coherent Micro-Instabilities in J-TEXT Ohmic Plasmas

Quasi-coherent micro-instabilities is one of the key topics of magnetic confinement fusion. This work focuses on the quasi-coherent spectra of ion temperature gradient (ITG) and trapped-electron-mode instabilities using newly developed far-forward collective scattering measurements within ohmic plasmas in the J-TEXT tokamak. The ITG mode is characterized by frequencies ranging from 30 to 100 kHz and wavenumbers (kθρ s) less than 0.3. Beyond a critical plasma density threshold, the ITG mode undergoes a bifurcation, which is marked by a reduction in frequency and an enhancement in amplitude. Concurrently, enhancements in ion energy loss and degradation in confinement are observed. This ground-breaking discovery represents the first instance of direct experimental evidence that establishes a clear link between ITG instability and ion thermal transport.

Effects of edge-localized electron cyclotron current drive on edge-localized mode suppression by resonant magnetic perturbations in DIII-D

According to recent DIII-D experiments (Logan et al 2024 Nucl. Fusion 64 014003), injecting edge localized electron cyclotron current drive (ECCD) in the counter-plasma-current (counter-I p) direction reduces the n = 3 resonant magnetic perturbation (RMP) current threshold for edge-localized mode (ELM) suppression, while co-I p ECCD during the suppressed ELM phase causes a back transition to ELMing. This paper presents nonlinear two-fluid simulations on the ECCD manipulation of edge magnetic islands induced by RMP using the TM1 code. In the presence of a magnetic island chain at the pedestal-top, co-I p ECCD is found to decrease the island width and restore the initially degraded pedestal pressure when its radial deposition location is close to the rational surface of the island. With a sufficiently strong co-I p ECCD current, the RMP-driven magnetic island can be healed, and the pedestal pressure fully recovers to its initial ELMing state. On the contrary, counter-I p ECCD is found to increase the island width and further reduce the pedestal pressure to levels significantly below the peeling-ballooning-mode limited height, leading to even stationary ELM suppression. These simulations align with the results from DIII-D experiments. However, when multiple magnetic island chains are present at the pedestal-top, the ECCD current experiences substantial broadening, and its effects on the island width and pedestal pressure become negligible. Further simulations reveal that counter-I p ECCD enhances RMP penetration by lowering the penetration threshold, with the degree of reduction proportional to the amplitude of ECCD current. For the ∼1 MW ECCD in DIII-D, the predicted decrease in the RMP penetration threshold for ELM suppression is approximately 20%, consistent with experimental observations. These simulations indicate that edge-localized ECCD can be used to either facilitate RMP-driven ELM suppression or optimize the confinement degradation.

Confinement mechanism of cross-shaped steel reinforced concrete columns

A novel confined concrete constitutive model for steel-reinforced concrete (SRC) columns with cross-shaped steel (CSS) sections is proposed. The model is based on the similar characteristics of descending branches in the stress–strain curves of confined and unconfined concrete (UCC), considering confinement degradation caused by buckling of the steel section or longitudinal reinforcement. The confined region of SRC columns with CSS sections was divided into four parts: highly steel-confined concrete (HSCC), partially steel-confined concrete (PSCC), stirrup-confined concrete (StCC) and UCC. Relevant effective confinement coefficient expressions were also derived. Simulations of existing tests showed that (a) the load–strain curves obtained using the modified stress–strain constitutive model agreed well with the experimental results, with the error in the descending branch less than 5%; (b) the HSCC region for SRC columns with CSS sections in the finite-element model was in good agreement with that in the proposed region division; (c) the simulation of confined concrete region boundaries for SRC columns with CSS sections were determined using plumb lines for improving the calculation efficiency.

Tailoring tokamak error fields to control plasma instabilities and transport

A tokamak relies on the axisymmetric magnetic fields to confine fusion plasmas and aims to deliver sustainable and clean energy. However, misalignments arise inevitably in the tokamak construction, leading to small asymmetries in the magnetic field known as error fields (EFs). The EFs have been a major concern in the tokamak approaches because small EFs, even less than 0.1%, can drive a plasma disruption. Meanwhile, the EFs in the tokamak can be favorably used for controlling plasma instabilities, such as edge-localized modes (ELMs). Here we show an optimization that tailors the EFs to maintain an edge 3D response for ELM control with a minimized core 3D response to avoid plasma disruption and unnecessary confinement degradation. We design and demonstrate such an edge-localized 3D response in the KSTAR facility, benefiting from its unique flexibility to change many degrees of freedom in the 3D coil space for the various fusion plasma regimes. This favorable control of the tokamak EF represents a notable advance for designing intrinsically 3D tokamaks to optimize stability and confinement for next-step fusion reactors.

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.

Comparison of partial and deep energy detachment behaviors with Ar seeding on EAST new corner slot divertor

It is necessary for future fusion reactor to reduce the heat fluxes on the entire divertor target, especially if view of long pulse high performance operation. In recent EAST experiments, partial energy detachment without confinement degradation, and deep energy detachment with protection of the entire divertor target have both been confirmed on EAST corner slot divertor by argon (Ar) seeding, which can provide reference for the divertor protection on future fusion reactors. In the deep energy detachment state, the electron temperature T et along entire lower outer divertor target decreases to less than 10 eV and heat fluxes are also strongly mitigated with peak heat flux reduction of more than 90%. Compared to the attached state, there is a moderate confinement degradation with H 98,y2 from ∼1 to ∼0.9 because of Ar radiation in the core region. This confinement degradation can be avoided in the partial energy detachment state, where the radiative power losses in the core are reduced. The experiment and SOLPS-ITER simulation results show that there is no decrease of particle flux js on the divertor target in the partial energy detachment state because the momentum loss in the SOL region is not strong enough. With increasing Ar seeding, there is a js decrease in the deep energy detachment state. The increases of momentum and power losses in the SOL region, and the decrease of upstream pressure all contribute to the js reduction.

First results of high density H-mode operation in metal-wall EAST tokamak

In metal-wall EAST superconducting tokamak, H-mode operation with plasma density close to the Greenwald density limit nGW has been achieved with radio frequency and NBI heating for the first time. Both gas puffing from horizontal plane and HFS pellet fueling were used for density ramp-up during the experiment. The confinement of H-mode gradually deteriorates with plasma density increasing. And the H-L transition can be observed after heating power dropping or pellet injection. In the density range of (0.6–1)×nGW, the divertor detachment occurs and causes an obvious confinement degradation. The maximum accessible density ne, max in H-mode phase deviates from the Greenwald scaling. It has been observed that the fraction ne, max/nGW is almost independent of the total heating power, but a high heating power is helpful to extend the duration of high density H-mode. And ne,max/nGW has an increase relation with the safety factor q95 varied by changing plasma current. Besides, it is also found that the discharges with low plasma current have a higher ne, max/nGW than those with high plasma current. All these findings will provide a good reference to the high density plasma operation for future metal-wall fusion devices.

Tailoring resonant magnetic perturbation to optimize fast-ion confinement during ELM control in KSTAR

3D resonant magnetic perturbation (RMP) is one promising way to control edge localized modes that can cause excessive material erosion of tokamak first walls. However, RMP can lead to undesired degradation of plasma confinement, including fast-particle losses, which can impact the performance and safety of the reactor. This work investigates the optimization of the poloidal spectrum of the 3D field to optimize fast ion confinement during edge localized mode (ELM) suppression. In the initial step, the validity of the modeling framework is tested against experimental data. Simulations successfully replicate an increase in poloidal limiter temperature with different poloidal spectra. Then, the simulation shows improvement of fast ion confinement with a reduction of core resonant response, while edge resonant magnetic fields are maintained above the threshold to sustain the ELM suppression. Reduction of the core resonant fields keeps the Kolmogorov–Arnold–Moser surface and reduces the fast particle losses due to the stochastic magnetic field lines. The results highlight the potential of edge localization of the resonant fields to enhance the performance of fusion reactors, but further investigation is needed to improve the validation of this approach.

Mitigation of divertor edge localised mode power loading by impurity seeding

One of the major challenges for the design of future thermonuclear reactors is the problem of power exhaust—the removal of heat fluxes deposited by plasma particles onto the plasma-facing components (PFCs) of the reactor wall. In order for the reactor to work efficiently, the power loading of the PFCs has to stay within their material limits. A substantial part of these heat fluxes can be deposited transiently during the impact of edge localised modes (ELMs), which typically accompany the high confinement mode, a regime foreseen for tokamak ITER and next-step devices. One of the possible ways to mitigate the deposition of localised heat fluxes during ELMs is injection of impurities, which could similarly to inter-ELM detachment dissipate part of the energy carried by plasma particles, the so-called ELM buffering effect. In this contribution, we report on experimental observations in impurity seeded discharges in ASDEX Upgrade, where injection of argon is capable of reducing the ELM energy by up to 80 (60 without degradation of confinement). A simple model of ELM cooling is in some cases capable of providing quantitative prediction of this effect. The ELM peak energy fluence was reduced by a factor 8 without a degradation of the pedestal pressure. Should such mitigation be achieved in ITER, the resulting power loading would satisfy the material limits of divertor tungsten monoblocks (Eich et al 2017 Nucl. Mater. Energy 12 84–90) and as such avoid the risk of their melting. The most favourable results in terms of confinement and divertor heat flux mitigation were achieved by use of a mixture of argon and nitrogen, where the later impurity helped to improve the confinement. The ELM frequency was identified as a scaling factor for in discharges with impurity seeding, suggesting that high frequency ELMs are favourable for future devices.

Integrated RMP-based ELM-crash-control process for plasma performance enhancement during ELM crash suppression in KSTAR

The integrated Resonant Magnetic Perturbation (RMP)-based Edge-Localized Mode (ELM)-crash-control process aims to enhance the plasma performance during the RMP-driven ELM crash suppression, where the RMP induces an unwanted confinement degradation. In this study, the normalized beta () is introduced as a metric for plasma performance. The integrated process incorporates the latest achievements in the RMP technique to enhance efficiently. The integrated process triggers the n = 1 Edge-localized RMP (ERMP) at the L–H transition timing using the real-time Machine Learning (ML) classifier. The pre-emptive RMP onset can reduce the required external heating power for achieving the same by over 10% compared to the conventional onset. During the RMP phase, the adaptive feedback RMP ELM controller, demonstrating its performance in previous experiments, plays a crucial role in maximizing during the suppression phase and sustaining the -enhanced suppression state by optimizing the RMP strength. The integrated process achieves up to ∼2.65 during the suppression phase, which is ∼10% higher than the previous KSTAR record but ∼6% lower than the target of the K-DEMO first phase ( = 2.8), and maintains the suppression phase above the lower limit of target (= 2.4) for ∼4 s (∼60). In addition to enhancement, the integrated process demonstrates quicker restoration of the suppression phase and recovery of compared to the adaptive control with the n = 1 Conventional RMP (CRMP). The post-analysis of the experiment shows the localized effect of the ERMP spectrum in radial and the close relationship between the evolution of and the electron temperature.

Prevention of mode coupling by external applied resonant magnetic perturbation on the J-TEXT tokamak

Toroidal coupling between m/n = 2/1 and m/n = 3/1 modes frequently occurs in the J-TEXT, where m (n) is the poloidal (toroidal) mode number. These coupled modes destabilize each other, leading to confinement degradation and even triggering a major disruption. This paper presents two control strategies for preventing the mode coupling through the application of a proper static resonant magnetic perturbation (RMP) field. Experimental results demonstrate that moderate 2/1 RMP can suppress the small, rotating 2/1 mode thus prevent coupling between the 2/1 and 3/1 modes. The 3/1 static RMP can excite a large 3/1 locked island while leave the small 2/1 mode rotating at 8 kHz. Enlarging the frequency difference between 2/1 and 3/1 modes makes mode coupling more difficult. Both strategies can break the frequency coupling condition between the 2/1 and 3/1 modes, and hence avoid coupling and mutual destabilizing.

Experimental study of density gradient-driven micro-instabilities and the confinement degradation during H-mode in EAST

A broadband (BB) mode is observed by collective Thomson scattering diagnostics in repeatable shots of EAST and analyzed for the first time. This BB mode usually grows during L–H transitions, featuring a BB quasi-coherent mode with increasing frequency. During H-mode operations, it is characterized by steady-state BB in the high-frequency range (f ∼ 200–2000 kHz), at the electron scale (k θ ρ s = 1‒2), mainly driven by the density gradient, and is sensitive to the value of η e in the region of interest (ρ = 0.4‒0.8), where is the ratio of the normalized electron temperature gradient and density gradient, and the region ρ = 0.4‒0.8 usually has a relatively low collisionality (v eff < 5). The frequency of BB is found to be dependent on the electron temperature and density gradient, which is a typical feature of electron-driven turbulence. A negative correlation between the energy confinement and the intensity of the BB turbulence during H-mode has been found, which indicates a strong electron thermal transport induced by the BB turbulence. The BB significantly decreases the electron temperature and causes flatter electron temperature profiles in the region of interest (ρ = 0.4‒0.8), thus making η e decrease and the BB destabilize further. These characteristics of BB are related to the theoretical density gradient-driven trapped electron mode. It should be noted that this mode is not observed by other diagnostics in EAST, and shows very different features to the coherent modes in the edge.

Use of intrinsic hysteresis for the active control of internal transport barriers in magnetically confined fusion plasmas

In magnetically confined fusion devices, control of internal transport barriers (ITBs) is important both to enhance and suppress the turbulent transport to improve confinement control. Barrier control should allow for the improvement of confinement to aid in achieving the needed fusion criteria while also permitting the degradation of confinement to control profiles and clean the device by moving out the impurities accumulated near the core. In this work, we present a novel control scenario that takes advantage of the hysteresis intrinsic to transport barriers to easily cycle through enhanced and degraded confinement regimes. The control scenario is illustrated using a five-field simplified transport model for an ITB using typical parameters of a neutral beam injection-heated DIII-D tokamak discharge. Pellets and ion cyclotron resonance frequency power are used as control knobs for this active control scenario. These knobs adequately modify at will the local gradients and, therefore, the growth rates and shearing rates, allowing for easy and efficient control of the barrier by taking advantage of the barrier hysteresis. The result is a control cycle that could be operated with a relatively small amount of power in high performance regimes which, nowadays, typically require large power to control. It may also have advantages to avoid, or at least ameliorate, the appearance of magnetohydrodynamic instabilities in the barrier region.

First-principles based plasma profile predictions for optimized stellarators

In the present Letter, first-of-its-kind computer simulations predicting plasma profiles for modern optimized stellarators—while self-consistently retaining neoclassical transport, turbulent transport with 3D effects, and external physical sources—are presented. These simulations exploit a newly developed coupling framework involving the global gyrokinetic turbulence code GENE-3D, the neoclassical transport code KNOSOS, and the 1D transport solver TANGO. This framework is used to analyze the recently observed degradation of energy confinement in electron-heated plasmas in the Wendelstein 7-X stellarator, where the central ion temperature was ‘clamped’ to keV regardless of the external heating power. By performing first-principles based simulations, we provide key evidence to understand this effect, namely the inefficient thermal coupling between electrons and ions in a turbulence-dominated regime, which is exacerbated by the large ratios, and show that a more efficient ion heat source, such as direct ion heating, will increase the on-axis ion temperature. This work paves the way towards the use of high-fidelity models for the development of the next generation of stellarators, in which neoclassical and turbulent transport are optimized simultaneously.

Experimental observations of naturally occurring dust using a high-speed vacuum ultraviolet imaging system in EAST

In the ELMy H-mode experiment, naturally occurring dust originating at the high-field side is clearly observed using the high-speed vacuum ultraviolet imaging system developed on the Experimental Advanced Superconducting Tokamak (EAST). The main ablation cloud shape is similar to the classical shape observed in pellet fueling experiments. However, during the dust penetration, an erupted secondary cloudlet with a bent ‘cigar’ shape is observed and moves upwards along the direction perpendicular to the magnetic field line, which is different to the obviation in the pellet fueling experiments. This may be due to the ion diamagnetic drift effect. The velocities of the secondary cloudlet are estimated to be 50‒80 m s−1. In addition, a significant degradation of the plasma confinement is observed during the dust penetration.