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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.

The intermediate mode between the ohmic, or low confinement (L-mode), and the increased confinement (H-mode) regimes, or the so called I-phase, which is characterized by the existence of zonal flows in the form of limit cycle oscillations (LCOs), was observed on the Globus-M tokamak. Depending on the LCO frequency, the I-phase resulted in either a transition to H-mode or back to L-mode. The possibility of L-I-H transition initiation induced by LCOs and the effect of LCO frequency were studied by means of numerical modeling of the density profile evolution, taking into account turbulence suppression by the inhomogeneous radial electric field. The modeling results show that lower LCO frequency could be a factor facilitating the L-H transition, whereas higher frequency LCOs are more likely to cause the backward transition to L-mode. The results are in qualitative agreement with the results of the studies of geodesic acoustic mode (GAM)–initiated L-H transition in the TUMAN-3M tokamak, where a lower GAM frequency was found to be beneficial for L-H transition initiation. [ABSTRACT FROM AUTHOR]

Geodesic acoustic mode (GAM) is the finite frequency counterpart of zonal flow in toroidal plasmas, which could be excited by and regulate turbulence. This article reports a stationary eigenmode GAM at the inner side of the edge radial electric field contributing to an extreme high confinement plasma with $H_{98}\gt 1.3$ . Such GAM could be observed exclusively in the vicinity of the pedestal top, but not in the pedestal region itself, which represents a previously unreported phenomenon. The bicoherence analysis demonstrates a robust interaction between the GAM and quasi coherent mode (QCM). Statistical evidence indicates that GAMs with such characteristics display a proclivity for a large q _95 and low collisionality in all instances of H-mode accompanied by QCM. Based on the Gyro-kinetic simulation, the QCM is the low n trapped electron mode (TEM) and that the GAM is driven by TEM. Furthermore, the simulations show that the amplitude of GAM declines in conjunction with an increase in the collisionality, which is consistent with the experimental statistics. Additionally, the simulations demonstrate that the turbulent transport exhibits a corresponding decrease as GAM amplitude declines, implying that GAM and its interaction with the QCM would play an important role in the stability and confinement enhancement of such H-mode.

The heavy ion beam probe (HIBP) diagnostics at the TUMAN-3M tokamak was updated to provide measurements in the regime with neutral beam injection co-directed with plasma current (co-NBI). By means of HIBP, plasma potential measurements in the center of plasma were carried out. Plasma potential evolution in the discharge with the LH transition (transition to the improved confinement mode) is in good agreement with the concept of negative radial electric field generation during formation of the transport barrier. [ABSTRACT FROM AUTHOR]

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.

As part the DTE2 campaign in the JET tokamak, we conducted a parameter scan in T and D-T complementing existing pulses in H and D. For the different main ion masses, type-I ELMy H-modes at fixed plasma current and magnetic field can have the pedestal pressure varying by a factor of 4 and the total pressure changing from β N = 1.0 to 3.0. We investigated the pedestal and core isotope mass dependencies using this extensive data set. The pedestal shows a strong mass dependence on the density, which influences the core due to the strong coupling between both plasma regions. To better understand the causes for the observed isotope mass dependence in the pedestal, we analysed the interplay between heat and particle transport and the edge localised mode (ELM) stability. For this purpose, we developed a dynamic ELM cycle model with basic transport assumptions and a realistic neutral penetration. The temporal evolution and resulting ELM frequency introduce an additional experimental constraint that conventional quasi-stationary transport analysis cannot provide. Our model shows that a mass dependence in the ELM stability or in the transport alone cannot explain the observations. One requires a mass dependence in the ELM stability as well as one in the particle sources. The core confinement time increases with pedestal pressure for all isotope masses due to profile stiffness and electromagnetic turbulence stabilisation. Interestingly, T and D-T plasmas show an improved core confinement time compared to H and D plasmas even for matched pedestal pressures. For T, this improvement is largely due to the unique pedestal composition of higher densities and lower temperatures than H and D. With a reduced gyroBohm factor at lower temperatures, more turbulent drive in the form of steeper gradients is required to transport the same amount of heat. This picture is supported by quasilinear flux-driven modelling using TGLF -SAT2 within Astra. With the experimental boundary condition TGLF -SAT2 predicts the core profiles well for gyroBohm heat fluxes > 15 , however, overestimates the heat and particle transport closer to the turbulent threshold. [ABSTRACT FROM AUTHOR]

Experiments performed in DIII-D demonstrate that higher plasma current and heating power combined with impurity seeding facilitate the achievement of divertor detachment with a higher pedestal pressure and higher plasma performance in H-mode plasmas with a baffled closed divertor compared with an open divertor. Dedicated experiments were carried out to study the impact of power, plasma current and impurity seeding on divertor detachment with ion B × ∇ B directed into the divertor favorable for the L–H transition. With a factor of three variation in heating power and with only D2 puffing, no significant difference in the separatrix density at detachment onset was found. The higher heating power leads to higher impurity concentration and wider scrape-off layer (SOL) width, and reduces the detachment onset density to one similar to that in lower-power plasmas. Higher current requires higher pedestal and line-averaged densities to achieve divertor detachment; however, the increase in separatrix density at increasing plasma current is found to be less pronounced. Initial calculations found that both power scan and plasma current scan datasets are qualitatively consistent with theory after considering the change in impurity concentration and heat flux width. This also motivates the future extensive study of transport and divertor impurity behavior in order to have a quantitative comparison between experiment and theory. Compared with an open divertor, a closed divertor facilitates detachment onset at ∼40% lower line-averaged plasma density. Additional N2 seeding facilitates the achievement of detachment at a lower separatrix density and thus a higher pedestal temperature, which is beneficial for advanced tokamak scenarios. Higher heating power requires a higher N2 puffing rate to achieve the same degree of detachment, while a higher N2 puffing rate leads to lower detachment onset line-averaged density, both of which agree with theory. In contrast to the narrower pedestal in an open divertor approaching detachment, the pedestal density width in a closed divertor increases with density. The density gradient increases with line-averaged density at higher plasma current, but remains nearly unchanged at lower plasma current. In particular, compared with discharges with low power, at high heating power the pedestal density gradient is much weaker, while the SOL density is significantly higher and wider. At the same plasma current, both pedestal pressure gradient and temperature gradient decrease linearly with the line-averaged density but remain similar across different heating powers. Even with different plasma current and heating power, the normalized pressure gradient remains identical. As a result, achievement of divertor detachment with a higher pedestal pressure and higher plasma performance is shown in a closed divertor, which is important for improving core–edge integration as one of the critical issues for future tokamak fusion reactors. [ABSTRACT FROM AUTHOR]

Recent progress towards an increased understanding of detached divertor physics has been made with the highly closed divertor geometry in HL-2A. Non-intrinsic impurities were injected into the outer divertor chamber, and increased divertor neutral pressure and enhanced volumetric energy loss in the divertor were observed. Meanwhile the neutral pressure in the main chamber decreased slightly, and neutral compression between the divertor and main chamber increased greatly. This led to divertor detachment with a low upstream plasma line-averaged electron density ( ∼ 0.5 n GW ). In the H-mode, slight degradation of the core confinement, characterized by a decrease in plasma stored energy and pedestal pressure and an increase in edge-localized mode frequency, was observed, but the H-mode was still sustained well with H98 > 1. Pedestal density fluctuation was increased during detachment, implying that the enhanced pedestal transport might be responsible for the degradation. During the divertor detachment phase, the impurities were well controlled in the divertor without strong radiation near the X-point region, and the main plasma density did not increase but decreased slightly; this could be a benefit of the highly closed divertor geometry. The experimental results suggest that a closed divertor geometry has the advantages of volumetric energy loss, gas pumping and impurity control in the divertor without significant effects on the plasma confinement, thus giving a wider operating window for divertor detachment. [ABSTRACT FROM AUTHOR]

This paper describes scalings of the H-mode confinement database of neutral beam injection (NBI)-lower hybrid wave (LHW) and electron cyclotron resonant heating-LHW plasmas in the EAST tokamak. Fundamental information of the EAST H-mode database and the details of calculation of the thermal energy confinement time in EAST are presented. The result of the scaling model τ E,EASTS,H,NBI - LHW of the thermal energy confinement time in NBI-LHW plasmas with root-mean-square-error RMSE = 5.5 % , compared with the multi-machine scaling model τ th,98 y 2 derived from the multi-machine database (mainly NBI heating) with RMSE = 12.9 % , is obtained. The scaling τ E,EAST,H,NBI - LHW presents a similar dependency on plasma current and power loss and a lower dependency on plasma density and elongation, but stronger dependency on toroidal field than in τ th,98 y 2 . The scaling τ E,EAST,H,ECRH - LHW demonstrates a similar dependency on plasma current and power loss and lower dependency on plasma density, but a negative dependency with large error bars on elongation compared with τ E,EAST,H,NBI - LHW in NBI-LHW plasmas. The comparison of EAST models with the more recent multi-tokamak scaling ITPA20-IL shows a similar engineering dependency on plasma current, power loss and plasma density, but opposite dependency on toroidal field. [ABSTRACT FROM AUTHOR]

Based on a heuristic model of the scrape-off-layer (SOL) plasma dynamics, a new type of limit-cycle-oscillation (LCO) is shown to be possible in the SOL region. The key trigger process is the positive feedback amplification of the SOL radial electric field through the SOL turbulence suppression and temperature profile steepening. This SOL-LCO can make the edge plasma inside the separatrix also have an oscillation motion, taking the features qualitatively similar to those observed in the recent ASDEX Upgrade experiment (Cavedon et al 2017 Nucl. Fusion 57 014002). With the above positive feedback process much more feasible when the ion magnetic curvature drift direction is toward the X -point, the present model also provides an explanation of why the LCO is mostly observed in such a favorable magnetic configuration. [ABSTRACT FROM AUTHOR]

The impurity transport driven by kinetic ballooning mode (KBM) is theoretically studied in the DIII-D H-mode strong gradient pedestal plasmas. From the electromagnetic gyrokinetic equation, including the correction of the strong radial electric field, the dispersion relationship of KBM instability with non-trace impurity is firstly derived. Then, the turbulent impurity flux and ion heat flux, as well as the associated transport coefficients, are further calculated. Through the parametric dependence analysis of analytical results, it is found that dilution effects of light fully ionized impurities can reduce the drive of KBM by affecting the kinetic pressure gradient parameter $\alpha $ and diamagnetic effects, thus leading to a decrease in both the absolute value of the real frequency $\left| {\omega_\text{r}^{\prime}}\right|$ and the growth rate $\gamma_{\text{k}}^{\prime}$ of KBM instability. Stronger dilution effects by increasing the impurity charge number $Z$ or steepening the impurity density profile correspond to stronger effects. Moreover, the removal efficiency of light fully ionized impurities, quantified by the ratio between the impurity diffusivity and effective ion heat conductivity $\frac{{{D_{\text{z}}}}}{{\chi _{\text{i}}^{{\text{eff}}}}}\approx \frac{{1 - {b_{\text{z}}}+ {{4{\omega _{{\text{Dz}}}}}\mathord{\left/ {\vphantom {{4{\omega _{{\text{Dz}}}}}{\omega_\text{r}^{\prime}}}}\right. }{{{\omega_\text{r}^{\prime}}}}}}}{{{{3\left({1+{1\mathord{\left/ {\vphantom {1 {{\eta _{\text{i}}}}}}\right. }{{\eta _{\text{i}}}}}}\right)}\mathord{\left/ {\vphantom {{3\left( {1 + {1 \mathord{\left/ {\vphantom {1 {{\eta _{\text{i}}}}}}\right. }{{\eta _{\text{i}}}}}}\right)}2}}\right. }2}}}$ , increases with an increase of $Z$ mainly due to the smaller impurity finite Larmor radius (FLR) effects reflected by ${b_{\text{z}}}\propto {1 \mathord{\left/ {\vphantom {1 Z}}\right. }Z}$ . Besides, the increase of the impurity density gradient can significantly enhance ${{{D_{\text{z}}}}\mathord{\left/ {\vphantom {{{D_{\text{z}}}}{\chi _{\text{i}}^{{\text{eff}}}}}}\right. }{\chi _{\text{i}}^{{\text{eff}}}}}$ , and this is because stronger impurity dilution effects make a larger magnetic drift term ${{\left| {{\omega _{{\text{Dz}}}}}\right|}\mathord{\left/ {\vphantom {{\left| {{\omega _{{\text{Dz}}}}}\right|}{\left| {\omega_\text{r}^{\prime}}\right|}}}\right. }{\left|{\omega_\text{r}^{\prime}}\right|}}$ ( ${\omega _{{\text{Dz}}}}$ is the magnitude of impurity magnetic drift frequency) and ${\eta _{\text{i}}}$ (the ratio of ion density gradient scale length to ion temperature gradient scale length). For heavy metal impurities with a concentration of ${10^{ - 4}}$ , the peaking factor (PF) is positive, which means that its density profile is inwardly peaked, and the PF decreases with the enhancement of impurity FLR effects. These results may provide some theoretical reference on understanding the physical mechanism of impurity transport in the pedestal of H-mode plasmas.

Recent JET D–T campaigns opened the possibility of unique isotope studies to investigate the L–H transition physics in view of reactor plasmas and to study the origin of the observed power threshold minimum. In the present paper, we characterise L–H transitions in the low and high-density branches of JET NBI-heated D–T plasmas. As discussed in the paper, L–H transition has been hypothesised to be determined by the transport power losses of plasma ions, i.e. the so-called ion heat flux ( Q _i ). We present the first power balance analysis of JET NBI-heated D–T plasmas to evaluate the ion heat flux at the transition. Due to the experimental setting being similar to previous JET D experiments, we also directly compare the results, discussing the isotope effect and similarities between datasets. First, we find an isotope effect between D and D–T Q _i , with a lower Q _i in D–T plasmas. We confirm that the ion heat flux deviates from density linearity compared to the linear trend observed in wave-heated D plasmas of other tokamaks. The deviation we observe in NBI-heated L–H transitions happens at an isotope-dependent density. Plasma edge rotation correlates with Q _i deviation from density linearity in the low-density branch. However, further investigations would be required to assess the role of rotation on Q _i and the power threshold minimum at JET. At low plasma density, NBI power dominates Q _i , while increasing the density makes the equipartition power dominant. We finally compare our results with hypotheses proposed from evidence in other tokamaks to present a complete overview of ion heat flux analyses in D and D–T NBI-heated plasmas at JET.

The impact of plasma shaping on the properties of high density H-mode scrape-off layer (SOL) profiles and transport at the outer midplane has been investigated on Tokamakà configuration variable. The experimental dataset has been acquired by evolving the upper triangularity while keeping the other parameters constant. The scan comprises $\delta_\mathrm{up}$ values between 0.0 and 0.6, excluding negative triangularity scenarios. Within this study, a transition from type-I edge localised modes to the quasi-continuous exhaust regime takes place from low to high $\delta_\mathrm{up}$ . The modification of the upstream SOL profiles has been assessed, in terms of separatrix quantities, within the $\alpha_\mathrm{t}$ turbulence control parameter theoretical framework (Eich et al 2020 Nucl. Fusion 60 056016). The target parallel heat load and the upstream near-SOL density profiles have been shown to broaden significantly for increasing $\alpha_\mathrm{t}$ . Correspondingly, in the far SOL a density shoulder formation is observed when moving from low to high $\delta_\mathrm{up}$ . These behaviours have been correlated with an enhancement of the SOL fluctuation level, as registered by wall-mounted Langmuir probes as well as the thermal helium beam diagnostic. Specifically, both the background and the filamentary-induced fluctuating parts of the first wall ion saturation current signal are larger at higher $\delta_\mathrm{up}$ , with filaments being ejected more frequently into the SOL. Comparison of two pulses at the extremes of the $\delta_\mathrm{up}$ scan range, but with otherwise same input parameters, shows that the midplane neutral pressure does not change much during the H-mode phase of the discharge. This indicates that indirect effects of the change in geometry, linked to first wall recycling sources, should not play a significant role. The total core radiation increases at high $\delta_\mathrm{up}$ , on account of a stronger plasma–wall interaction and resulting larger carbon impurity intake from the first wall. This is likely associated to the enhanced first wall fluctuations, as well as a smaller outer gap and the close-to-double-null magnetic topology at high shaping.

In the pedestal region of the Experimental Advanced Superconducting Tokamak (EAST) during high confinement mode plasma operations with radio-frequency heating, two distinct fluctuations are observed: high-frequency fluctuations (HFFs) and edge harmonic oscillation-like (EHO-like) modes. The HFFs are characterized by intermittent fluctuations with a broadband frequency range of $1{-}3{\text{ MHz}}$ and a poloidal wave number ( ${k_\theta }$ ) greater than $0.9{\text{ c}}{{\text{m}}^{ - 1}}$ . On the other hand, the EHO-like mode exhibits characteristics similar to magnetohydrodynamics (MHD)-like modes with n = 1−5 and lower poloidal wave numbers ( ${{\text{k}}_{{\theta }}} \unicode{x2A7D} 0.12{\text{ c}}{{\text{m}}^{ - 1}}$ ). During the pedestal establishing phase following the L–H transition, a significant concurrent presence of HFF and EHO-like modes in high-density pedestal regions has been noted. In this phase, the EHO-like mode not only modulates the amplitude of the HFF but also engages in nonlinear interactions. The occurrence of EHO-like mode and HFF is associated with particle transport toward the divertor, though it is notably less than that caused by edge coherent modes. During the inter-edge localized mode (ELM) period, a significant decrease in the ${D_\alpha }$ baseline is observed whenever the low frequency fluctuation (LFF) weakens and the HFF grows, prior to each large ELM. One possible explanation is that the rapid increase of $E \times B$ shear stabilizes the LFF and destabilizes the HFF, which lowers the pedestal transport and enables the further growth of the pedestal until the onset of the ELM.

To achieve their goals, future thermonuclear reactors such as ITER and DEMO are expected to operate plasmas with a high magnetic field, triangularity and confinement. To address the corresponding challenges, the concept of the high-field ( $B_{\textrm{T}} \unicode{x2A7D}$ 5 T), high-current ( $I_{\textrm{P}} \unicode{x2A7D}$ 2 MA) COMPASS Upgrade tokamak was established, and the device is currently being constructed in Prague, Czech Republic. This contribution provides an overview of the priority physics topics for the future physics programme of COMPASS Upgrade, namely: (i) characterisation of alternative confinement modes, (ii) a power exhaust including liquid metals, (iii) operation with a hot first wall and (iv) the influence of plasma shape on pedestal stability and confinement. The main scenarios are presented, as predicted by METIS and FIESTA codes. Pedestal pressure and density are estimated using EPED, multi-machine semi-empirical scaling and a neutral penetration model. Access to detachment is estimated using a detachment qualifier.

The RFX-mod2 device, the upgraded version of the previous RFX-mod with a modified magnetic boundary, is presently under realization and will start to be operated in 2025. Significant upgrades of the diagnostic capabilities have been proposed and are under development. These include a largely increased number of in-vessel magnetic and electrostatic sensors, a new fast reciprocating manipulator for the exploration of the edge plasma in a wide range of experimental conditions, the improved Thomson scattering and soft x-ray diagnostics system for a detailed determination of the behavior of the electron temperature profile, new dedicated systems for the space and time resolved analysis of x-ray spectra and neutron rate, a reflectometric diagnostic for real-time determination of plasma position, two diagnostics devoted to the imaging of light impurities and influxes behavior along with arrays of halo current sensors. These diagnostic upgrades will be accompanied by a significant effort to improve the control of the electron density and of the impurity influxes by means of proper treatment of plasma facing components with in-vessel fixed electrodes distributed over the first wall. The described advancements will allow a deeper understanding of physics phenomena in the wide variety of magnetic configurations, including the tokamak, the reversed-field pinch and the Ultra-low q, which can be produced in RFX-mod2 thanks to its flexibility and unique MHD control capabilities.

Understanding the physics of low-confinement (L-), improved-confinement (I-), and high-confinement (H-) modes is critical for fusion reactors. The finding herein reports observations of two types of turbulence coexisting near the L-mode edge, one magnetohydrodynamic (MHD)-like and another micro-tearing mode (MTM)-like, linked to the H-mode and I-mode confinement in the DIII-D tokamak. Ion-scale magnetic and density turbulence is measured using a Faraday-effect radial-interferometer-polarimeter and beam-emission-spectroscopy (BES). Broadband turbulence spectra of up to ∼600 kHz are observed in two discharges where transitions between L-mode, I-mode, and H-mode occurs. Turbulence is found to be inversely correlated with confinement, meaning lower turbulence power at higher confinement. Distinctively, the high-frequency (HF, >∼100 kHz) magnetic turbulence power changes by the most (55%) during transitions primarily involving energy confinement change, whereas the low-frequency (LF,

A series of dedicated experiments from the DIII-D tokamak provide spatially and temporally resolved measurements of electron density and temperature, and multiscale and multichannel fluctuations over a wide range of conditions. Measurements of long wavelength density fluctuations in the type-I ELMing H-mode pedestals routinely reveal a coexistence of multiple instabilities that exhibit dramatic different dynamic behaviors as q _95 and temperature gradients are varied, apparently responsible for limiting pedestal temperature profiles. Two distinct frequency bands of density fluctuations are modulated by an ELM cycle with frequency above 200 kHz propagating in the electron diamagnetic direction in the lab frame (electron mode) and below 200 kHz propagating in the ion diamagnetic direction (ion mode). The electron mode amplitude peaks near the electron temperature gradient region and increases with q _95 which seems to be correlated with the increased χ _e at higher q _95 , similar to the characteristics expected for the micro-tearing mode (MTM). At higher q _95 , during the inter-ELM period, the ion mode decays at the later phase of the ELM cycle. Consistently, the poloidal correlation length of the ion mode is also found to reduce, which suggests the possible E × B flow shear suppression of the ion mode at the later phase of the ELM cycle as the E _r well recovers. In contrast, the electron mode grows during the ELM cycle and reaches saturation at around 50%–60% of the ELM period. Linear gyrokinetic simulations find the MTMs to be the most unstable mode in the pedestal electron temperature gradient region. The higher q _95 and lower magnetic shear destabilize the MTMs. These observations provide key insights into the underlying physics of multifield properties and a rich dataset of experimental ‘fingerprints’ that enable new tests of theoretical pedestal models and lead to the development of a predictive model for pedestal formation on the ITER and future burning plasma experiments.

For its initial operational phase, ITER has until recently considered using non-nuclear hydrogen (H) or helium (He) plasmas to keep nuclear activation at low levels. To this end, the Tokamak Exploitation Task Force of the EUROfusion Consortium carried out dedicated experimental campaigns in He on the ASDEX Upgrade (AUG) and JET tokamaks in 2022, with particular emphasis put on the ELMy H-mode operation and plasma-wall interaction processes as well as comparison to H or deuterium (D) plasmas. Both in pure He and mixed He + H plasmas, H-mode operation could be reached but more effort was needed to obtain a stable plasma scenario than in H or D. Even if the power threshold for the LH transition was lower in He, entering the type-I ELMy regime appeared to require equally much or even more heating power than in H. Suppression of ELMs by resonant magnetic perturbations was studied on AUG but was only possible in plasmas with a He content below 19%; the reason for this unexpected behaviour remains still unclear and various theoretical approaches are being pursued to properly understand the physics behind ELM suppression. The erosion rates of tungsten (W) plasma-facing components were an order of magnitude larger than what has been reported in hydrogenic plasmas, which can be attributed to the prominent role of He ^2+ ions in the plasma. For the first time, the formation of nanoscale structures (W fuzz) was unambiguously demonstrated in H-mode He plasmas on AUG. However, no direct evidence of fuzz creation on JET was obtained despite the main conditions for its occurrence being met. The reason could be a delicate balance between W erosion by ELMs, competition between the growth and annealing of the fuzz, and coverage of the surface with co-deposits.

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.

We study numerically the feasibility of achieving the L-H transition in the current EU-DEMO tokamak baseline using uniquely direct electron heating. The ASTRA code coupled to the TGLF turbulent transport model is used to predict steady-state kinetic plasma profiles for diverse numerical scans. Among them, we have varied the separatrix electron density, the total amount of ECRH power, the microwave beam deposition profile and the plasma impurity content. The solutions are then compared to L-H transition scaling laws to assess whether the found plasma state would enter into H-mode. We find the plasma density and impurity content to be the key variables setting the boundaries in parameter space where the L-H transition is feasible. When impurities can be controlled under a certain threshold, given here for a fully shaped DEMO plasma, the L-H transition is achieved in all the studied conditions.

A predictive model for the electron temperature profile of the H-mode pedestal is described, and its results are compared with the pedestal structure of JET-ILW plasmas. The model is based on a scaling for the gyro-Bohm normalized, turbulent electron heat flux qe/qe,gB resulting from electron temperature gradient (ETG) turbulence, derived from results of nonlinear gyrokinetic (GK) calculations for the steep gradient region. By using the local temperature gradient scale length LTe in the normalization, the dependence of qe/qe,gB on the normalized gradients R/LTe and R/Lne can be represented by a unified scaling with the parameter ηe=Lne/LTe, to which the linear stability of ETG turbulence is sensitive when the density gradient is sufficiently steep. For a prescribed density profile, the value of R/LTe determined from this scaling, required to maintain a constant electron heat flux qe across the pedestal, is used to calculate the temperature profile. Reasonable agreement with measurements is found for different cases, the model providing an explanation of the relative widths and shifts of the Te and ne profiles, as well as highlighting the importance of the separatrix boundary conditions. Other cases showing disagreement indicate conditions where other branches of turbulence might dominate. [ABSTRACT FROM AUTHOR]

The dithering H-mode phase, characterized by oscillations, is generally observed at input power values close to the L-H transition power threshold and low plasma collisionalities (low electron density and/or high plasma temperature). Measurements to characterize the dithering phase are presented for the low aspect ratio, high magnetic field tokamak, ST40. The dithering phase oscillation frequency is observed between 400 and 800 Hz and demonstrates an inverse relationship with core plasma density. Dithering phase H-modes are documented across a nonlinear, low-density power threshold operational space, with signature low- and high-density branches. The minimum power threshold for dithering H-mode access is measured at a core, line average electron density of 4.7(±0.5)×1019 m−3, close to a predicted value of 4.1(±0.4)×1019 m−3 from multimachine studies. ASTRA calculated values of power coupled to the ion species, at the dithering H-mode transition, exhibit a similar nonlinear dependence on density. This analysis points to the important contribution of the ion thermal channel to the L-H phase transition. The low-frequency plasma density and D-alpha dithers appear to be accompanied by sudden bursts of magnetohydrodynamic (MHD) activity. A simple model is tested to demonstrate a possible scenario of self-regulation among turbulence, zonal flows, pressure (density) gradient and MHD activities. [ABSTRACT FROM AUTHOR]

The transition in the turbulence spectrum from ion-scale dominated regimes to multiscale transport regimes that couple ion and electron scales is studied with gyrokinetic simulations of turbulent transport. The simulations are based on DIII-D high-confinement mode (H-mode) plasma parameters in the tokamak pedestal. The transition is initiated by varying the ion temperature gradient. To our knowledge, no full multiscale simulations of pedestal-like transport have been done previously. The experimental parameters lie in a bifurcation region between the two regimes. At long wavelengths, a complex, ion-direction hybrid mode is the dominant linearly unstable drift wave, while an electron temperature gradient-driven mode is unstable at short wavelengths. In the transition from the multiscale branch to the ion-scale branch, the magnitude of the ion-scale poloidal wavenumber spectrum of the nonlinear turbulent energy flux increases and the magnitude of the high-wavenumber spectrum decreases. The decrease in the electron-scale transport is due to nonlinear mixing with ion-scale fluctuations and the ion-scale-driven zonal flows. A shift in the total energy associated with the fluctuating electrostatic potential intensity from dominantly drift kinetic energy in the multiscale regime to dominantly potential intensity in the ion-scale regime is well-correlated with the trend in the total energy flux. [ABSTRACT FROM AUTHOR]

This paper presents the first results of a study of the LH transition on the new spherical Globus-M2 tokamak using the Doppler backscattering (DBS) diagnostic. New data characterizing the H-mode of discharges with higher values of the plasma parameters, such as magnetic field Bt up to 0.9 T and plasma current Ip up to 450 kA, were collected and analyzed. An upgraded neutral beam injection (NBI) system was used to initiate the LH transition. DBS allows the measurement of the poloidal rotation velocity and the turbulence amplitude of the plasma. The multi-frequency DBS system installed on Globus-M2 can simultaneously collect data in different areas spanning from the separatrix to the plasma core. This allowed for the radial profiles of the rotation velocity and electric field to be calculated before and after the LH transition. In addition, the values and temporal evolution of the velocity shear were obtained. The associated turbulence suppression after the transition to the H-mode was investigated using DBS. [ABSTRACT FROM AUTHOR]

Net erosion of plasma-facing materials was investigated at the low-field-side (outer) strike-point area of the ASDEX Upgrade (AUG) divertor during H-mode discharges with small and frequent ELMs. To this end, Au and Mo marker samples with different surface morphologies and geometries were exposed to plasmas using the DIM-II divertor manipulator. The results were compared to existing erosion and deposition patterns from various L- and H-mode experiments, in the latter case the main difference was the size and frequency of the ELMs.We noticed that increasing surface roughness reduces net erosion but less than what is the case in L-mode. On the other hand, net-erosion rates in H-mode are generally 2–5 times higher than the corresponding L-mode values, in addition to which exposure in H-mode conditions results in strong local variations in the poloidal and toroidal erosion/deposition profiles. The latter observation we associate with the large migration length, on the order of several cm, of the eroded material, resulting in strong competition between erosion and re-deposition processes especially at poloidal distances > 50 mm from the strike point. Considerable net erosion was measured throughout the analysed poloidal region unlike in L-mode where the main erosion peak occurs in the vicinity of the strike point. We attribute this qualitative difference to the slow decay lengths of the plasma flux and electron temperature in the applied H-mode scenario.Both erosion and deposition require detailed analyses at the microscopic scale and the deposition patterns may be drastically different for heavy and light impurities. Generally, the rougher the surface the more material will accumulate on locally shadowed regions behind protruding surface features. However, rough surfaces also exhibit more non-uniformities in the quality or even integrity of marker coatings produced on them, thus complicating the analyses of the experimental data.We conclude that local plasma parameters have a huge impact on the PFC erosion rates and, besides incident plasma flux, surface morphology and its temporal evolution have to be taken into account for quantitative estimates of erosion rates and PFC lifetime under reactor-relevant conditions.

Particle and heat fluxes were successfully controlled by using a continuously flowing liquid Li (FLiLi) limiter in the H-mode discharges with high plasma heating power in the Experimental Advanced Superconducting Tokamak device. There were strong interactions between the FLiLi limiter and high-power plasma with a ∼ 8.3 MW source heating power, and successively, a bright Li radiation ring was produced, which effectively decreased fuel particle recycling by approximately 50%. Due to Li efflux from FLiLi during a series of high-power discharges, an obvious real-time wall conditioning effect was produced, and fuel particle recycling further decreased. Moreover, the value of Zeff decreased from 2.3 to 1.6 due to a decrease in impurity sources; this was attributed to the accumulation of Li deposited on the first wall, which effectively protected the wall materials. The decreased recycling and impurity radiation achieved high-energy confinement plasma, and the average stored energy increased up to ∼ 290 kJ. Moreover, due to the effect of Li vapor shielding, nearly 30% plasma heat flux was dissipated before it arrived at the Li limiter. These results promote further exploration of liquid Li solutions for the critical challenge of heat flux handling and particle control in fusion power plants.

The high confinement mode (H-mode) is the widely adopted standard operation scenario for the path to fusion in toroidal confinement devices. Since its discovery in 1982, the H-mode and access to the H-mode (the low to high and high to low transitions) remain two of the most actively researched areas in magnetically confined fusion programmes across the world. Significant progress has been made in the understanding of the intricate H-mode phase dynamics in recent years, from improvement in experimental diagnostic capability, theoretical development and modelling. The ‘H-mode transition and pedestal studies’ Special Issue provides a timely overview of recent progress in the study of H-modes covering experimental studies, further theoretical inquiry and computational modelling. [ABSTRACT FROM AUTHOR]

Recent evidence points toward the microtearing mode (MTM) as an important fluctuation in the H-mode pedestal for anomalous electron heat transport. A study of the instabilities in the pedestal region carried out using gyrokinetic simulations to model an ELMy H-mode DIII-D discharge (USN configuration, 1.4 MA plasma current, and 3 MW heating power) is presented. The simulations produce MTMs, identified by predominantly electromagnetic heat flux, small particle flux, and a substantial degree of tearing parity. The magnetic spectrogram from Mirnov coils exhibits three distinct frequency bandsâ€"two narrow bands at lower frequency (∼35â€"55 kHz and ∼70â€"105 kHz) and a broader band at higher frequency (∼300â€"500 kHz). Global linear GENE simulations produce MTMs that are centered at the peak of the ω * profile and correspond closely with the bands in the spectrogram. The three distinctive frequency bands can be understood from the basic physical mechanisms underlying the instabilities. For example (i) instability of certain toroidal mode numbers (n) is controlled by the alignment of their rational surfaces with the peak in the ω * profile, and (ii) MTM instabilities in the lower n bands are the conventional collisional slab MTM, whereas the higher n band depends on curvature drive. While many features of the modes can be captured with the local approximation, a global treatment is necessary to quantitatively reproduce the detailed band gaps of the low- n fluctuations. Notably, the transport signatures of the MTM are consistent with careful edge modeling by SOLPS. [ABSTRACT FROM AUTHOR]

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Geodesic acoustic modes (GAMs) represent the oscillating counterpart of zonal flow in tokamak plasma and can affect transport due to their interaction with turbulence eddies. GAMs have been observed in many experiments and modelled under different conditions, but because of their variety of characteristics, we do not yet have a complete picture of their dynamics. It has been demonstrated that optical methods can be efficiently used to describe and predict several characteristics of the GAM radial structures that can be interpreted as ‘waves’ propagating in the space-time. We exploit complex eikonal theories to investigate the behavior of GAMs that are commonly observed in experiments, and find that their periodic modulation and intermittency can be explained by the properties of the equilibrium temperature profile. Theoretical results obtained in this work are supported by gyrokinetic simulations for several equilibria. Implications for existence criteria and GAM dynamics in different plasma equilibrium conditions are discussed, with particular attention to the edge plasma in low and high confinement modes.

The quasi-continuous exhaust (QCE) regime, formerly known as either type-II ELM or small ELM regime is studied in ASDEX Upgrade. The regime is a natural type-I ELM-free H-mode. The operational space of QCE discharges in ASDEX Upgrade with respect to their separatrix conditions and their power exhaust capabilities are presented. A significant broadening of the power fall-off length is observed, correlating to an increased separatrix density and pressure. Moreover, the possible reactor relevance of this regime is demonstrated by expanding the operational space to low edge safety factor and demonstrating the benign tungsten impurity behaviour. A discharge without any type-I ELM from start to end reaching a partially detached divertor at high normalised energy confinement time is presented.

During the last several years, the HL-2A experiment has made significant progress in the following areas: (1) lower-hybrid wave (LHW) heating and current drive, (2) plasma confinement and turbulent transport, (3) magnetohydrodynamic (MHD) instabilities and energetic particle physics and (4) H-mode and edge localized mode (ELM) control. The results show that the LHW system working in the co-current mode can reach higher driving efficiency and full non-inductive lower-hybrid current drive (LHCD) has been achieved. The intrinsic poloidal torque characterized by the divergence of the residual stress is deduced from synthesis for the first time. The dynamics of spectral symmetry breaking in drift wave turbulence is in good agreement with the development of the poloidal torque to drive the edge poloidal flow. The influence of the cross-phase dynamics on turbulent stress was also investigated. The ion internal transport barrier has been observed in the NBI-heated plasma, and inside the barrier the ion thermal transport is reduced to the neoclassical level. Besides, micro-turbulence is modulated by the rotation frequency of the magnetic island, and this modulation effect is related to a critical island width. Strong E × B shear is found at the island boundary. Three kinds of axisymmetric modes, beta-induced Alfven eigenmode (BAE), toroidal Alfven eigenmode (TAE) and the ellipticity-induced Alfven eigenmode (EAE), are found to be driven unstable by nonlinear mode coupling between Alfven eigenmodes and tearing mode which is well explained by the nonlinear gyrokinetic theory. The fishbone and tearing modes were actively controlled by the electron cyclotron resonance heating (ECRH). The dynamics of the edge plasma flows and turbulence during the L–I–H transition have been dedicatedly investigated. The geodesic acoustic mode (GAM) and limit cycle oscillation (LCO) coexist for a short time and disappear in the H-mode plasma with the increasing of E × B shear flow before the I–H transition, which plays an important role in the turbulence suppression. Different techniques, such as LHW, ECRH, resonant magnetic perturbation (RMP), and impurity seeding by the laser blow-off (LBO) and supersonic molecular beam injection (SMBI), have been successfully applied to control the large ELMs. It has been found that pedestal turbulence enhancement might be responsible for the observed mitigation effect. [ABSTRACT FROM AUTHOR]

Here we report the results of the turbulence study in the high-confinement mode (H-mode) with and without edge localized modes (ELMs). The study was performed by the Doppler backscattering (DBS) method on the Globus-M tokamak. Two types of ELMs were observed in the Globus-M tokamak during NBI heating and accordingly, two different transitions to transient ELM-free modes were discovered. It was demonstrated that the transition from the H-mode with ELMs accompanied by small sawtooth oscillations to the ELM-free H-mode is characterized by the drop of the turbulence amplitude near the periphery while the transition in the case of ELMs is accompanied by large sawtooth oscillations, which occur without suppression of peripheral turbulence. [ABSTRACT FROM AUTHOR]

Midplane separatrix density is a crucial parameter in tokamaks since it strongly impacts divertor conditions. Scaling midplane separatrix density, ne, SEP, and pedestal density, ne, PED, as function of engineering parameters such as auxiliary heating Pinjected, toroidal magnetic field BT, and plasma poloidal current Ip are relevant to observe the effect of tuning these parameters on, for example, quality of confinement and divertor regime governed by ne, PED and ne, SEP, respectively. Thus, a dataset of JET H‐mode pulses performed with Iter like wall (ILW) has been analysed. Midplane density data are collected from an HRTS (high‐resolution Thomson scattering) diagnostic and ne, SEP is determined using the power balance method. Parallel heat flux model is chosen using transport code SOLEDGE2D (S2D) applying power balance method over a simulated ne, SEP and Te, SEP profiles to obtain separatrix positions. The parameters are averaged over time windows with order of (85–185 ms) and the magnetic configuration has been fixed to avoid divertor geometrical effect on ne, SEP determination, configuration chosen is corner–corner. A ratio between separatrix density and pedestal density at outer midplane ranges between 0.3 and 0.7 on the data set. A scaling law of ne, SEP/ne, PED is obtained as function of Pinjected, BT, and IP. [ABSTRACT FROM AUTHOR]

Screening of high-Z (W) impurities from the confined plasma by the temperature gradient at the plasma periphery of fusion-grade H-mode plasmas has been demonstrated in the JET-ILW (ITER-like wall) tokamak. Through careful optimisation of the hybrid-scenario, deuterium plasmas with sufficient heating power ( $\gtrsim$ 32 MW), high enough ion temperature gradients at the H-mode pedestal top can be achieved for the collisional, neo-classical convection of the W impurities to be directed outwards, expelling them from the confined plasma. Measurements of the W impurity fluxes between and during edge-localised modes (ELMs) based on fast bolometry measurements show that in such plasmas there is a net efflux (loss) between ELMs but that ELMs often allow some W back into the confined plasma. Provided steady, high-power heating is maintained, this mechanism allows such plasmas to sustain high performance, with an average D–D neutron rate of ${\sim} 3.2 \times 10^{16}$ s ^−1 over a period of ∼3 s, after an initial overshoot (equivalent to a D–T fusion power of ∼9.4 MW), without an uncontrolled rise in W impurity radiation, giving added confidence that impurity screening by the pedestal may also occur in ITER, as has previously been predicted (Dux et al 2017 Nucl. Mater. Energy 12 28–35).

Experiments on the high field spherical tokamak ST40 have led to the recent observation of interplay between beam-driven modes of sweeping frequency (chirping modes) and transitions to the enhanced global confinement regime (H-mode) and back to the low confinement regime (L-mode). The H-modes of plasma confinement are identified from decreased intensity of D _α signal and from clear distinctions in the edge gradients of the visible plasma boundary (observed as a sharp plasma edge in camera images). The beam-driven chirping modes, identified as ideal magnetohydrodynamics beta-induced Alfvén acoustic eigenmodes modes, are observed in Mirnov coil signals, interferometry, and soft x-ray diagnostics. A moderate amplitude ‘primer’ chirping mode usually precedes an H–L transition. This is followed by a ‘dominant’ chirping mode with higher amplitude during the L-mode. The L–H transition back to the improved confinement occurs on a longer time scale of tens of ms, consistent with the slowing down time scale of fast beam ions. A dramatic decrease in toroidal plasma rotation is systematically observed associated with chirping modes sweeping down to zero frequency. Resonance maps built for the beam-driven chirping modes with the ASCOT (accelerated simulation of charged particle orbits in torodoial devices) code show that the resonant beam ions have orbits near the trapped-passing boundary. The ASCOT modelling assesses how losses of the resonant fast ions caused by the chirping modes with high enough amplitude modify the torque, potentially affecting the plasma rotation.