Divertor Detachment Research Articles

The effect of impurity seeding into the closed divertor on plasma detachment in the HL-2A tokamak

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 (). 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.

Multi-fluid and dynamic simulation of tungsten impurity in tokamak boundary plasma

Accumulation of tungsten (W) in core is a serious challenge for achieving high-performance plasmas in future tokamak reactors, thus W impurity transport is a highly concerned topic in the tokamak physics researches. Multi-fluid model and kinetic model are the numerical tools widely used for investigating and/or predicting impurity behaviors in the boundary of tokamak plasma. Generally, the applicability of multi-fluid model for impurity transport modeling requires that the collision mean-free-path should be smaller than the gradient scale lengths of particles, which may not be always satisfied. It is performed and comparatively investigated to evaluate the applicability of multi-fluid model for W impurity transport modeling, multi-fluid (SOLPS-ITER) modeling and kinetic (DIVIMP) modeling of W impurity transport in the edge of high-confinement plasma in Experimental Advanced Superconducting Tokamak (EAST) during neon impurity seeding. It is found that low-charge-state W ions are mainly located in the divertor region near the target plate where plasma collisionality is relatively high due to the relatively low/high local plasma temperature/density. Hence, the fluid assumption for transport of lowly-charged W ions can be well satisfied. Consequently, the density of lowly-charged W ions predicted by SOLPS-ITER and that calculated by DIVIMP are almost similar. Owing to the fact that the density of highly-charged W ions is relatively low and these particles mainly exist in the upstream (e.g. the main SOL and core) where plasma collisionality is relatively low, the fluid approximation cannot be well satisfied. However, the total W impurity density calculated by the kinetic code DIVIMP and the multi-fluid model SOLPS-ITER are found to be in agreement with each other within a factor of 1.5 for the simulation cases presented in this contribution. Besides, the multi-fluid simulation with bundled charge state model has also been performed, the obtained results are compared with those from the multi-fluid modeling with W ions treated as 74 fluids. It is revealed that in simulation cases with neon impurity seeding and with divertor plasmas in high-recycling or partially detached regimes, the bundling scheme, which is commonly used for saving the computation cost in multi-fluid modeling, tends to overestimate the average charge state of W ions and thus tends to underestimate the radiation power loss, especially in the divertor region. Consequently, under the circumstance that W impurity radiation dominates the radiative power loss in divertor region, plasma temperature/density can be largely overestimated/underestimated, leading to the underestimation of W ion ionization source and W impurity density. Moreover, simulation results demonstrate that W accumulation in core can decrease effectively during divertor detachment promoted by neon seeding.

Fascinating physics at the edge of magnetic fusion devices

The physics of the processes at the edge of magnetic fusion devices is multifaceted and exhibits complex, nonlinear synergistic effects. Even though this region occupies only a small portion of the whole device, it plays a crucial role in overall plasma confinement, heat exhaust, and plasma–wall interactions. The latter affects not only the performance but also the lifetime of plasma-facing components and therefore remains an outstanding challenge for future fusion reactors. At the edge of fusion devices, researchers are dealing with phenomena including classical and anomalous plasma transport, atomic physics effects, and physics of plasma-facing material under strong irradiation by particle and energy fluxes. The diversity of edge physics makes it particularly attractive for young scientists. Working in this field, they can find endless possibilities to demonstrate their talents and creativity. This short review describes just some of the basic scrape-off layer and divertor plasma phenomena including divertor plasma detachment, intermittent bursts of anomalous cross-field plasma transport, plasma–material interactions, and dust in fusion plasmas, which are of particular interest in relation to fusion reactors.

A Method to Identify the Heat Flux From Photons and Neutrals at the Divertor Target

The surface eroding thermocouple (SETC) is a diagnostic designed for fast surface heat flux measurement and has been successfully applied in DIII-D tokamak to investigate divertor detachment. A novel method has been developed of using a combination of surface-facing and recessed SETCs to distinguish between the heat flux from charged particles, and that from neutrals and radiative heating. This technique suggests that at least 20% of the total incident heat flux is attributable to the latter sources not trapped in the magnetic field in a fully detached open divertor condition. It not only indicates the noncharged particle heating significantly contributes to the total surface heat flux in detached divertor conditions but also demonstrates the feasibility of using two operational modes of SETCs to distinguish between the two surface heating mechanisms.

E × B flow driven electron temperature bifurcation in a closed slot divertor with ion B × ∇B away from the X-point in the DIII-D tokamak

An electron temperature bifurcation is observed in the small angle slot divertor, which has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Experiments in the DIII-D tokamak and associated SOLPS-ITER modeling with full drifts find a strong interplay between drifts and divertor geometry on divertor dissipation. The coupling of divertor geometry and drift flows can strongly affect the path towards divertor detachment onset as the plasma density is raised. With the strike point on the inner slanted surface and ion B × ∇B away from the magnetic X-point, bifurcative transitions were observed with sharp decrease of T e towards detachment onset both experimentally and computationally. This differs from the situation for the open divertor where the T e cliff was only observed for ion B × ∇B towards the X-point. SOLPS-ITER modeling with full drifts demonstrates that the magnitude of the E × B drift flow is comparable with the main plasma flow. The reversal of both the poloidal and radial E × B flows near the strike point leads to rapid density accumulation right near the separatrix, which results in bifurcative step transition of divertor conditions with cold plasma across the entire divertor target plate. These results indicate that the interplay between geometry and drifts should be fully taken into account in future fusion reactor divertor designs.

Divertor detachment of high β N H-mode discharges with the HL-2A closed divertor geometry

Divertor detachment of high H-mode discharges has been achieved by the injection of radiating impurities in the very closed divertor geometry of HL-2A. Non-intrinsic impurity is injected into the outer divertor chamber, and enhanced volumetric energy loss in the divertor and decreased divertor heat flux are then observed. Divertor detachment leads to a little degradation of the core confinement, characterized by slight decrease of plasma stored energy and pedestal pressure as well as increase of edge localized mode (ELM) frequency, but the high H-mode is still sustained with > 2 and H98 > 1. After impurity injection, the pedestal density fluctuation is increased, implying that the enhanced pedestal transport might be responsible for the degradation. During divertor detachment phase, the impurities are well controlled in the divertor without strong radiation losses near X-point region, and the main plasma density does not increase, but decreases slightly, which could be contributed by the very closed divertor geometry. These results are encouraging to achieve better heat load control and high plasma performance simultaneously for tokamak operations and to evaluate the effect of the divertor geometry for future tokamak devices.

Dependence of particle and power dissipation on divertor geometry and plasma shaping in DIII-D small-angle-slot divertor

• The compact SAS divertor in DIII-D provides a good testbed for understanding the effects of a tightly closed divertor on particle and power dissipation, and for application to core-edge integration solutions. • A longer outer leg facilitates the achievement of divertor dissipation in the SAS divertor, while a shorter leg leads to higher divertor temperatures and requires higher upstream densities to achieve the same level of divertor detachment. • The divertor conditions strongly depend on the divertor target shaping and Bt direction. • In-slot gas puffing can reduce detachment onset density by about 10% when the outer strike point is placed at the inner slanted surface. • Completed divertor detachment and a high confinement core with normalized energy confinement factor H 98 >1.0 can be simultanesouly achieved with the SAS divertor, demonstrating the benefit of a closed divertor for exploration of core-edge integration. Dedicated experiments in DIII-D find that magnetic shaping and divertor target geometry significantly affect the divertor plasma conditions and divertor detachment process in the small-angle-slot (SAS) divertor. The compact SAS divertor in DIII-D provides a good testbed for understanding the effects of a tightly closed divertor on particle and power dissipation, and for application to core-edge integration solutions. A longer outer leg facilitates the achievement of divertor dissipation in the SAS divertor, while a shorter leg leads to higher electron temperatures near the divertor target plate and requires higher upstream densities to achieve the same level of divertor detachment. In addition, with the ion B × ∇ B drift away from the SAS divertor and the outer strike point (OSP) near the outer corner, the target temperature is lower for a particular upstream density than with the OSP on a slanted or flat surface, leading to lower heat flux even when the particle flux remains similar. In contrast, with the ion B × ∇ B drift into the SAS divertor, a strike point at the inner slanted surface exhibits a lower upstream density to achieve divertor detachment than a strike point either at the outer corner or the outer slanted target. Experimental results and SOLPS-ITER simulations with full drifts suggest the strong interplay between drift flows and the neutral distribution resulted from target shaping. Furthermore, in-slot gas puffing has been shown to achieve global divertor detachment with an onset density about 10% lower than that using main-chamber gas puffing when the outer strike point is placed at the inner slanted surface. Corresponding modelling reveals that the local gas puffing enhances the neutral ionization which potentially facilitates the achievement of divertor dissipation. However, such improvement diminishes when the strike point is at the outer corner, which also indicates the geometric dependence on divertor performance in the SAS divertor. Even with different strike point locations, complete divertor detachment with very low particle and heat fluxes at the divertor targets and a high confinement core with normalized energy confinement factor H 98 >1.0 can be simultanesouly achieved with the SAS divertor with ion B × ∇ B drift into SAS divertor, demonstrating the benefit of a closed divertor for exploration of core-edge integration.

Control of tungsten impurity source and edge transport using different gas injection with full tungsten divertor on EAST

• With the newly installed EAST lower W divertor, W source and edge transport are experimentally investigated for different Ne and D 2 gas injection conditions. • W erosion on the whole divertor target can be well suppressed or mitigated by suitable edge Ne or D 2 injection when divertor detachment is achieved. • A W screening enhancement is obtained during the attached divertor condition due to D 2 injection. • After the divertor detachment, the core W content is observed to be kept in a high level, although W source from the divertor is strongly reduced. Tungsten (W) erosion and edge transport are investigated for EAST L-mode discharges with different gas injection. It is found that W erosion can be suppressed or mitigated by Ne or D 2 seeding when divertor detachment is achieved. Compared to edge D 2 fueling, Ne seeding from the divertor target is favorable for full detachment condition and thus W erosion suppression. Increasing the upstream plasma density by edge D 2 fueling can affect the divertor condition, which may lead to a W erosion mitigation. D 2 puffing at divertor target is less effective on increasing the upstream plasma density than OMP D 2 puffing, and thus it is less effective on reduction of W erosion rates. W gross erosion profiles with different amount of injected D 2 at the divertor are reproduced by a mixed material W erosion model, which indicates that there exist a Li-C overlayer on the W surface of EAST divertor. Ne and D 2 injection are also found to have different impact on the normalized core W density. For the attached divertor condition, divertor Ne seeding will increase W leakage, but a suitable D 2 fueling from divertor target can strengthen the edge W screening. After the divertor detachment, although the W source is dramatically reduced, the W core density is kept in a high level for the Ne seeding discharges, and even increased for the D 2 fueling discharges.

Data-driven model for divertor plasma detachment prediction

We present a fast and accurate data-driven surrogate model for divertor plasma detachment prediction leveraging the latent feature space concept in machine learning research. Our approach involves constructing and training two neural networks: an autoencoder that finds a proper latent space representation (LSR) of plasma state by compressing the multi-modal diagnostic measurements and a forward model using multi-layer perception (MLP) that projects a set of plasma control parameters to its corresponding LSR. By combining the forward model and the decoder network from autoencoder, this new data-driven surrogate model is able to predict a consistent set of diagnostic measurements based on a few plasma control parameters. In order to ensure that the crucial detachment physics is correctly captured, highly efficient 1D UEDGE model is used to generate training and validation data in this study. The benchmark between the data-driven surrogate model and UEDGE simulations shows that our surrogate model is capable of providing accurate detachment prediction (usually within a few per cent relative error margin) but with at least four orders of magnitude speed-up, indicating that performance-wise, it has the potential to facilitate integrated tokamak design and plasma control. Comparing with the widely used two-point model and/or two-point model formatting, the new data-driven model features additional detachment front prediction and can be easily extended to incorporate richer physics. This study demonstrates that the complicated divertor and scrape-off-layer plasma state has a low-dimensional representation in latent space. Understanding plasma dynamics in latent space and utilising this knowledge could open a new path for plasma control in magnetic fusion energy research.

Three-dimensional structure of radiative cooling in impurity seeded plasmas in the Large Helical Device

• Three-dimensionally localization of radiative cooling due to nitrogen seeding for divertor detachment was detected experimentally. • Principal component analysis was applied to two-dimensional radiation images. • A radiation feature in nitrogen seeding was extracted as a principal component. • EMC3-EIRENE code suggested the ionization enhancement at one of the divertor legs. • The extracted feature along the magnetic field lines which pass through the divertor leg. Three-dimensionally localization of radiative cooling due to nitrogen (N 2 ) seeding for divertor detachment was detected experimentally. Since the localization along some magnetic field lines induces toroidal asymmetry of heat load reduction on divertor plates, it should be avoided for fusion reactors. The three-dimensionally localized structure was extracted using Principal Component Analysis (PCA) from two-dimensional radiation images measured with an InfraRed imaging Video Bolometer (IRVB). By applying PCA to 34 images each in N 2 seeded plasmas with toroidally-asymmetric heat load reduction and in neon (Ne) seeded plasmas with toroidally-symmetric heat load reduction, a radiation feature in N 2 seeded plasmas was found as one of the principal components (PC). The three-dimensional transport code EMC3-EIRENE indicated that the ionization in one of the divertor legs is enhanced in nitrogen seeding compared with Ne seeding due to the difference in the first ionization energy. The magnetic field lines from the divertor leg were along the extracted radiation structure and were terminated by the divertor where the heat load decreased due to the N 2 seeding. These results indicate that three-dimensionally localized structure of radiative cooling was detected experimentally.

Turbulence in the near scrape-off layer towards the L-mode density limit in ASDEX-Upgrade

The L-mode density limit in ASDEX Upgrade is investigated with focus on the scrape-off layer. We observe a chain of events that involves the collapse of the E × B outer shear layer (just outside the separatrix), widening of the electron pressure fall-off length, outer divertor detachment, density shoulder formation, reduction and flattening of the parallel Mach number, and enhanced turbulent cross-field transport. The latter effect was associated with higher amplitude density fluctuations at low frequencies, where the turbulent cross-phase between density and velocity fluctuations pointing radially outward was favourable.

Evolution of the high-field-side radiation belts during the neon seeding plasma discharge in EAST tokamak

Divertor detachment achieved by injecting impurities or increasing density is always accompanied with various local radiation phenomena in the boundary or core plasma. This paper presents the formation and evolution of the high-field-side (HFS) radiation belts during the neon seeding plasma discharge in upper single null configuration with two directions of toroidal magnetic field in EAST tokamak. The neon mixed with deuterium seeding can induce the divertor detachment with strong radiation belts in the HFS scrape-off layer (SOL) region. With the increase of the radiation power, the plasma discharge will transit from H-mode to L-mode, and meanwhile the radiation belts move away from the near X-point to HFS SOL. When the radiation power is high enough, the radiation belts begin to move further to the other X-point along the HFS SOL, and even cause plasma disruption. The results indicate that the behavior of the radiation belts is related to the radiation power, plasma confinement performance and state of divertor detachment, which is useful for developing better feedback control methods to achieve high-performance radiative divertor operation mode.

Mitigation of plasma–wall interactions with low-Z powders in DIII-D high confinement plasmas

Experiments with low-Z powder injection in DIII-D high confinement discharges demonstrated increased divertor dissipation and detachment while maintaining good core energy confinement. Lithium (Li), boron (B), and boron nitride (BN) powders were injected in H-mode plasmas (I p = 1 MA, B t = 2 T, P NB = 6 MW, ⟨n e⟩ = 3.6–5.0 ⋅ 1019 m−3) into the upper small-angle slot divertor for 2 s intervals at constant rates of 3–204 mg s−1. The multi-species BN powders at a rate of 54 mg s−1 showed the most substantial increase in divertor neutral compression by more than an order of magnitude and lasting detachment with minor degradation of the stored magnetic energy W mhd by 5%. Rates of 204 mg s−1 of boron nitride powder further reduce edge localized mode-fluxes on the divertor but also cause a drop in confinement performance by 24% due to the onset of an n = 2 tearing mode. The application of powders also showed a substantial improvement of wall conditions manifesting in reduced wall fueling source and intrinsic carbon and oxygen content in response to the cumulative injection of non-recycling materials. The results suggest that low-Z powder injection, including mixed element compounds, is a promising new core-edge compatible technique that simultaneously enables divertor detachment and improves wall conditions during high confinement operation.

Visible imaging system with changeable field of view on the HL-2A tokamak.

A new visible imaging system characterizing a flexible optical design and delivering high resolution frames is established on the HL-2A tokamak. It features a modular configuration, consisting of a front-end imaging lens, a set of bilateral telecentric relay lenses, and a camera. To avoid the effects of plasma radiation (x and gamma-rays) and magnetic field variation on the camera, it should be away from the coils. Therefore, the length of the relay lenses determines the total size of the imaging system. The main feature of this imaging system is to realize the variation of field of view (FOV) by interchanging the front-end prime lenses or by using a zoom lens directly rather than designing the optical system afresh, which lowers the cost drastically. The primary purpose of varying FOV is to enrich the versatility of this system, i.e., focusing on a narrow FOV such as gas puff imaging or a wide FOV such as the plasma cross sections. During the HL-2A experiments, this visible imaging system is used to provide high quality pictures of the plasma-wall interaction, divertor detachment, pellet injections, and so on. The frames confirmed that a strong radiation close to the X point is correlated with the completely detached inner target.

Divertor plasma behaviors with neon seeding at different locations on EAST with ITER-like divertor

For the problem of excessively high divertor heat flux, active impurity seeding is an effective method to radiate the plasma energy reaching the divertor and thus achieve the divertor detachment. Neon is a very effective radiation impurity on many current tokamaks, which is also a candidate species to be applied on ITER. In the EAST 2019 experimental campaign, a series of experiments were performed by seeding a mixture of neon and deuterium (Ne-D2) for detachment and core-edge-divertor integration in H-mode plasmas. The divertor partial detachment with high-confinement core plasma has been achieved by using Ne-D2 seeding in EAST with ITER-like tungsten divertor. Both the plasma stored energy and H 98,y2 > 1.1 are maintained, with the divertor electron temperature, heat flux and the surface temperature near the strike point being all significantly reduced. The differences between Ne-D2 seeding at the scrape-off layer (SOL) upstream and downstream have been experimentally investigated in detail. It is found that impurity seeding at SOL downstream is more beneficial to reducing the divertor electron temperature and peak heat flux. By comparison with experiments using divertor D2 fueling, it is further demonstrated that gas seeding in the SOL downstream will enrich more particles near the strike point, while the seeding in the SOL upstream will influence the entire outer target more evenly. Furthermore, in most of the experiments, gas seeding does not cause obvious toroidal asymmetry in the divertor plasma. However, when D2 is injected in an amount similar to that used to build the plasma, it causes the particle flux near the gas-puff to increase locally, i.e., much more than that at the toroidal location far from the gas-puff location. It is a competition between particle source and transport. When the particle source is stronger, it will naturally increase the local particles. In addition, dedicated experiments with different poloidal distances between impurity seeding and strike point on the radiation ability were carried out. Both experimental results and SOLPS simulation show that the seeding close to the strike point is more conducive to neon ionization and energy radiation.

Experimental investigation of scrape-off layer blob high density transition in L-mode plasmas on EAST

Lithium Beam Emission Spectroscopy systems in the outer midplane and divertor Langmuir probe arrays embedded in the divertor target plates, are utilized to investigate the scrape-off layer (SOL) blob transition and its relation with divertor detachment on EAST. The blob transition in the near SOL is observed during the density ramp-up phase. When the plasma density, normalized to the Greenwald density limit, exceeds a threshold of f GW ∼ 0.5, the blob size and lifetime increases by 2 – 3 times, while the blob detection rate decreases by about 2 times. In addition, a weak density shoulder is observed in the near SOL region at the same density threshold. Further analysis indicates that the divertor detachment is highly correlated with the blob transition, and the density threshold of blob transition is consistent with that of the access to the outer divertor detachment. The potential physical mechanisms are discussed. These results could be useful for the understanding of plasma-wall interaction issues in future devices that will operate under a detached divertor and high density conditions (over the blob transition threshold).

Divertor detachment operation in helium plasmas with ITER-like tungsten divertor in EAST

Detachment in helium (He) discharges has been achieved in the EAST superconducting tokamak equipped with an ITER-like tungsten divertor. This paper presents the experimental observations of divertor detachment achieved by increasing the plasma density in He discharges. During density ramp-up, the particle flux shows a clear rollover, while the electron temperature around the outer strike point is decreasing simultaneously. The divertor detachment also exhibits a significant difference from that observed in comparable deuterium (D) discharges. The density threshold of detachment in the He plasma is higher than that in the D plasma for the same heating power, and increases with the heating power. Moreover, detachment assisted with neon (Ne) seeding was also performed in L- and H-mode plasmas, pointing to the direction for reducing the density threshold of detachment in He operation. However, excessive Ne seeding causes confinement degradation during the divertor detachment phase. The precise feedback control of impurity seeding will be performed in EAST to improve the compatibility of core plasma performance with divertor detachment for future high heating power operations.

Investigation of the role of hydrogen molecules in 1D simulation of divertor detachment

The role of neutral and charged hydrogenic molecules in detached regimes of tokamak plasmas is investigated using a simplified 1D numerical model. Using MAST Upgrade like conditions, simulations are implemented to study the rollover of target flux in upstream density scan. It is found that if and are considered in simulations a lower target temperature and a larger upstream density will be required to trigger divertor detachment under the same input power and particle flux, and the critical detachment threshold (the critical ratio of upstream static pressure to the power entering the recycling region) is found to be at rollover. Molecule–plasma interactions are found to be as crucial as atom–plasma interactions during divertor detachment, both of which account for the majority of plasma momentum loss in the cases studied here. Further analysis of the momentum loss decomposition shows molecule-plasma elastic collisions dominate molecule-plasma interactions, while molecular charge exchange cannot effectively reduce plasma momentum. In terms of emission, a strong rise of signal is found to be due to molecular excitation channels when the upstream density further increases after rollover.

Model for access and stability of the X-point radiator and the threshold for marfes in tokamak plasmas

Based on particle and energy balances, a reduced model is derived for the physical mechanisms leading to the occurrence of stable and unstable X-point radiators (XPRs), the latter also known as marfes. The leading roles of the neutral deuterium density in the divertor region for initiating XPRs is highlighted. An access condition is formulated whose parameter dependencies are consistent with experimental observations and which could also apply to the process of divertor detachment. With an exponential increase of the recombination rate at low temperature, the XPR becomes magnetohydrodynamically unstable, leading to a marfe and, possibly, to a disruption. A critical density for marfe occurrence is formulated with the upstream density and safety factor as leading parameters, as in the experiment. Marfes are predicted to be more likely in carbon devices than in impurity-seeded plasmas in tungsten devices. The edge plasma parameter domain where marfes occur resembles that used for active marfe avoidance schemes. Both the XPR and marfe occurrence parameter can be used to guide active discharge control.

Achievements of actively controlled divertor detachment compatible with sustained high confinement core in DIII-D and EAST

The compatibility of efficient divertor detachment with high-performance core plasma is vital to the development of magnetically controlled fusion energy. The joint research on the EAST and DIII-D tokamaks demonstrates successful integration of divertor detachment with excellent core plasma confinement quality, a milestone towards solving the critical plasma–wall-interaction (PWI) issue and core-edge integration for ITER and future reactors. In EAST, actively controlled partial detachment with T et,div ∼ 5 eV around the strike point and H 98 > 1 in different H-mode scenarios including the high β P H-mode scenario have been achieved with ITER-like tungsten divertor, by optimizing the detachment access condition and performing detailed experiments for core-edge integration. For active long-pulse detachment feedback control, a 30 s H-mode operation with detachment-control duration being 25 s has been successfully achieved in EAST. DIII-D has achieved actively controlled fully detached divertor with low plasma electron temperature (T et,div ⩽ 5 eV across the entire divertor target) and low particle flux (degree of detachment, DoD > 3), simultaneously with very high core performance (β N ∼3, β P > 2 and H 98 ∼ 1.5) in the high βP scenario being developed for ITER and future reactors. The high-β P high confinement scenario is characterized by an internal transport barrier (ITB) at large radius and a weak edge transport barrier (ETB, or pedestal), which are synergistically self-organized. Both the high-β P scenario and impurity seeding facilitate divertor detachment. The detachment access leads to the reduction of ETB, which facilitates the development of an even stronger ITB at large radius in the high β P scenario. Thus, this strong large radius ITB enables the core confinement improvement during detachment. These significant joint DIII-D and EAST advances on the compatibility of high confinement core and detached divertor show a great potential for achieving a high-performance core plasma suitable for long-pulse operation of fusion reactors with controllable steady-state PWIs.