Locally decodable codes (LDCs) are error correction codes that allow recovery of any single message symbol by probing only a small number of positions from the (possibly corrupted) codeword. Relaxed locally decodable codes (RLDCs) further allow the decoder to output a special failure symbol on a corrupted codeword. While known constructions of RLDCs achieve much better parameters than standard LDCs, it is intriguing to understand the relationship between LDCs and RLDCs. Separation results (i.e., the existence of q-query RLDCs that are not q-query LDCs) are known for q=3 (Gur, Minzer, Weissenberg, and Zheng, arXiv:2512.12960, 2025) and q ≥ 15 (Grigorescu, Kumar, Manohar, and Mon, arXiv:2511.02633, 2025), while any 2-query RLDC also gives a 2-query LDC (Block, Blocki, Cheng, Grigorescu, Li, Zheng, and Zhu, CCC 2023). In this work, we generalize and strengthen the main result in Grigorescu, Kumar, Manohar, and Mon (arXiv:2511.02633, 2025), by removing the requirement of linear codes. Specifically, we show that any q-query RLDC with soundness error below some threshold s(q) also yields a q-query LDC with comparable parameters. This holds even if the RLDC has imperfect completeness but with a non-adaptive decoder. Our results also extend to the setting of locally correctable codes (LCCs) and relaxed locally correctable codes (RLCCs). Using our results, we further derive improved lower bounds for arbitrary RLDCs and RLCCs, as well as probabilistically checkable proofs of proximity (PCPPs).

W/Cu flat-type components are promising candidates for plasma-facing components in future fusion devices due to their flexible and efficient heat sink design. They are expected to be used in the inner/outer Reflector plates and Dome of ITER. To investigate their damage behaviors, a high-resolution infrared camera was employed to monitor the damaged C4 W/Cu flat-type component with a hypervapotron heat sink, located in the outer horizontal target (OHT) of the EAST lower divertor. The results show that the formation of W protrusions (‘Hills’) connects the early crack initiation stage with the final W exfoliation. To prevent further damage progression, it is recommended that for tokamaks, the damaged area be located two decay lengths away from the strike point—this distance is sufficient to guarantee safety. In such a scenario, if the damage lies within the private flux region, the parallel heat flux will be reduced to nearly zero. Although thermal performance degraded over time, the damaged OHT component was still able to maintain an approximate thermal equilibrium during each discharge, and no abnormal W impurity levels were observed. Linear fitting analysis estimates a total lifetime of approximately 3257 ± 277 shots, with a plasma stored energy of 298 ± 20 kJ corresponding to the onset of W melting in the surface protrusions. These findings shed light on the damage behavior of W/Cu flat-type components, provide key support for the development and validation of finite element analysis models, and offer practical guidance for the design and evaluation of ITER-like W/Cu flat-type components.

Recent experiments in EAST have demonstrated the compatibility of tungsten (W) as main wall plasma facing material with high confinement H-mode with low or no boron coverage. The experiments were conducted in plasmas with $q_{95}$ $\approx$ 6.0, which allowed access to both type-I and type-II ELMy H-modes in EAST with a W wall and low normalized input torque similar to ITER. Central electron cyclotron (EC), neutral beam injection (NBI) and lower hybrid waves (LHW) were applied as auxiliary heating in a range of total power injected into the plasma from 3 MW to 5 MW. Small amplitude high frequency type-II ELMs allow maintaining good H-mode energy confinement even when the distance between the separatrix and the main W limiter is as low as 4 cm. The normalized H-mode energy confinement can reach $H_{98}$ factors up to 1.1 for both EC+NBI and EC+LHW power combination in the type-II ELMy H-mode regime. In addition, for these type-II ELMy H-mode conditions, nitrogen puffing from divertor region has been used to achieve partial detachment without significant impact on $H_{98}$ nor on the core W concentration. On the contrary, EAST operation in the type-I ELMy H-mode is strongly affected by the main wall W source. The use of $n$ = 2 resonant magnetic perturbations (RMPs) to achieve type-I ELM suppression in EAST reduces the core W level but at the cost of about 10% reduction on energy confinement. These results complement existing W-wall findings, providing a preliminary foundation from EAST for evaluating the impact of the W first wall in ITER, as proposed in the new ITER baseline, and highlight that the achievement of sufficient ELM control levels with low deterioration of energy confinement is key to minimize this impact.

Neoclassical tearing modes (NTMs) are critical magneto-hydrodynamic (MHD) instabilities in tokamak plasmas, significantly impacting plasma confinement and potentially leading to disruptions. In the Experimental Advanced Superconducting Tokamak (EAST) with boronization wall conditions, the repetitive bursting phenomena of NTMs are observed experimentally in the high confinement mode plasma with metal impurity accumulation. The evolution characteristics of NTM behaviors are investigated in the article. The m/n = 2/1 NTMs are excited by mode couple with ideal kink modes. The NTMs exhibited two distinct growth phases: a rapid growth phase (phase A) with small saturated amplitude and a secondary growth phase (phase B) characterized by a slow growth rate but larger saturated amplitude. Plasma confinement is enhanced during NTM phase A but degraded in phase B. The phase B of NTMs can lead to fast electron loss and H–L transitions. In addition, the NTMs are observed to periodically recur on a timescale of approximately 200 ms. It is found that the high-Z impurity radiation is also modulated synchronously by the long-period NTM behavior. A strong radiation region near the m/n = 2/1 magnetic island develops in NTM phase B due to outward impurity transport and the radiation profile becomes hollow. The effect of metal impurity radiation on NTM dynamics is worth further investigating for fusion devices with full tungsten wall like ITER.

Impurity seeding is typically used to achieve plasma detachment. In this study, nitrogen (N 2 ) seeding detachment experiments are performed on EAST in high-confinement mode (H-mode) plasmas. The experiments are conducted under full-metal-wall conditions, featuring a tungsten divertor and boronized first wall surfaces. The radiation characteristics near the lower X-point region during plasma detachment are systematically investigated. The formation of the X-point radiation region is observed, primarily contributed by the radiation from seeded impurities. The X-point radiation intensity is related to the N 2 seeding rates and the degree of detachment. During the partial detachment with low N 2 seeding rates, the X-point radiation remains weak. However, as the N 2 seeding rate increases, a strong X-point radiation region emerges along with the deep detachment process. The strong X-point radiation can be a direct consequence of deep divertor detachment. When the amount of N 2 injected is excessive, the strong X-point radiation region can expand upwards toward the bulk plasma region, leading to modifications of the plasma parameter profiles. Nevertheless, the upward expansion of the strong X-point radiation region does not adversely affect the sustained deep detachment state and remains compatible with H-mode confinement, with only a slight confinement degradation of no more than 10%.

Measurements of the total radiated power and its spatial distribution are crucial for fusion research. On the experimental advanced superconducting tokamak (EAST), both the metal foil resistive bolometer and the absolute extreme ultraviolet (AXUV) photodiodes have been used to quantify the radiated power. This article introduces the latest improvement of the bolometer diagnostic system on EAST. It also details the successful design and installation of new divertor AXUV cameras, which are dedicated to the investigation of divertor physics. The shielding components of the bolometer detector have been refined, and the article provides a detailed exposition of the double shielding structures that have been verified as effective in microwave shielding. Additionally, the changes in the radiated power distribution in the divertor region during the plasma detachment process are measured using the divertor AXUV camera. Finally, the radiated power measured by the AXUV detector and metal foil resistive bolometer are compared, and different detector performances are presented.

Radiative divertor is an effective method for the divertor heat flux control, but excessive core impurity radiation can deteriorate the confinement in tokamaks. In recent EAST experiments, the compatibility of divertor detachment and impurity control with good core confinement (H98,y2 ∼1) has been achieved simultaneously by combining an upstream deuterium (D2) puff and divertor argon (Ar) seeding. Both experimental and SOLPS-ITER modeling reveal that additional D2 puff can further mitigate the target heat load and facilitate the detachment. The SOLPS-ITER results also show that the D2 puff makes a limited contribution to the Ar line radiation but a significant contribution to the neutral radiation in the SOL and divertor regions. The physical mechanism behind the effect of the D2 puff on the Ar retention is also revealed by the modeling. The increase in the Ar velocity is the major reason for the improved Ar retention in the partial detachment state. The increase in vAr is mainly caused by the strengthening of frictional drag of D+, which is because the additional D2 puff increases the pressure gradient force (FPG) and the electric field force (FE) on D+. Further analysis reveals that in the deep detachment state, some other negative effects play a dominant role in determining the Ar retention, such as the shift of the region of the Ar+ ionization source, the decrease in the strength of the Ar+ ionization source, and the net force on the Ar ions near the target.

Since the last IAEA-FEC in 2021, significant progress on the development of long pulse steady state scenario and its related key physics and technologies have been achieved, including the reproducible 403 s long-pulse steady-state H-mode plasma with pure radio frequency (RF) power heating. A thousand-second time scale (similar to 1056 s) fully non-inductive plasma with high injected energy up to 1.73 GJ has also been achieved. The EAST operational regime of high beta(P) has been significantly extended (H-98y2 > 1.3, beta(P) similar to 4.0, beta(N) similar to 2.4 and n(e)/n(GW) similar to 1.0) using RF and neutral beam injection (NBI). The full edge localized mode suppression using the n = 4 resonant magnetic perturbations has been achieved in ITER-like standard type-I ELMy H-mode plasmas with q(95) approximate to 3.1 on EAST, extrapolating favorably to the ITER baseline scenario. The sustained large ELM control and stable partial detachment have been achieved with Ne seeding. The underlying physics of plasma-beta effect for error field penetration, where toroidal effect dominates, is disclosed by comparing the results in cylindrical theory and MARS-Q simulation in EAST. Breakdown and plasma initiation at low toroidal electric fields (<0.3 V m(-1)) with EC pre-ionization is developed. A beneficial role on the lower hybrid wave injection to control the tungsten concentration in the NBI discharge is observed for the first time in EAST suggesting a potential way toward steady-state H-mode NBI operation.

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 similar to 1 to similar 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.

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.

Massive gas injection(MGI)is a traditional plasma disruption mitigation method.This method directly injected mas-sive gas into the pre-disruption plasma and had been developed on the Experimental Advanced Superconducting Tokamak(EAST).Different noble gas injection experiments,including He,Ne,and Ar,were performed to compare the mitigation effect of plasma disruption by evaluating the key parameters such as flight time,pre-thermal quench(pre-TQ),and current quench(CQ).The flight time was shorter for low atomic number(Z)gas,and the decrease in flight time by increasing the amount of gas was insignificant.However,both pre-TQ and CQ durations decreased considerably with the increase in gas injection amount.The effect of atomic mass on pre-TQ and CQ durations showed the opposite trend.The observed trend could help in controlling CQ duration in a reasonable area.Moreover,the analysis of radiation distribution with different impurity injections indicated that low Z impurity could reduce the asymmetry of radiation,which is valuable in mitigating plasma disruption.These results provided essential data support for plasma disruption mitigation on EAST and future fusion devices.

To better understand divertor detachment and asymmetry in the Experimental Advanced Superconducting Tokamak (EAST), drift modeling via the comprehensive edge plasma code SOLPS-ITER of neon impurity seeded plasmas in favorable/unfavorable toroidal magnetic field ( B T ) has been performed. Firstly, electrostatic potential/field ( ϕ / E ) distribution has been analyzed, to make sure that ϕ and E are correctly described and to better understand drift-driven processes. After that, drift effects on divertor detachment and asymmetry have been focused on. In accordance with the corresponding experimental observations, simulation results demonstrate that in favorable B T the onset of detachment is highly asymmetric between the inner and outer divertors; and reversing B T can significantly decrease the magnitude of in-out asymmetry in the onset of detachment, physics reasons for which have been explored. It is found that, apart from the well-known E × B drift particle flow from one divertor to the other through the private flux region, scrape-off layer (SOL) heat flow, which is much more asymmetrically distributed between the high field side and low field side for favorable B T than that for unfavorable B T , is also a critical parameter affecting divertor detachment and asymmetry. During detachment, upstream pressure ( P u ) reduction occurs and tends to be more dramatical in the colder side than that in the hotter side. The convective SOL heat flow, emerging due to in-out asymmetry in P u reduction, is found to be critical for understanding divertor detachment and asymmetry observed in EAST. To better understand the calculated drastic power radiation in the core and upstream SOL, drift effects on divertor leakage/retention of neon in EAST with both B T directions have been addressed for the first time, by analyzing profile of poloidal neon velocity and that of neon ionization source from atoms. This work can be a reference for future numeric simulations performed more closely related to experimental regimes.

In this study, the multifaceted asymmetric radiation from the edge (MARFE) was observed in high-density discharges or during impurity gas injection on the EAST tokamak. The MARFE onset indicated by spectral and radiation signals can also be detected by the POlarimeter-INTerferometer (POINT) diagnostic, which measures the horizontal line-integrated density and the Faraday rotation. The fluctuation amplitude of the density signal resulting from the MARFE oscillation increases with the edge safety factor, which is consistent with the thermal instability theory. By combining density and the Faraday rotation, the local current shrinkage in the MARFE region is observed during the MARFE movement. The density and the current profile calculated by the POINT become more peak during the MARFE, which may lead to a strong magnetohydrodynamic instability that can result in disruption.

The multifaceted asymmetric radiation from the edge (MARFE), which is generally considered to be the result of a radiation thermal instability in the edge and usually occurs in high density operation, has been first observed to move up and down along the poloidal cross-section due to edge cooling after cryogenic deuterium pellet injections in EAST tokamak with tungsten divertor. A maximum electron density of 0.84 × n GW has been obtained using continuous cryogenic pellet fueling. In the meantime, MARFEs, initially located near the divertors of EAST, moves to the inner wall on high field side after each pellet injection. This movement should be attributed to the asymmetry of the power flow to the two sides of the MARFE after pellet injections. Accompanied with MARFE movement, two kinds of strong density fluctuations have been observed. The ones with continuous and regular frequency spectrum, which does not cause a reduction of main plasma density, are confirmed to be induced by MARFE. The others, appearing with magnetic fluctuations, have been determined to be induced by the m / n = 2/1 magnetohydrodynamic activities after pellet injections. All the investigations in this paper will be meaningful for the steady high density operation of future fusion reactors, such as ITER.

In this paper, a comparison of dynamical features between the fast H-L and the H-I-L transition, which can be identified by the intermediate phase, or ‘I-phase’, has been made for radio-frequency (RF) heated deuterium plasmas in EAST. The fast H-L transition is characterized by a rapid release of stored energy during the transition transient, while the H-I-L transition exhibits a ‘soft’ H-mode termination. One important distinction between the transitions has been observed by dedicated probe measurements slightly inside the separatrix, with respect to the radial gradient of the floating potential, which corresponds to the E × B flow and/or the electron temperature gradient. The potential gradient inside the separatrix oscillates and persists during the stationary I-phase, and shows a larger amplitude than that before the fast H-L transition. The reduction of the gradient leads to the final transition to the L-mode for both the fast H-L and the H-I-L transition. These findings indicate that the mean E × B flow shear and/or edge electron temperature gradient play a critical role underlying the H-L transition physics. In addition, the back transition in EAST is found to be sensitive to magnetic configuration, where the vertical configuration, i.e., inner strike-point located at vertical target, favours access to the H-I-L transition, while the horizontal shape facilitates achievement of the fast H-L transition. The divertor recycling level normalized to electron density is higher before the fast H-L transition, as compared to that before the I-phase, which strongly suggest that the density of the recycled neutrals is an important ingredient in determining the back transition behaviour.

The influence of a molybdenum dust buildup on plasma edge turbulence has been studied in the EAST tokamak. The motion of the dust from the upper divertor region is detected by a fast visible CCD camera, the XUV spectrometer arrays, and the EUV spectrometer. The MoXV emission intensity sharply increases compared with the spectral lines of various ionization states of other elements, which implies that the dust particles are the molybdenum impurities. The radial distribution of Mo14+ ion simulated by a simplified 1D transport model indicates that the molybdenum dust mainly deposits in the pedestal bottom region. Moreover, it is observed that the coherent mode (CM) appears at ρ = 0.94 after the molybdenum impurities enter the main plasma region. The influx of molybdenum impurities results in increasing pedestal electron density and decreasing pedestal electron temperature in contrast to that before the event of impurities dropping. It is also found that the electron density gradient in the pedestal increases when the ablation of the molybdenum impurities is observed in the pedestal region. The qualitative experimental results indicate that the onset of CM is likely related to the increase of the density gradient and edge collisionality in the pedestal. In comparison to the density gradient, the enhancement of CM amplitude largely depends on the increase of the edge collisionality.

Modulated lower hybrid waves, are injected into the Experimental Advanced Superconducting Tokamak (EAST), to investigate the modulated evolutions of pedestal structure and unstable edge-localized modes (ELMs). The spectrum patterns of D-alpha show that the ELM amplitude only slightly decreases while the ELM frequency becomes almost 3-5 times larger as the lower hybrid radio frequency (LHRF) jumps into the lower power level. Pedestal structure analysis indicates that higher LHRF power injection can increase the plasma stored energy and core plasma parameters, accompanying with the pedestal shift and gradient increase. Ion saturation contour current plots suggest that LHRF power can deposit in the plasma edge region to profoundly change the three-dimensional (3D) magnetic topology, similarly to the effect of resonant magnetic perturbations (RMPs), by inducing the helical current filaments (HCFs) following along the magnetic field lines in the scrape-off layer (SOL), which may cause the density pump-out effect by increasing the radial plasma transport. The modulated higher temperature and lower density of pedestal top-values can reduce the pedestal collisionality, which will strongly increase the bootstrap current density and gradually shift the Peeling-Ballooning mode (PBM) boundary. ELITE code stability analysis indicates that LHRF injection can lead to the relatively unstable ELMs with a higher pedestal gradient, simultaneously accompanying with the normalized pressure gradient and edge bootstrap current density increase. All of these results imply that there is a significant correlation between the ELM behavior and pedestal structure modulation because of LHRF dissipation at the plasma edge for the unstable ELM modulated case.

The phenomenon of multifaceted asymmetric radiation from the edge (MARFE) is investigated with impurity gas puff from the upper divertor on the Experimental Advanced Superconducting Tokamak (EAST). A typical process in which dense radiation region in the vicinity of the X point (X-point MARFE) further evolves into MARFE on the high field side (wall MARFE) after the plasma makes a transition from H-mode to L-mode confinement is observed in discharges with the impurity gas seeding. The electron density distribution and evolution in the divertor volume are measured by means of spectroscopy. It is observed in the experiments that the final position of MARFE is related to the ion backward difference B drift direction. After the phase difference of n = 1 resonant magnetic perturbation (RMP) change from 90 degrees to 270 degrees, MARFE begins to appear on the high field side, which may be caused by the density pump-out induced by RMP. At the same heating power, the density threshold for MARFE formation is somewhat higher in the Ne seeding discharge than in the Ar seeding discharge. This may be attributed to the fact that the divertor radiation fraction, P-rad,P-div/P-rad,P-main, in the Ne seeding discharge is lower than that in the Ar seeding discharge. In addition, MARFE is successfully suppressed by total radiated energy feedback control in radiative divertor experiments. Therefore, radiation feedback control may play a vital role in avoiding major disruption induced by impurity radiation in radiative divertor experiments.

A hybrid scenario with the minimum safety factor above unity and a safety factor profile with a wide range of low magnetic shear in the core is one of the candidates for ITER operations. Theory predicts that such equilibria are prone to quasi-interchange mode, which was confirmed by observations from several devices. The presence of those 3D quasi-interchange helical modes has both positive and negative effects on plasma performance. On the one hand, they can avoid low mode number instabilities such as sawteeth. On the other hand, they would lead to confinement degradation, significant fast ion losses and toroidal rotation damping. Such modes are observed for the first time in EAST high beta(p) hybrid plasmas achieved with off-axis neutral beam injection (NBI) heating. The mode manifested itself as long-lived saturated helical instability. Notably, when the mode was destabilized, the toroidal rotation profile in the core was completely flattened. The extent of the flattened rotation profile reaches up to half of the minor radius. As the minimum safety factor qmin approaches unity, strong n = 2 and n = 3 harmonics are successively driven unstable as nonlinear consequences of the n = 1 mode. A 3D resistive nonlinear simulation with a realistic tokamak configuration has been applied to reproduce the evolution of long-lived helical instability with the M3D code. The simulated magnetic perturbation shows a strong n = 2 component. As the mode grows, the amplitude of the n = 2 harmonic grows immensely with respect to that of the n = 1 harmonic. The n = 2 harmonic gradually becomes dominant as the mode saturates. In saturation, the initial weak shear safety factor profile in the core becomes completely flat. This change of safety factor profile in the core corroborates a self-regulating magnetic flux pumping mechanism which is considered responsible for maintaining stationary non-sawtoothing hybrid discharges.