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All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Results from the last DD and DT JET campaigns in the framework of the EUROfusion Tokamak Exploitation Work Package activity

JET, the only tokamak capable of operating with deuterium–tritium (D–T) fuel (since TFTR was shutdown in 1999), has provided essential experimental data to support ITER and DEMO design and operation. Within the EUROfusion Tokamak Exploitation Work Package, JET completed its final campaigns (2022–2023), culminating in the third D–T campaign (DTE3). These experiments addressed key challenges in plasma scenarios, exhaust control, and tritium management under reactor-relevant conditions. Significant progress was achieved in demonstrating ITER-like integrated scenarios with impurity seeding, achieving partial divertor detachment and high confinement ($H_{98}(y,2)$ ≈ 0.85) at 3 MA in D–T plasmas. Advanced exhaust regimes such as quasi-continuous exhaust (QCE) and X-point radiator (XPR) were successfully achieved first in D–D and then extended to D–T operation, confirming their relevance for mixed isotope operation. Operational milestones included a new world record of 69 MJ fusion energy in tritium-rich hybrid plasmas and long-pulse H-mode operation up to 60 s, contributing with unique data to the CICLOP database. Physics studies focused on peeling-limited pedestals in support of ITER and improved understanding of edge stability and impurity screening in metallic environments. Extensive usage of the shattered pellet injector (SPI) on JET provided critical information for the design of the ITER disruption mitigation system (DMS). Real-time control systems for D/T ratio control and plasma exhaust were deployed and demonstrated in D–D and D–T, while energetic particle physics investigations unfolded the role of fast ions in turbulence suppression mechanisms. Comprehensive tritium retention studies using gas balance method, post-mortem analysis, and ITER-relevant laser induced desorption spectroscopy (LIDS) diagnostics provided essential input for tritium accountancy strategies. These results are validating the ITER operational concepts, inform DEMO design, and deliver critical experience in nuclear operation and scenario integration.

disruptions↗
DOE OSTI2024

Examining Graphite Degradation in Molten Salt Environments: A Chemical, Physical, and Material Analysis

Molten-salt reactors (MSRs) are Generation IV nuclear reactors that use liquid salt as a coolant and/or fuel. In several MSR designs, graphite serves as a moderator and/or reflector. However, due to limited experimental data and operational experience, our understanding of graphite behavior in molten salt environments remains incomplete. This report aims to identify the degradation mechanisms of nuclear graphite in MSRs, detail the mechanisms of each factor, and provide an initial assessment of their impact on the structural integrity of graphite components. This assessment is based on an extensive literature review and insights from subject matter experts. Furthermore, given the limited data, a modeling strategy using existing Grizzly software is proposed for a more thorough analysis where appropriate. Additionally, it presents mitigation strategies where applicable. The report covers physical degradation mechanisms such as infiltration, erosion, and abrasion, as well as chemical degradation mechanisms including fluorination, intercalation, corrosion, and oxidation. Molten salt can infiltrate the porous structure of graphite, leading to several detrimental effects. Entrapment of fissile products within the graphite pores can cause radiation damage and could pose challenges in the handling and disposal of contaminated components. The differential thermal expansion between the infiltrated salt and graphite, along with internal stress from pressurized molten salt and volumetric heating, can compromise the structural integrity of graphite. To mitigate these effects, employing ultra-fine graphite grades and applying sealants and coatings are effective strategies. A computational model based on coupled solid mechanics and heat transfer phenomena could be used to predict the internal stresses using Grizzly software. In pebble-bed MSRs, graphite fuel pebbles can cause abrasion against reactor components due to friction and wear. The severity of wear is influenced by various factors such as temperature, environment, and the presence of lubricants. Tribological studies reveal that higher temperatures and molten salt environments, such as FLiBe, significantly reduce wear rates compared to dry conditions. Additionally, the chemical composition of the salt can further optimize graphite's tribological performance. Long-term wear effects can be modeled by incorporating surface defects into the geometry and predict stresses under thermal and radiation effects using Grizzly software. Chemical degradation of graphite in a molten salt environment can occur through fluorination and intercalation. Fluorination can occur via replacement of hydrogen or oxygen atoms, or at the active sites, but does not cause structural degradation. Intercalation, on the other hand, can lead to exfoliation, where layers of graphite separate and peel away, damaging the graphite. Protective coatings can enhance graphite's resistance to intercalation. Graphite generally exhibits good chemical stability in molten salt environments, though it can corrode under specific conditions, particularly in the presence of impurities or oxidants. Studies have shown that protective coatings, such as plasma-sprayed partially stabilized zirconia (PSZ), can effectively prevent such degradation. Corrosion behavior varies significantly with different graphite grades and coating applications, underscoring the need for detailed studies on uncoated and coated graphite to understand and mitigate corrosion mechanisms in MSRs. Research indicates that the presence of oxidants and impurities can accelerate graphite degradation in molten salts, making it essential to explore acceptable impurity limits. Oxidation is another critical degradation mechanism, leading to weight loss and structural damage due to the formation of CO and CO 2 from the reaction of carbon atoms with oxygen. This process creates new porosity and compromises graphite's integrity. While extensive research on graphite oxidation has been conducted for gas-cooled reactors, studies specific to MSRs are limited. Findings from the coal industry suggest that molten alkali metal salts can significantly accelerate graphite oxidation, a hypothesis worth exploring for fluoride salts in MSRs. Understanding oxidation behavior in MSRs is vital for developing protective measures. The analysis of post-irradiated graphite from the MSRE experiment demonstrated exceptional chemical compatibility with molten fluoride salt, suggesting that the extent of chemical attack on graphite largely depends on the salt's infiltration capability. Therefore, the use of ultra-fine grade graphite could help mitigate chemical degradation effects. Existing oxidation modeling capabilities in Grizzly, which use reaction-diffusion equations to model graphite-air interactions, could be adapted to simulate the chemical degradation effects of graphite in molten salt environments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Simulations of divertor plasmas with inverse sheaths

The effect of strong electron emission from material surfaces has been proposed to form an “inverse sheath”: a region with a positive potential relative to the near-wall plasma which prevents the flow of ions to the wall [M. D. Campanell, “Negative plasma potential relative to electronemitting surfaces,” Phys. Rev. E. 88, 033103 (2013); M. D. Campanell and M. V. Umansky, “Strongly emitting surfaces unable to float below plasma potential,” Phys. Rev. Lett. 116, 1–5 (2016); M. D. Campanell and G. R. Johnson, “Thermionic cooling of the target plasma to a sub-ev temperature,” Phys. Rev. Lett. 122, 1–5 (2019)]. We assess the viability of this regime in a tokamak device using the 2D edge plasma transport code UEDGE [T. Rognlien et al ., “A fully implicit, time dependent 2-D fluid code for modeling tokamak edge plasmas,” J. Nucl. Mater. 196–198, 347–351 (1992)]. Since the UEDGE code does not consider the sheath region directly, we apply boundary conditions at the divertor targets which emulate the physics of both “standard” and “inverse” sheath regimes [R. Masline et al ., “Influence of the inverse sheath on divertor plasma performance in tokamak edge plasma simulations,” Contrib. Plasma Phys. 60, e201900097 (2020)]. Using these boundary conditions, we perform scoping studies to assess plasma parameters near the target by varying the density at the core-edge interface. We observe a smooth transition in the resultant profiles of plasma parameters for the standard sheath, and a bifurcation across the simulation set for plasmas with an inverse sheath. The cause of this bifurcation is assessed by performing the parameter scan both with and without impurity radiation; we observe that the bifurcation persists in both cases, indicating that this bifurcation is caused by plasma recombination.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Electric field effects during disruptions

Tokamak disruptions are associated with breaking magnetic surfaces, which makes magnetic field lines chaotic in large regions of the plasma. The enforcement of quasi-neutrality in a region of chaotic field lines requires an electric potential that has both short and long correlation distances across the magnetic field lines. The short correlation distances produce a Bohm-like diffusion coefficient ∼Te/eB and the long correlation distances aT produce a large scale flow ∼Te/eBaT. This cross-field diffusion and flow are important for sweeping impurities into the core of a disrupting tokamak. The analysis separates the electric field in a plasma into the sum of a divergence-free, E→B, and a curl-free, E→q, part, a Helmholtz decomposition. The divergence-free part of E→ determines the evolution of the magnetic field. The curl-free part enforces quasi-neutrality, E→q=−∇→Φq. Magnetic helicity evolution gives the required boundary condition for a unique Helmholtz decomposition and an unfortunate constraint on steady-state tokamak maintenance.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Study of solid molecular deuterium D 2 growth under gas pressure

The injection of high-speed cryogenic pellets made of frozen hydrogen-isotopes, represents to date the most effective method to fuel magnetically confined thermonuclear fusion plasmas. Additionally, the injection of very large pellets composed of cryogenic solid of some suitable impurity (typically a noble-gas such as H 2 , Ne, or H 2 /Ne, D 2 /Ne mixtures), shattered in relatively small fragments just before entering the plasma, seems to be the most promising method to reduce the damage risks for the plasma-facing components in case of a plasma disruption. This technology, known as "Shattered Pellet Injection" (SPI), allows to spread out the plasma energy and mitigate possible damage to the in-vessel components, as well as to densify the plasma to suppress the formation of runaway electrons, and/or dissipate their energy. Several techniques to produce and launch cryogenic pellets have been investigated in the past decades. "Pipe gun" injectors are reliable and relatively simple devices are still commonly used today. They make use of single- or two-stage pneumatic light-gas guns to accelerate the pellet at high speeds. In these injectors, the cryogenic pellets are formed “ in situ ” (i.e., inside the launching barrel), by de-sublimating them directly from the gas phase, i.e., at temperatures and pressures below those of the triple point. The simplest case is pure deuterium pellets ($\mathcal{T}$ < 18.7 K, $\mathcal{P}$ < 171.3 hPa). The production of good quality solid deuterium, capable of withstanding the mechanical stress during the acceleration of the pellets, is a key issue. To this end the phase transition of deuterium from gas to solid (and vice versa) is modeled with extensive molecular-dynamics (MD) simulations. Furthermore, the solid growth from the gas phase is simulated in an ample range of temperatures and pressures, to find the best compromise between growth velocity and mechanical properties of the resulting solid system.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Exploration of ICP platforms for measuring elemental impurities in uranium ore concentrates

Presented here is a look at different inductively coupled plasma (ICP) platforms including sector field, triple quadrupole mass spectrometric configurations, and optical emission spectroscopy. The goal of this paper was to convey the feasibility of trace elemental analysis of a uranium ore concentrate with varying ICP-based platforms. It was predicted that low mass elements would be better determined through optical means, or the mass collision/reaction technology of the triple quadrupole mass spectrometer, in comparison to high resolution measurements of magnetic sector-based configurations. Ultimately, it is difficult to achieve an “optimal” analysis of 20+ elements, for numerous samples, on a single platform, especially when time, measurement reproducibility, sensitivity, and robustness should be considered. The work presented here highlights the advantages and disadvantages of the respective trace elemental analysis platforms.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Characterizing W sources in the all-W wall, all-RF WEST tokamak environment

In this work, experimental data, together with interpretive modeling tools, are examined to study trends in the tungsten (W) source in the all-W environment of the WEST tokamak, both from the divertor and from the main chamber. In particular, a poloidal limiter protecting an ion cyclotron resonance heating (ICRH) antenna is used as proxy for main chamber sourcing. The key study is carried out by stepping up lower hybrid current drive (LHCD) power, as the only auxiliary power source. Limiter and divertor W sources exhibit a qualitatively similar proportionality to the total power crossing the separatrix, P SEP , although the main chamber source remains substantially lower than the divertor source, for the range of P SEP accessible in the experiments. Intepretive modeling of the limiter source is carried out with a particle-in-cell (PIC) sheath model coupled to a surface sputtering model. Oxygen is used as a proxy for all light impurity species allowing for characterization of the critical W erosion regions. To get a good quantitative match to the data, it is necessary to assume that the oxygen arrives at the surface mostly at high ionization stages (4+ and above). A separate simulation with SOLEDGE-EIRENE, constrained to measured upstream scrape-off-layer plasma profiles, gives oxygen fractional abundances that are compatible with the PIC simulation result. This is understood to arise from transport processes that dominate over recombination. Substituting the LHCD by ICRH, in an equivalent experiment, the local W source exhibits a 3× enhancement. This can be matched by the simulation, by assuming local RF electric field rectification, based on ~100 eV peak-to-peak, near-antennna electric field. This work has highlighted the particular importance of understanding the ion charge state balance of light impurities as these are most likely the dominant sputtering species in fusion devices with high-Z walls.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Integrated modeling of RF-induced tungsten erosion at ICRH antenna structures in the WEST tokamak *

This paper introduces STRIPE (Simulated Transport of RF Impurity Production and Emission), an advanced modeling framework developed to analyze material erosion and the global transport of eroded impurities originating from radio-frequency (RF) antenna structures in magnetic confinement fusion devices. STRIPE integrates multiple physics modules: SolEdge3x for scrape-off-layer plasma profiles, COMSOL for 3D RF rectified sheath potentials, RustBCA for erosion yields and surface interactions, and global impurity transport for 3D ion energy-angle distributions and impurity transport. The framework is applied to an ion cyclotron RF-heated L-mode discharge (#57877) in the WEST tokamak, where it predicts a thirty-fold increase in gross tungsten erosion at antenna limiters during the transition from ohmic to ICRH operation. Additionally, under ICRH conditions, a tenfold enhancement in erosion is observed when comparing RF sheath effects to purely thermal sheath conditions. High-charge-state oxygen ions ($\mathrm{O}$ 6+ and above) are identified as the dominant contributors to tungsten sputtering. To validate the model, a synthetic diagnostic tool based on inverse photon efficiency (S/XB coefficients) from the ColRadPy collisional-radiative model enables direct comparison with spectroscopic measurements. Model predictions using a plasma composition of 1% oxygen and 99% deuterium show good agreement with observed W − I (400.9 nm) emission for discharge #57877, supporting the accuracy of the STRIPE framework. This study focuses specifically on gross erosion calculations to demonstrate STRIPE’s capabilities. Future extensions of this work will incorporate net erosion, re-deposition, self-sputtering effects, and whole-device modeling of sputtered tungsten impurity transport. STRIPE is also being applied to other RF-heated linear and toroidal devices, offering valuable insights for antenna design, impurity control, and performance optimization in next-generation fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Filterscope Techniques for Parasitic Signal Screening

Here, the filterscope diagnostic uses bandpass filters and photomultiplier tubes (PMTs) to detect specific spectral emission lines. A filterscope was used to measure the W I 400.88 nm line emission as a function of ion energy on the Radio Frequency Plasma Interaction Experiment (RF PIE) for the purpose of assessing W erosion on plasma-facing components (PFCs). Different filter techniques are being explored and compared in order to effectively screen out nearby impurity lines, like the Ar II 401.39 nm line. The effectiveness of these techniques is determined by comparing the measurements to a high-resolution 1.0 m Czerny–Turner spectrometer with 0.012 nm spectral resolution. The ability to filter out nearby impurity emissions is useful when imaging PFCs in fusion devices including divertor and antenna guard limiters. Initial results with a helium plasma show little difference between the two techniques at higher bias voltages. In a helium plasma at lower dc bias voltages, a two-filter technique with filters at two separate wavelengths was shown to be more effective at screening out background signals. Data collected with an argon plasma however show the technique with overlapping filters on the line of interest is a closer match to spectrometer data at lower dc bias voltages.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Mechanisms behind impurity spreading in a channeled, long-leg divertor configuration

One suggested benefit of long-leg, tightly baffled divertor configurations, a common feature included in the designs of many next-generation fusion devices, is access to a delocalized “spread” of impurity radiation along the leg, allowing for a substantial radiation region while maintaining both sufficient distance from the core to prevent pollution of the plasma fuel and avoiding local overheating of the plasma-facing materials. However, it is unclear whether this increased connection length and divertor volume afforded by the long-leg configuration is compatible with impurity transport, and whether it will enable this desired impurity radiation spread at all. Here, the SOLPS 4.3 code is used to assess the distribution of nitrogen and neon impurity radiation in a long-leg, tightly baffled divertor geometry. Further, there is a strong correspondence between the distribution of impurity radiation in the long leg and impurity recycling on the divertor side walls with both the nitrogen simulation set and the neon simulation set, suggesting that impurity spreading along these long legs is due to the interplay between the cross-field transport and associated neutralization of the impurity on the material surfaces.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Modeling of small tungsten dust grains in EAST tokamak with NDS-BOUT ++

In order to investigate the transport of small dusts as well as their evolution property along their trajectories, the NDS module is developed under the BOUT++ framework, a highly desirable C++ code package to perform parallel plasma fluid simulations with an arbitrary number of equations in three-dimensional curvilinear coordinates. Due to the severe dust ablation in fusion plasmas, the dust size would decrease from micrometer to nanometer, resulting in impurities. Small dusts in the simulations here are specified as tungsten spheres with the radii on or below the order of submicrometer. The Rayleigh limit is included in the charging process when the dust is ablated to the droplet phase. The simulation results from the NDS module show that a 200 nm radius spherical tungsten dust originated from upper divertor region of EAST Tokamak is ablated completely due to the intense heating from the incoming plasma inside the core region, well consistent with the CCD footage of EAST shot # 81459. Furthermore it is found that the magnetic field dominates the dust transport when the dust radius is below 100 nm during the ablation along the trajectory. Our simulations predict that a 10 nm radius spherical tungsten dust injected from the inner midplane is well constrained by the magnetic field, and it reaches the inner divertor target with a velocity on the order of km/s.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Extended-MHD simulations of disruption mitigation via massive gas injection in SPARC

Recent developments to the M3D-C1 code enable higher fidelity modeling of disruptions, and can be applied in the design verification of reactor-scale tokamaks. Among these new capabilities is a method to mesh conducting vessel structures such as coils and passive plates, packing of the toroidal mesh around gas injectors, as well as anisotropic resistivity inside the vessel structures. We present extended-magnetohydrodynamic (MHD) simulations of disruption mitigation via massive gas injection (MGI) in SPARC. The goal of this study is to inform the disruption mitigation layout of SPARC and aid in the design of an effective gas injector configuration. Fully three-dimensional simulations with M3D-C1 are carried out for various injector configurations with the primary goal of determining the effect of different MGI parameters on heat loads and vessel forces. The simulations include a model for impurity ionization, recombination, advection and radiation, as well as spatially resolved conducting structures around the plasma. A localized mixture of deuterium and neon with a small toroidal and poloidal width is injected in up to six locations. We demonstrate that M3D-C1 can model a rapid shutdown via MGI using narrow and more realistic gas plumes than in previous simulations. As a result of the q = 1 surface in the SPARC baseline case a sawtooth is observed early in the simulations. Despite the sawtooth and the onset of edge MHD instabilities, the impurity distribution remains localized around the injector locations, but enables a radiative shutdown of the plasma. We find that using the maximum of six gas injectors results in a lower peaking factor and leads to a more even distribution of radiation toroidally than using two injectors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Elucidating Hydrogen Reaction-Induced Water Desorption from Oxide-Passivated Metal Surfaces for Plasma Applications

Elucidating the mechanisms responsible for sub-microsecond desorption of water and other impurities from electrode surfaces at high heating rates is crucial for understanding pulsed power behavior. Ionization of desorbed impurities in the vacuum regions causes power or current loss; devising methods to limit such desorption during the short time scale of pulsed power is needed to improve corresponding applications. Previous molecular modeling studies have strongly suggested that, under high vacuum conditions, the amount of water impurity desorbing from oxide surfaces on metal electrodes is at a sub-monolayer level at room temperature, which appears insufficient to explain observed pulsed power energy losses at high current densities. In this work, we apply Density Functional Theory (DFT) techniques to show that hydrogen trapped inside iron metal can diffuse into hematite (α-Fe 2 O 3 ) on the metal surface, ultimately reacting with the oxide to form Fe(II) and H 2 O. The latter desorbs at elevated temperature and may explain the anomalous amount of desorbed impurity inferred from pulsed-power experiments. We also apply a suite of characterization techniques to demonstrate that when iron metal is heated to 650 °C, the dominant surface oxide component becomes α-Fe 2 O 3 . The oxide facets exposed are found to be a mixture of (0001), (10-10), and others, in agreement with the DFT models used.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2021

Stabilization of kink/peeling modes by coupled rotation and ion diamagnetic drift effects in quiescent H-mode plasmas in DIII-D and JT-60U

Magnetohydrodynamic stability at the edge pedestal in several quiescent H-mode (QH-mode) plasmas in DIII-D and JT-60U experiments was analyzed by considering plasma rotation and ion diamagnetic drift effects. It was identified that a kink/peeling mode, which is a prime candidate for a trigger of edge harmonic oscillation in QH-mode, is stabilized by plasma rotation when considering the ion diamagnetic drift simultaneously in both experiments. Furthermore, the stabilizing effect by rotation becomes more effective in case using the rotation profile of the main ion species evaluated by assuming radial force balance. In addition, when inverting the rotation direction, it was found that the kink/peeling mode is more stabilized when considering the rotation of the main ion species, though the mode is less stabilized by the rotation of impurity ion species. The result implies that the kink/peeling mode stability in QH-mode plasmas is sensitive to how the rotation profile is evaluated, but it is shown that a qualitative trend stabilizing the kink/peeling mode by rotation can be reproduced with the poloidal rotation profile of an impurity predicted numerically based on the neoclassical theory.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Implementation of an impurity diagnostic suite on the Pegasus-III experiment

A suite of diagnostics used to assess impurity content and dynamics has been updated, upgraded, and installed on the Pegasus-III Experiment. Typical plasma parameters during local helicity injection start-up are τ shot ~10 ms, n e ~ 1 × 10 19 m –3 , and T e ~ 50 eV. The deployed diagnostics are compatible with this modest temperature and density regime and provide species identification, source localization, and estimation of radiation losses. Impurity species are determined by recording time-evolving, single line-of-sight spectra at 1.25 kfps using a SPRED (Survey, Poor Resolution, Extended Domain) vacuum ultraviolet spectrometer. SPRED is equipped with 450 g/mm grating, giving a spectral resolution of 0.33 nm and a spectral range from ~10 to 110 nm, useful to identify light impurity species in this temperature and density range. An absolutely calibrated spectrometer that collects light from the plasma at R tan = 15.9 cm and Δt ≥ 2 ms is used as a visible survey spectrometer and for continuum measurements. The radiated power from the plasma is estimated with a photodiode-based diagnostic. Two 16-channel absolute extreme ultraviolet diode arrays are placed behind pinhole apertures, resulting in 32 lines of sight at Z = 0, with a spatial resolution of 2–3 cm and a time response of 60 kHz. A photometrically calibrated collinear D α /near infrared filtered photodiode-based system measures the D α emission and around 1040 nm. All these instruments have been designed to suppress electromagnetic interference from megawatt-class switching power supplies.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Overview of the KSTAR experiments and future plan

The Korea Superconducting Tokamak Advanced Research (KSTAR) device has recently entered a new phase of operation following the installation of a tungsten mono-block divertor, providing a unique platform to investigate high-performance, long-pulse plasma scenarios for future reactors. The 2023–2024 experimental campaigns demonstrated significant progress in commissioning the tungsten divertor, developing plasma operation scenarios, and achieving real-time control. Comparative studies revealed increased core radiation losses and associated performance degradation relative to the carbon divertor environment. Alleviation strategies, such as optimized neutral beam injection timing, boron powder seeding, and impurity control through gas puffing, were shown to reduce tungsten accumulation and extend high-performance phases, including record H-mode operation exceeding 100 s. International collaboration enabled the first demonstration of high poloidal beta scenarios with an internal transport barrier on KSTAR. Advances in control included upgraded real-time resonant magnetic perturbation schemes and disruption forecasting. Furthermore, divertor detachment control using a new surrogate-model-based approach with real-time radiation imaging demonstrated active impurity and heat flux management. Supported by enhanced heating, current drive, and diagnostic systems, these achievements offer valuable insights into tungsten impurity behavior, transport physics, and control strategies for future reactors including ITER. Planned upgrades toward a full-tungsten wall and ITER-aligned real-time plasma control will further establish KSTAR as a leading experimental platform for developing operational scenarios required to achieve steady-state, burning plasma conditions.

KSTAR↗