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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

The effect of high-power transient events on tungsten and tungsten coatings used for radio frequency launcher applications

High-temperature plasma-facing material coatings used for radio frequency (RF) launchers need to be robust enough to survive RF breakdown arcing or other transient events from the plasma (e.g., an edge localized mode) without causing a catastrophic failure of the coating. High-power transient effects are being explored by using an RF-induced vacuum arc to determine the robustness of tungsten coatings made by a variety of manufacturing methods. A 1/4-wavelength resonant section of vacuum transmission line terminated with an open circuit electrode structure with a well-defined electric field (30-60 kV/mm) produces repeatable arcing conditions. The initial focus is on tungsten as a plasma-facing material, including sintered tungsten, tungsten coatings on steel produced via physical vapor deposition (PVD), and functionally graded tungsten/steel coatings deposited by low-pressure plasma-spraying (LPPS). Thin PVD coatings (1-2 microns) fail catastrophically from an arc and result in severe delamination of the coating. The arc-induced damage of thicker coatings, such as those made via LPPS, tend to be restricted to the top few microns of the surface. Arcing often initiates on sharp surface microstructures and causes localized melting of tungsten at the surface of all the materials studied and results in resolidified melt pools with surface cracks. The resolidified surface results in a reduction in overall deuterium retention when exposed to typical RF plasma sheath conditions.

Caughman, John [ORNL] (ORCID:0000000206091164)↗
DOE OSTI2024

Power and isotope effects in the ITER baseline scenario with tungsten and tungsten-equivalent radiators in DIII-D

Abstract Experiments in DIII-D document the ITER Baseline Scenario (IBS) at q 95 ∼ 3 and P IN / P LH ∼ 1–2, in both deuterium and hydrogen utilizing Kr and Xe as Tungsten-equivalent radiators. The power threshold for H-mode operation ( P LH ) was determined experimentally without added impurities and found to be about a factor of two higher than the scaling law. In recent IBS experiments in deuterium, intrinsic levels of metals such as Tungsten (W) or molybdenum and inconel are present that reduce the pedestal pressure by 20%–25%. A complete radiative collapse of deuterium IBS plasmas occurs at W core concentrations C W = 10 −5 . Simulations show that for core temperatures expected for ITER, the plasmas would not have a radiative collapse at C W = 1 × 10 −5 , moreover Q = 8–10 would still be achieved for C W up to 3 × 10 −5 . In contrast to deuterium, the IBS in hydrogen is not affected by intrinsic high-Z impurities, indicating that hydrogen H-modes in ITER may not inform the D-T phase with respect to W accumulation and discharge survival. Compared to deuterium, the pedestal pressure in hydrogen is ∼25% lower, with much higher ELM frequency of 150 Hz, decreasing with input power. Krypton was injected in a matrix scan of input power and impurity flow in IBS hydrogen discharges. Krypton impurity density profiles in hydrogen are similar to deuterium plasmas, but at Kr flows that are 2–3 times higher for the same input power. Krypton is transported into the core and affects the whole radius; at the highest injection rates a radiative collapse occurs at core radiation fractions of 0.3–0.35, consistent with the expected maximum W radiation fraction for ITER core plasmas. Comparing the results with previous International Tokamak Physics Activity database studies of the IBS confirms that at higher radiation fraction due to high-Z impurities, a drop in H 98 of >10% is observed. On the other hand, the results using Kr as a W-equivalent radiator indicate that metal (W) devices at lower core temperatures than ITER may provide overly pessimistic performance extrapolations to ITER for deuterium-tritium operation. The new DIII-D results support a more attractive option for the ITER Research Plan with a short hydrogen phase for system commissioning, transitioning to deuterium operations as soon as possible to provide relevant conditions for deuterium-tritium operations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Interpretive modeling of tungsten divertor leakage during experiments with neon gas seeding

Abstract Many existing and future tokamaks with tungsten divertors operate, or will operate, with low- Z impurity seeding, but the direct effect of these seeded impurities on tungsten Scrape-off-Layer (SOL) transport has not been explored in detail. This paper reports on a DIII-D experiment designed to test how tungsten divertor leakage from the Small-Angle Slot V-Shaped, tungsten-coated divertor is impacted by neon seeding at a variety of injection rates and poloidal injection locations. Measurements from the experiment show an inverse relationship between the neon injection rate and the tungsten core penetration factor. Interpretive modeling is performed with a combination of the SOLPS-ITER and DIVIMP codes to assess the underlying tungsten behavior. The modeling results show that the reduction in tungsten divertor leakage is driven by both an increase in the divertor collisionality as well as a reduction in the ion temperature gradient near the divertor target. Collisions between low- Z impurities and tungsten impurities are found to have a significant impact on the tungsten SOL transport, such that ignoring the low- Z impurity collisional effects on the tungsten transport can result in an overestimate of the divertor leakage by an order-of-magnitude. Given the importance of these localized interactions, neon seeding from the closed, slot-like divertor has a clear advantage in being able to reduce tungsten divertor leakage without the high levels of neon core contamination that occur when seeding from other poloidal locations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Corrigendum: Simulations of stand-off runaway electron beam termination by tungsten particulates for tokamak disruption mitigation (2024 Nucl. Fusion 64 056019)

In the original article, we inadvertently omitted a reference which previously introduced the concept of tungsten injection for 'Disruption Mitigation in Tokamak Reactor via Reducing the Seed Electrons of Avalanche.' In that reference, the authors have proposed injection of a tungsten cylinder of 80 mm by 8 mm dimensions, by rail guns, to absorb the runaway seed population. This can be contrasted with another interesting idea of using small tungsten pellets coated with a low Z material for depleting the runaway seeds as an option for upgrade of the ITER Disruption Mitigation System. Interested readers are referred to those two papers on the tungsten injection for runaway seed removal/reduction, and the related issue of not shortening the current quench excessively. The application focus of was not on reducing the runaway seeds in the initial Ohmic-to-runaway current conversion phase, but on safe termination of a fully formed and likely decaying runaway beam that is about to scrape off against the first wall, which can be accelerated by vertical displacement events. This scheme thus serves as a last line of defense against potential wall damage. The choice here is locally released tungsten particulates, similar to the previous idea of using tungsten particulates as a dust shield for the divertor. The most significant finding of is that strong pitch angle scattering, in addition to energy attenuation and absorption, of high-energy runaways by tungsten particulates, provides a transport mechanism by which runaways damage of the first wall can be mitigated by reducing and spreading the runaway wall load. We regret the omission in the original article, and thank the authors of the cited article for bringing this matter to our attention. We also find the possibility of using the injected tungsten rod from to terminate the runaways, as opposed to the cloud of tungsten particulates in intriguing. In that context, it is of interest to mention another possibility of a retractable tungsten metal arm that can be swung out for deployment in runaway termination, which would remove the need for post-mitigation recovery of tungsten debris in injection schemes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Synthesis and Photodriven Hydrogenation of Tungsten Nitride Complexes Prepared from Dinitrogen Cleavage

Oxidation of the Chatt-type tungsten dinitrogen compound, trans-(depe) 2 W(N 2 ) 2 (depe = Et 2 PCH 2 CH 2 PEt 2 ), with [(η 5 -C 5 H 5 ) 2 Fe][BAr F 4 ] (BAr F 4 = B(3,5-(CF 3 ) 2 C 6 H 3 ) 4 ) resulted in isolation of [(depe) 2 WN][BAr F 4 ], a rare example of a tungsten(IV) nitride prepared from N 2 cleavage. A bimetallic μ-N 2 ditungsten intermediate supported by terminal N 2 ligands was identified, and irradiation with visible light promoted dinitrogen cleavage and formation of [(depe) 2 WN][BAr F 4 ]. Performing the analogous one-electron oxidation of the related tungsten dinitrogen compound, trans-(dppe) 2 W(N 2 ) 2 (dppe = Ph 2 PCH 2 CH 2 PPh 2 ), furnished the corresponding cationic, 17-electron tungsten dinitrogen complex, [(dppe) 2 W(N 2 ) 2 ][BAr F 4 ], that was characterized by X-ray diffraction and vibrational and EPR spectroscopies. The generation of [(dppe) 2 W(N)][BAr F 4 ] was observed in low yield from the in situ formed mixed N 2 -bridged compound, [(N 2 )(depe) 2 W(μ-N 2 )W(dppe) 2 (N 2 )][BAr F 4 ] 2 , and was confirmed by independent synthesis using 1-azidoadamantane. Addition of ammonia or water to [(depe) 2 WN][BAr F 4 ] resulted in formation of the cationic imide and hydroxide complexes, [(depe) 2 W(NH)(X)][BAr F 4 ] (X = NH 2 , OH). Irradiation of [(depe) 2 WN][BAr F 4 ] with 440 nm visible light in the presence of Ir(ppy) 3 (ppy = 2-phenylpyridine) under 4 atm of dihydrogen resulted in hydrogenation of the tungsten nitride to the cationic tungsten pentahydride, [(depe) 2 WH 5 ][BAr F 4 ], with the release of free ammonia in 21% yield, a rare example of ammonia generation from dinitrogen and dihydrogen from a well-defined tungsten nitride.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Drift-kinetic effects of tungsten on plasma response to RMP in ITER

Here, effects of high- Z ( Z is the particle charge number) tungsten impurity ions on the plasma response to the resonant magnetic perturbation (RMP) field are numerically investigated for the ITER 15 MA baseline scenario, where the tungsten contribution to the plasma response is computed with a drift-kinetic model while the bulk thermal particle contributions follow the fluid approximation. The study yields three highlights: (i) the drift-kinetic contribution of the tungsten impurity exerts minor influence on the plasma response compared to that computed by the pure fluid model without tungsten; (ii) a new figure of merit, based on the resonant spectrum perturbation at the plasma boundary surface, results in different optimal coil phasing compared to that previously obtained by maximizing the edge-peeling plasma response; (iii) the optimal $n = 3$ RMP (for edge localized mode (ELM) control, $n$ is the toroidal mode number) is found to induce a large tungsten particle influx near the plasma edge associated with the neoclassical toroidal viscosity. The study thus provides useful data on the compatibility of the full tungsten wall with RMP ELM control in ITER.

ITER↗
DOE OSTI2024

Survey of tungsten gross erosion from main plasma facing components in WEST during a L-mode high fluence campaign

An initial high fluence campaign was performed in WEST, in 2023, on the newly installed actively cooled tungsten divertor composed of ITER-grade monoblocks. The campaign consisted in the repetition of a 60 s long Deuterium L-mode pulse in attached divertor conditions, cumulating over 10000s of plasma exposure. A maximum deuterium fluence of approximately 5⋅1⁢026 m−2 was reached in the outer strike point region, representative of a few high performance ITER pulses. Gross tungsten erosion inferred from visible spectroscopy shows that the most eroded plasma facing component is the inner divertor target with rates ten times larger than on the outer divertor target. The outer midplane tungsten bumpers, located a few centimeters from the plasma, show gross erosion rates two times lower than at the outer divertor. We conclude that the outer midplane bumpers have a negligible contribution to the long range tungsten migration and deposition onto the lower divertor. The cumulated gross erosion rate on the inner divertor translates in an effective gross erosion thickness of about 20μ⁢m, while it is about 2μ⁢m for the outer divertor. Strikingly, these orderings coincide with the thickness of deposits found locally on the divertor: the exposed surfaces of high field side monoblocks are covered with several tens of μ⁢m tungsten deposits, while on the lower field side, few μ⁢m thin tungsten deposits are only found on the magnetically shadowed parts of monoblocks. The strong impact of those deposits on WEST operation, namely perturbation of surface temperature measurement with infra-red thermography, and the emission of flakes causing radiative perturbation of the confined plasma, calls for anticipating similar issues in ITER. In particular, the start of research operation shall consider the definition of a divertor erosion budget in order to anticipate the formation of deleterious deposits.

Fedorczak, N.↗
DOE OSTI2024

Discovering tungsten-based composites as plasma facing materials for future high-duty cycle nuclear fusion reactors

Abstract Despite of excellent thermal properties and high sputtering resistance, pure tungsten cannot fully satisfy the requirements for plasma facing materials in future high-duty cycle nuclear fusion reactions due to the coupled extreme environments, including the high thermal loads, plasma exposure, and radiation damage. Here, we demonstrated that tungsten-based composite materials fabricated using spark-plasma sintering (SPS) present promising solutions to these challenges. Through the examination of two model systems, i.e., tungsten-zirconium composite for producing porous tungsten near the surface and dispersoid-strengthened tungsten, we discussed both the strengths and limitations of the SPS-fabricated materials. Our findings point towards the need for future studies aimed at optimizing the SPS process to achieve desired microstructures and effective control of oxygen impurities in the tungsten-based composite materials.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

The effect of ICRH and NBI on tungsten transport in neon edge radiation cooled discharges in TEXTOR

Tungsten is planned to be the plasma-facing material for the main chamber and divertor in future devices like ITER, SPARC, and DTT. To address risks associated with tungsten, R&D is being carried out on various toroidal confinement devices. One particular research question is related to the optimum heating mix for ITER. To investigate the influence of the heating scheme on the release and transport of tungsten, a comparison of neutral beam injection (NBI) and ion cyclotron resonance frequency (ICRF) heating was carried out in TEXTOR tungsten test limiter experiments. The experiments were performed under standard L-mode conditions and in radiative improved-mode operation with neon seeding and boronized walls covering the graphite plasma-facing components. The plasma was heated with hydrogen or deuterium NBI alone or with deuterium NBI in combination with H-minority, ion cyclotron resonance heating. A movable solid tungsten limiter was inserted through a limiter lock system into the edge plasma. The impurity release from this limiter was evaluated from visible spectroscopy. The tungsten concentrations in the plasma core were determined by extreme ultraviolet spectroscopy and bolometry. With deuterium NBI alone, strong central radiation and accumulation of W was observed. This can, however, be avoided by adding ICRF heating.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Rapid synthesis of phase-engineered tungsten carbide electrocatalysts via flash joule heating for high-current-density hydrogen evolution

Fabricating durable and high-performance electrocatalysts operating at high current densities for industrial acidic hydrogen evolution remains a daunting challenge. Tailoring the phase composition of electrocatalysts is a promising strategy to harness synergistic effects and improve charge transfer, thereby optimizing their performance. Here, this work presents a fast, green method based on flash joule heating (FJH) to synthesize phase-engineered tungsten carbide electrocatalysts for the acidic hydrogen evolution reaction (HER) at high current densities. Tungsten carbide electrodes with varying FJH treatment durations (3, 10, 30, and 60 s) are fabricated to fine-tune the mixture of tungsten monocarbide (WC) and tungsten semicarbide (W 2 C) phases. Results show that samples with a 30-s treatment (TC-3) exhibit an optimal balance between these phases, leading to a low overpotential of 180.97 and 387 mV at current densities of 10 mA/cm 2 and 4 A/cm 2 , respectively. TC-3 exhibits significantly lower charge transfer resistance compared to the other electrocatalysts, which can be attributed to its optimal phase ratio. Notably, the TC-3 electrocatalyst remains stable for over 9 days at 4 A/cm 2 due to their controlled phases and excellent corrosion-resistant properties. This work highlights a new method to fabricate cost-effective, high-performance tungsten carbide electrocatalysts with well-controlled phase compositions.

Acidic hydrogen evolution reaction↗
DOE OSTI2025

Operational space for lower hybrid heating scenarios in the full tungsten environment of WEST

In tungsten—W—Environment in Steady-state Tokamak (WEST), the lower hybrid current drive (LHCD) system is key for achieving long pulse operation by providing most of the non-inductive plasma current, as well as a crucial source of electron heating. Therefore, determining the operational space for its application is fundamental. In the present study, the LHCD operational space is deeply analyzed for 0.5 MA pulses. This space is bounded by three limits: (i) the ratio of the LHCD power over density must be above a threshold to compensate tungsten radiation with enough core heating, (ii) the line-averaged density must be high enough to allow good coupling of the hybrid wave with the plasma, and (iii) fast electron ripple losses must be below a limit to avoid reaching a thermal threshold on plasma-facing components. If the tungsten radiation peak or burn-through phase is not safely overcome, a maximum electron temperature of 1.5 keV is obtained, confinement is degraded, and magnetohydrodynamic activity is frequently triggered, potentially causing a disruption. From experimental measurements and interpretative simulations, we highlight the main mechanisms that prevent the plasma from heating up during LHCD power ramp-up. Three parameters play a major role: plasma density, tungsten concentration and LHCD power deposition. A strategy to overcome this limitation is found: a precise density ramp-up performed simultaneously with the increase in LHCD power. Additionally, we show that boronization greatly facilitates the burn-through of tungsten by lowering its content during the heating phase. Finally, taking into account the three constraints given above, the LHCD operational space is determined at power ramp-up and during constant heating phases.

lower hybrid heating and current drive↗
DOE OSTI2026

Radiation-front position control with scenario adjustments during the first tungsten divertor operation on KSTAR

This study reports the first experimental demonstration of a radiation front-based detachment control system during the initial tungsten divertor campaign in KSTAR. The system employs real-time infrared video bolometry with tomographic reconstruction to track the inboard radiation-front position, which served as a feedback variable for impurity seeding. In a baseline scenario without pre-D 2 fueling, tungsten accumulation triggered core radiation surges and repetitive H–L back transitions. In contrast, with pre-D 2 fueling, detachment was stabilized by maintaining higher edge-localized mode frequency, suppressing core tungsten buildup, and reducing sputtering, thereby enabling sustained feedback control. The suitability of the inboard front position as a detachment control metric was confirmed by its correlation with the degree of detachment: with D 2 fueling the relation exhibited a broad hysteresis loop characteristic of momentum-loss-dominated detachment, whereas with N 2 fueling, it became narrow and nearly vertical, consistent with radiation-loss-dominated pinning. These results provide a proof-of-principle for radiation-front-based detachment control, demonstrating its viability as a control approach. At the same time, these results clarify that the detachment outcomes were further conditioned by impurity transport characteristics of tungsten divertor operation in KSTAR, providing guidance for future scenario optimization and controller development in tungsten devices.

KSTAR↗
DOE OSTI2025

Understanding the limitations and potential of micro tensile testing of tungsten and needs for crystal plasticity modelling

Tungsten is the leading plasma facing material candidate due to its exceptional properties. Understanding the response to neutron irradiation is crucial for the lifetime evaluation of tungsten. The limited space in nuclear reactors and the high levels of radioactivity of the specimens after irradiation are significant barriers to accurately measuring the mechanical properties after neutron irradiation. Testing of micro tensile specimens is one approach to reduce the total amount of irradiated material needed for a set of mechanical testing experiments. However, micro samples are not necessarily measuring the bulk material properties, as size effects produce higher measured mechanical properties than is observed in engineering size counterparts. Determining the minimal sample size for reliable bulk property measurement, combined with modeling efforts, is essential. We conducted room temperature tensile tests on tungsten micro tensile specimens fabricated with focused ion beam, plasma focused ion beam and femto-second laser ablation system to dimension of 2×2×7?µm³, 7×7×18?µm³, and 80×100×233?µm³ (width×thickness×gauge-length), respectively. Elevated tensile testing was performed up to 475°C on 5×5×18?µm³ specimens fabricated with focused ion beam. The smallest specimens exhibited a high degree of ductility and strength, whereas the largest specimens demonstrated behavior akin to bulk tungsten. To investigate the effects introduced by the micro specimen fabrication processes and to obtain the necessary bulk dislocation density for the crystal plasticity model, we employed X-ray diffraction depth profiling. This measurement was performed using different X-ray sources with varying penetration depths. Finally, the potential and limitations of micro mechanical tests for tungsten will be discussed.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Comparison of tungsten versus molybdenum for double shell capsules using machine learning design optimization

Double shell targets are an alternative ignition platform for inertial confinement fusion. One design consideration for double shell targets is the choice of inner shell material to help trap radiation emitted by the hot fuel to aid ignition. Materials such as molybdenum and tungsten are of interest for the inner shell layer of the targets. While molybdenum has a lower density that could inhibit instability growth and allow for radiography and code benchmarking, tungsten has a higher density that could provide more compression and confinement. These tradeoffs have been explored using optimized designs for each material. Our previous work [Vazirani et al., “Coupling 1D xRAGE simulations with machine learning for graded inner shell design optimization in double shell capsules,” Phys. Plasmas 28, 122709 (2021); Vazirani et al., “Coupling multi-fidelity xRAGE with machine learning for graded inner shell design optimization in double shell capsules,” Phys. Plasmas 30, 062704 (2023); and Vazirani et al., “Bayesian batch optimization for molybdenum versus tungsten inertial confinement fusion double shell target design,” Stat. Anal. Data Min. 17, e11698 (2024)] resulted in a multi-fidelity Bayesian optimization framework to find yield-optimized double shell target geometries. By leveraging simulations of varying fidelities (one-dimensional and two-dimensional) to inform one another, the multi-fidelity optimization was able to optimize a design in the highest fidelity with significantly fewer simulations than would be used in a systematic parameter scan. In this work, we apply the multi-fidelity Bayesian optimization to explore the optimized designs of double shell targets with molybdenum and tungsten inner shells as well as the physics producing the high performing implosions. A physics exploration of all the simulations used in this study shows trends in designs that contribute to high yields, ion temperatures, and fuel areal densities. Comparison of molybdenum and tungsten simulations shows that they can produce similar implosion conditions with different geometries, which would be important to study in experiments. Graded density layers produce varying performances with the two materials but continue to be of interest for future studies along with studies of doped inner shell materials and applied surface roughness.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Electron-impact excitation data for W 2+ in support of tungsten spectroscopy and re-deposition measurements for magnetically-confined plasmas

Abstract To better understand plasma wall interactions involving tungsten, accurate atomic structure and electron-impact driven collisional processes for near-neutral ion stages of tungsten are required. Complementing existing work on neutral and singly ionised tungsten, atomic structure and collisional calculations for W 2+ electron-impact excitation have been completed. These excitation calculations are an important component of S/XB coefficients for near-neutral charge states, which may be used to spectroscopically infer re-deposition of tungsten at the plasma-solid boundary of fusion relevant devices. With W 2+ in particular having emission lines that can be observed at ultraviolet (UV) wavelengths, while higher charge states of tungsten are unlikely to have lines possible to observe outside of the vacuum UV range. The atomic structure was generated using the General-purpose Relativistic Atomic Structure Package (GRASP 0 ), implementing the Multi-configuration Dirac Fock approach. This structure was the basis for a subsequent Dirac R -matrix electron-impact excitation calculation to provide Maxwellian averaged rate coefficients. A synthetic spectrum was generated from this data using a collisional-radiative model to predict the strongest W III spectral lines and these lines were compared to emission from the Compact Toroidal Hybrid (CTH) plasma device. Several of the strongest W III lines are observed in CTH and agree well with the modelled line wavelengths and intensities, a table of these lines is provided that could be observed in other devices.

McCann, M. (ORCID:0000000215321240)↗
DOE OSTI2024

WEST full tungsten operation with an ITER grade divertor

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗