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The article considers different methods of divertor plasma diagnostics planned for using in the T‑15MD tokamak. Technical problems arising during operation of optical systems in the divertor zone are discussed, including degradation of in-vessel optical elements. The main attention is paid to the conceptual design of the "Passive Spectroscopy in Divertor" diagnostic system. The optical system, including in-vessel mirrors, and methods for protecting its components from the negative effects of plasma are described in detail. Using synthetic diagnostics and numerical simulation methods, the possibility of solving the problem of tomographic reconstruction of the two-dimensional profile of plasma radiation in the T-15MD divertor is demonstrated. Based on the results presented, it was concluded that passive spectroscopy can be used for obtaining data on plasma parameters in the divertor with good spatial resolution, which will make it possible to study the physics of processes and monitor the operation of the T-15MD divertor, including the operation in detachment regime. [ABSTRACT FROM AUTHOR]

The heat flux width (λq) is a key parameter determining the heat load at divertor targets. In recent years, drifts have been found to play a remarkable role in the edge plasma transport, while its influence on λq has not been well understood. Investigations of the influence of drifts on λq, systematic simulations using the SOLPS-ITER code are performed in this work. The statistics of the simulation results show that the drift under favorable/unfavorable B T tends to increase the λq in the outer/inner side and decrease the λq in the other side, which is consistent with the experiment observations. At the upstream and the target, the mechanisms of the influence of the drifts on λq are different. The upstream λq (λq,u) is directly affected by the drift-induced convective heat flux, while λq at the target (λq,t) is indirectly influenced through heat conduction (in the high-recycling regime) and the sheath (in the detached regime) due to the change of plasma parameters there. Furthermore, the synergetic effect of geometry and drift under favorable B T leads to an anomalously large λq,t in the inner side at high density. [ABSTRACT FROM AUTHOR]

Experiments performed during strongly-shaped high-power diverted negative triangularity (NT) experiments in DIII-D achieved detached divertor conditions and a transient-free edge, showcasing the potential for application of NT to a core-edge integrated reactor-like scenario and providing the first characterization of the parametric dependencies for detachment onset. Detached divertor conditions will be required in future devices to mitigate divertor heat fluxes. Access to dissipative divertor conditions was investigated via an increase in upstream density. Detachment onset at the outer strike point was achieved with H-mode level confinement H 98 − y 2 ∼ 1 and reactor-relevant normalized pressures β N ∼ 2 . Confinement degradation was observed with deeper detachment, associated with the loss of an electron temperature pedestal. Differences in geometry, radial transport, impact of cross field drifts are discussed to explain differences in access to detachment in NT discharges. Higher normalized densities, with respect to equivalent discharges in positive triangularity, were necessary to achieve detachment, partially explained by the shorter parallel connection length to the targets. The effect of cross-field particle drifts (E × B, B × ∇ B) on access to detachment was demonstrated by the lower upstream density needed to access detachment with ion B × ∇ B drift directed outside of the active divertor (Greenwald fraction f Gw ∼ 0.9–1.0 vs f Gw ∼ 1.3). The upstream density at detachment onset was observed to increase linearly with plasma current with ion B × ∇ B drift into the divertor, consistent with the observed narrowing of the scrape-off layer heat flux width λ q . Edge fluid simulations capture separatrix densities needed to achieve detachment in NT plasma and their dependence on drift direction. The ability to reproduce detachment dynamics in NT plasma increases the confidence in future design studies for NT divertors. [ABSTRACT FROM AUTHOR]

A melting failure of W monoblock components of the upper divertor outer target in EAST occurred during the plasma campaigns in 2019. The failure characters and microstructure evolution of the failed W monoblock have been well investigated on one string (W436 string). Near the strike point region where heat flux density is highest, macroscopic cracks and severe surface damage such as dimensional change, melting and solidification are visible in several W monoblocks. At the same time, debonding, melting and migration of Cu/CuCrZr cooling tube components introduced fatal damage to the structure and function. The heat-induced microstructure evolution in the rest part has been examined via hardness tests and metallography. From the heat flux surface to the cooling tube, hardness increased gradually and the recrystallized grains could be found in the region with the highest temperature, while recrystallization grains also appear in some W monoblocks near the cooling tube area. The detailed microstructure has been investigated by metallography and EBSD. Such cases in EAST provide experiences on the extreme condition of accidental loss of coolant or higher discharge power in future devices.

Tungsten (W) has a high melting point, excellent thermal conductivity, and irradiation resistance, making it the most promising plasma facing material for divertors in fusion reactors, which are currently under development. However, since the divertor is exposed to an extremely harsh environment, it is considered necessary to develop suitable and cost-effective repair techniques. In this study, the applicability of the atmospheric plasma spraying (APS) method using a gas shroud as a repair technique for W components was investigated, in particular the possibility of strengthening the repaired part by applying friction stir processing (FSP) as a post-treatment. It was found that the application of a gas shroud can suppress in-flight oxidation to some extent, even when the W is deposited in air. In addition, the FSP treatment reduced grain size and porosity, resulting in an increase in microhardness of approximately 37.5% compared to the base material (W substrate) and 203.5% compared to the as-sprayed material. The gas shroud APS and FSP post-treatments have been shown to have potential as repair techniques for tungsten components in future fusion reactors. [ABSTRACT FROM AUTHOR]

In order to simulate hydrogen (H) plasma in the linear plasma device NAGDIS‐II, we have modified the fluid code LINDA‐NU to allow the simultaneous calculation of multiple ion species consisting of hydrogen atomic ions (H+$$ {\mathrm{H}}^{+} $$) and molecular ions (H2+,H3+$$ {\mathrm{H}}_2^{+},{\mathrm{H}}_3^{+} $$). In this simulation, H and H2$$ {\mathrm{H}}_2 $$ neutrals are assumed to be uniformly distributed in space in order to obtain initial qualitative results. The fraction of H3+$$ {\mathrm{H}}_3^{+} $$ ions increases as the H2$$ {\mathrm{H}}_2 $$ molecular density increases, and the recombination process between H3+$$ {\mathrm{H}}_3^{+} $$ and electrons is observed to reduce the particle flux to the target plate. With an increase in H density, the electron density increases due to the decrease in ion flow velocity due to the change exchange process, and the electron temperature decreases to less than 1 eV, leading to the detached plasma formation attributed to the electron‐ion recombination process. [ABSTRACT FROM AUTHOR]

The collisions between neutral particles and plasma in the divertor region determine divertor detachment and heat flux to the target. However, the diagnostic of neutral particle information in the tokamak experiment is very lacking. To study the behavior of neutral particles in the boundary region flexibly and comprehensively, a two‐dimensional kinetic neutral transport code based on Monte Carlo method is developed in this work. The code is applied to simulate the neutral particles in the boundary of EAST device. Firstly, the simulation results are compared with EIRENE using an identical plasma background, and good agreement is obtained. Secondly, the transport of methane and the corresponding hydrocarbon compounds is simulated using the developed code. The methane injection case is compared with the injection of carbon and deuterium molecule mixture gas (ratio 1:2) to investigate their impact on detachment and fueling. The simulation reveals that, compared to the direct carbon atom injection case, charge‐exchange collision is the dominant collision of neutral methane (hydrocarbons), which suppresses divertor power radiation and plasma detachment while promoting the improvement of fueling efficiency. [ABSTRACT FROM AUTHOR]

Efficient handling of high heat flux on the plasma‐facing components, particularly the divertor targets, poses a significant challenge for the Chinese Fusion Engineering Testing Reactor (CFETR) with fusion power of Gigawatt. This work investigates the divertor plasma detachment of CFETR with a standard ITER‐like divertor geometry by neon (Ne) or argon (Ar) impurity seeding using UEDGE code. The cross‐field drifts terms are switched off, and fluid neutral models and a "fixed‐fraction" impurity model are applied to enable efficient simulations for the study of CFETR detachment. In order to reduce the heat load on the divertor targets below the acceptable level (<10 MW/m2), the impurity fraction (f), pumping speed (S), and upstream density are varied to identify the suitable operations window during Ne seeding. The effects of Ne and Ar impurities on the plasma detachment are compared. It is found that with the power across the core‐edge interface PSOL = 200 MW and separatrix density of 2.8 ×$$ \times $$ 1019m−3$$ {\mathrm{m}}^{-3} $$, Ne impurity fraction ≥1.7%, and Ar impurity fraction ≥0.24% can achieve the partial detachment. Achieving similar total radiation power (˜148 MW), the Ne fraction is 2.3% and the Ar fraction is 0.24%. Moreover, the simulation results indicate that Ar exhibits better power radiation efficiency and core compatibility compared with Ne. [ABSTRACT FROM AUTHOR]

The burst of edge‐localized modes (ELMs) leads to an increase in the energy and particle fluxes to the divertor target. Tungsten (W) is chosen as the primary candidate material for plasma‐facing components (PFCs) in the future fusion devices, and EAST has already upgraded all divertors to use W. Therefore, understanding tungsten target erosion during ELMs and finding the correlation between erosion rate and key ELM parameters are crucial for steady‐state operation. In this work, based on the Vlasov–Poisson model (VPM), we develop a one‐dimensional kinetic parallel transport code to investigate the parallel transport of particles in the EAST device during ELMs and the resulting target erosion. The EAST experiment (#102182) is simulated by VPM code. The simulation results are compared with experimental data as well as free‐stream model (FSM) calculation, showing the accuracy of the code. Considering the presence of lithium (Li) impurities in EAST discharge, the erosion of the W target is simulated. The results indicate that during the burst of ELM, the total average tungsten erosion rate, ΓtotalAVG$$ {\Gamma}_{total}^{AVG} $$, is determined by both deuterium (D) and Li ions. D ions dominate the erosion when the ELM frequency (fELM) is low (ranging from 50 to 175 Hz), while Li impurities become more important than D+ in high‐frequency ELMs (fELM > 175 Hz). As fELM increases, the time‐averaged erosion of the W target first increases and then decreases. Therefore, the reduction of W erosion benefits from high‐frequency ELMs, with impurity ions being the primary contributor to the erosion. [ABSTRACT FROM AUTHOR]

The WEST (W Environment in Steady-state Tokamak) divertor serves as the primary element for heat exhaust and contributes critically to plasma control. The divertor receives intense heat fluxes, potentially leading to damage to the plasma facing units. Hence, it is of major interest for the safety of divertor operation to detect and characterize the hot spots appearing on the divertor surface. This is done through the use of infrared (IR) cameras, which provide a thermal mapping of the divertor surface. In this work, a knowledge-informed divertor hot spot detector is demonstrated, that explicitly accounts for hot spot structure and temperature repartition. A novel neural network, termed as Constrained U-Net, is proposed, which uses as input the bounding boxes of hot spots from prior automatic detection. The Constrained U-Net addresses jointly image segmentation and regression of physical parameters, while remaining compatible with the practical constraints of real-time use. The detector is trained on simulated data and applied to real-world infrared images. On simulated images, it yields a precision of 0.98, outperforming a classical U-Net, and Max-Tree. Visual results obtained on real-world acquisitions from the WEST Tokamak illustrate the reliability of the proposed method for safety studies on hot spots. [ABSTRACT FROM AUTHOR]

In nuclear fusion reactors, the heat removal from high-heat-flux components such as the divertor has always been a challenge as the heat flux is anticipated to easily overcome 10 MW/m2, as it is the case for the divertor tiles in the W7-X stellarator. In this paper a new cooling solution, relying on a Triply Periodic Minimal Surface lattice, is proposed and investigated. The analysis first focuses on the identification of the optimal lattice with a dedicated investigation on a reduced computational domain. The analysis with the best candidate is moved to a 30 cm × 10 cm divertor tile, to prove the compliance with the operating constraints. The tile with the gyroid lattice with unit cell size of 10 x 10 × 5 mm3 is shown to be capable to withstand a heat flux of 10 MW/m2, distributed over a heated length of 10 cm, when 1.25 kg/s of subcooled water are used for cooling, resulting in a pressure drop

Compared to CuCrZr, oxide dispersion strengthened copper alloy (ODS-Cu) exhibits higher stability of properties under irradiation and exposure to elevated temperatures, demonstrating broad application prospects in divertor components. Oxygen-free high thermal conductivity copper (Cu-OFHC) has been frequently employed as an interlayer between W and Cu-based alloy in the fabrication of W/Cu divertor components. This study investigates the effect of joining temperature together with the addition of a Ni interlayer on the interface microstructure and mechanical properties of the Cu-OFHC/ODS-Cu joints. As the joining temperature increased from 680 ℃ to 900 ℃, the interface bonding ratio of the Cu-OFHC/ODS-Cu joints improved from 40.4% to 95.8%, and the impact toughness increased from 23.4 J/cm2 to 133.1 J/cm2. With the addition of a Ni interlayer, the interface bonding ratio increased from 40.4% to 90.3%, and the impact toughness improved from 23.4 J/cm2 to 122.5 J/cm2. Increasing the joining temperature or adding a Ni interlayer effectively reduced interfacial voids, enhanced the interface bonding ratio, and consequently improved the impact toughness of Cu-OFHC/ODS-Cu joints. Then, the W/Cu/ODS-Cu monoblock mock-ups with good interfacial bonding were successfully fabricated under two conditions: at a joining temperature of 900 ℃ without an interlayer and at 680 ℃ with a Ni interlayer. These results provide a fundamental understanding for achieving high-quality Cu-OFHC/ODS-Cu joints and offer technical support for the engineering preparation of W/Cu/ODS-Cu components in future fusion devices.

A new Divertor Tokamak Test (DTT) facility is currently being built in Italy to investigate different divertor configurations under different plasma scenarios. The divertor and in particular its target, consisting of Plasma Facing Units (PFUs), is exposed to high heat loads due to plasma fluence. In this paper, the structural integrity of the PFU is evaluated for a reference Single Null (SN) divertor configuration under three different plasma scenarios. A comprehensive structural integrity analysis has been carried out in three stages. In the first stage, computational fluid dynamic (CFD) analyses of the divertor’s PFU with a cooling channel were performed at three different heat loads corresponding to three plasma scenarios. The temperature fields calculated by the CFD analyses were then used as input for the second stage, in which thermo-mechanical simulations were performed to predict the stresses and displacements in the PFU. Due to the high local heat loads, high stresses or even yielding are expected in the PFU’s structural components. Therefore, in the third stage, the structural integrity of critical cross-sections has been verified using the Structural Design Criteria for In-vessel Components (SDC-IC). It has been demonstrated that structural components of the PFU are able to withstand the expected loads, although some non-structural components experienced yielding while not exceeding the critical values.

The divertor must simultaneously withstand unprecedented high heat fluxes of up to 20 MW/m2 and high-energy neutron irradiation of 14-MeV during fusion reactor operation. Accordingly, it is necessary to simultaneously meet the excellent heat dissipation capacity of the divertor and maintain good material performance in harsh neutron irradiation environments. The flat-type divertor demonstrates better heat transfer performance compared to monoblock divertor. Nevertheless, under the condition of a heat flux of 20 MW/m2, the heat transfer and thermal fatigue performance of flat-type divertor using advanced materials are still unidentified. In this study, we conducted a thorough analysis of the heat transfer capabilities and thermal fatigue characteristics of potassium-doped tungsten (KW)/Cu/Oxide dispersion strengthened copper (ODS-Cu)/reduced activation ferritic/martensitic (RAFM) divertor mockup adopts hypervaportron (HV) structure by numerical simulations combined with experiments. The numerical results indicate that the flat-type KW/Cu/ODS-Cu/RAFM divertor mockup expresses excellent heat transfer capacity. The contact area between the edge of the fins and the bottom surface of the ODS-Cu heat sink induces a significant increase in water flow velocity to ∼15 m/s. The peak temperature of loaded KW surface is only ∼985 °C under the flow rate of 5 t/h and inlet temperature of 20 °C. During the high heat flux tests, the prepared flat-type divertor mockup successful endured 1000 cycles of 20 MW/m2 with the peak temperature of 883 °C, and the surface temperature experienced a fluctuation of 2.4 % during the thermal fatigue tests. This study can provide a strong data reference and technical support for the development of fusion reactors, and is of great significance in advancing the commercialization of fusion energy.

Liquid metal divertors are discussed as a potential alternative to divertor components made entirely of solid materials such as tungsten. However, liquid metal interaction with H isotope plasmas poses entirely new challenges, including the stability of the liquid surface. In this article, the interaction of low-temperature D plasmas with both free liquid tin surfaces, and the liquid metal confined in a capillary porous system (CPS) is presented. Free surfaces showed intolerable instabilities due to formation of gas bubbles in the melt, as opposed to a strong stabilizing effect of the CPS. A tendency of liquid tin to well up and entirely cover the surface of the CPS may be beneficial in terms of quickly re-establishing a tin surface after depletion by transient overloading. On the other hand, this appears to counteract efforts to fully suppress superficial bubble formation and bursting, and the resulting ejection of tin droplets by designing a CPS with very small pores.

The laser-induced grain structure of pure tungsten was evaluated using a nanosecond Q-switched Nd:YAG laser. The grain structure, crystal orientation, and grain boundary angular distribution were analyzed through electron backscatter diffraction analysis, both before and after laser processing. The mechanical properties of pure tungsten after laser processing were also measured. The results obtained indicate that laser processing-induced recrystallization can offer micro-nano surface treatment of materials without compromising the mechanical properties.

Radiation transport simulations of the deuterium Lyman line series using EIRENE predict a factor of 2 increase of Balmer-α emission in the JET-ILW low confinement divertor plasmas in high-recycling and detached conditions caused by the re-absorption of Lyman-β emission. The photon tracing model in the EIRENE code is applied to SOLPS-ITER pure deuterium plasma solutions of JET-ILW to evaluate the contribution of Lyman-β capture towards the population density of D(p=3) and, consequently, the Balmer-α emission. The local population escape factors of Lyman-α and Lyman-β lines in the JET-ILW divertor show steep spatial gradients and indicate local regions with high opacity in high-recycling and detached conditions. The relative contribution of Lyman-β capture towards the population density of D(p=3) is significant in high-recycling and detached conditions. This observation is further corroborated by line-integrated calculations using Balmer-α diagnostic line-of-sights.

Experimental data from NSTX and DIII-D discharges with the snowflake (SF) divertor configurations are analyzed toward the development of the X-point radiator (XPR) concept. The XPR divertor regime was recently realized in standard divertor configurations in several tokamaks The SF divertor configuration, with an additional poloidal field null nearby the main X-point, could provide additional benefits for the XPR: a higher flux expansion inside the separatrix and an extended private flux region. This may lead to lower temperatures and higher neutral and electron densities, which are thought to be essential for XPR stability, initiation, and impurity containment. In this work, 4 MW NBI-heated H-mode NSTX discharges and 3–5 MW NBI-heated H-mode DIII-D discharges with SF-minus and SF-plus divertors, with the ion B×∇B drift toward the lower divertor, with and without D2 and CD4 seeding, were analyzed. Many experimental XPR features were found, including good or slightly degraded H-mode confinement, significant ELM size reduction, nearly complete divertor power detachment and a significant divertor radiated power loss. However, evidence of the XPR extending into the confined region was inconclusive in the NSTX tokamak, while in DIII-D, a number of discharges demonstrated a stable MARFE-like structure inside the separatrix over a wide operating space. The present analysis supports the SF divertor as a good candidate for further XPR scenario development in DIII-D and NSTX-U.

For the first time, we have revealed the spatial structure of a detached plasma with electron–ion recombination (EIR) and molecular-activated recombination (MAR) in response to changes in ion temperature (Ti) and gas pressure. This was achieved using high-density sheet plasma generated by a linear divertor simulator (TPDsheet-ICR). The parallel-plate electrodes were positioned above and below the sheet plasma (∼1019 m−3). We varied the Ti from 3 to 7 eV using ion cyclotron resonance heating and investigated the parameters of the detached plasma near the target in the divergent field region. Plasma emission intensities at Balmer series (Hα and Hγ) and Fulcher band wavelengths (610 ± 10 nm) were measured with a high-speed camera equipped with an Arbaa prism. Additionally, a Langmuir probe measured the electron density and temperature of the plasma. Our results show that as the Ti increases, EIR generated near the plasma periphery disappears, while MAR gradually forms near the plasma center during the transition from detached to attached plasma. This transition occurs because the increase in Ti in the detached plasma transfers energy from ions to electrons, raising the high-energy component of the electrons, resulting in EIR disappearance and the onset of MAR.

We developed and performed a 3D Monte-Carlo simulation on local erosion and deposition in outer divertor target of JA-DEMO. The local erosion and deposition were evaluated considering the divertor plasma parameters and the roof shape of W-monoblocks. First, for local erosion simulation, the trajectories of D ions and impurity (He and Ar) ions were traced before irradiating the monoblock surface. The physical sputtering was calculated by these ions irradiation, considering their charge states, incident-energies and -angles. The highest sputtering flux was produced by multiply-charged Ar ions (Ar3+–Ar5+). The sputtering flux strongly depended on the poloidal positions along the divertor target. Although almost no erosion (sputtering) was obtained in low-temperature zone (∼1 eV), the erosion occurred in mid- (∼10 eV) and high-temperature (>10 eV) zone. Second, for local deposition simulation, the trajectories of W atoms emitted from the monoblock surface were traced. After transport (including ionization and recombination) in the plasma, W atoms and ions were deposited on the monoblock surface. The deposition occurred even on the area where plasmas did not irradiate (shadow). This was due to perpendicular diffusion to magnetic field lines, gyro-orbit effects for W ions and straight-line motion for neutral W atoms. The poloidal distribution of the deposition fluxes was similar to that of the sputtering fluxes. In low-temperature zone, no deposition flux occurred due to very low sputtering flux. The deposition flux in mid-temperature zone increased significantly and that in high-temperature zone remained high.

The estimation of the heat flux density distribution profiles in tokamak devices is a very important research topic for edge plasma physics purposes and also to ensure the safety of the machine. In the radial direction, the heat flux exhibits an exponential decay that could be captured by thermal sensors distributed in the plasma facing components. Radially distributed thermal sensors based on Fiber Bragg grating technology have been embedded in the WEST lower divertor to study the heat flux deposition profiles during plasma operation. The comparison between embedded measurements and a 3D finite element model shows a small decay length (5 – 10 mm) on top of a wider heat flux with a decay length around 30 to 50 mm. A tool using neural network has been developed in order to predict the values of the different parameters describing the deposited heat flux from embedded temperature measurements in steady state. A large span of deposited heat fluxes with maximum heat flux ranging from 1 to 9 MW/m2 and decay length from 5 to 50 mm were characterized using this tool over a database of more than 250 experimental L-mode pulses performed in WEST in attached divertor configuration. The comparison of the predicted heat flux parameters values with macroscopic plasma parameters have revealed the appearance of the narrow component with the increase of the divertor power load (Pdiv) with a threshold dependant of the plasma current (IP).

EAST has been upgraded with top and bottom tungsten divertors to facilitate the high power long pulse operation. A visible spectroscopy diagnostic has been developed progressively to monitor W source as well as line emissions from other particles (e.g., D, He, Li, B, C, N, O, Ne, Ar, Mo, D2, WD) in the divertors. Two sets of lens installed in mid-plane port view tangentially the top and bottom divertors, respectively. Two types of optical fibers transfer the plasma light collected by the lens to three different detecting modules. High spectral resolution (λ/Δλ ∼ 14000) can be achieved by the detecting module consisting of spectrometer and Electron-Multiplying Charge Coupled Device (EMCCD), high temporal resolution (50 μs) is obtained by filtered photoelectric multiplier (PMT), and high spatial resolution (1 mm) is acquired by Complementary Metal-Oxide-Semiconductor Transistor (CMOS) camera with collimation optics. The multichannel optical fiber array with 500 μm core diameter is used to connect the lens and the first two detecting modules. The imaging fiber bundle with 30 μm core diameter transmits the two-dimensional divertor plasma light to the third detecting module, i.e. collimation optics and CMOS camera. The diagnostic is operated automatically in normal EAST discharges to support experimental operation and divertor physics study. The paper presents a technical description of the diagnostic and typical measurements during EAST discharges.

In tokamak devices, the performance of plasma facing components (PFCs) under high heat loads is a key challenge for achieving long-pulse and high-performance plasma operations. The lower divertor of EAST adopts an ITER-like cassette modules (CMs) structure. However, regular distribution of damage along the toroidal direction have been observed on the divertor target plates, which not only affects the lifespan of PFCs but also limits the long-pulse and high-performance plasma operations. Therefore, investigating the toroidal distribution of the heat load on the divertor target in EAST is crucial for understanding the mechanism of damage on the target plate surface. The toroidal characteristics of heat loads on CMs of the lower divertor in EAST was analyzed using a three-dimensional magnetic field line tracing program (PFCFlux). For the inner vertical target (IVT) and outer vertical target (OVT) within a single CM, the heat load varies monotonically along the toroidal direction. In contrast, the heat load distribution on the outer horizontal target (OHT) is uniform, except at the chamfers, which is directly correlated with the target’s geometric structure and the magnetic field configuration. The surface heat load distribution exhibits toroidal non-uniformity, with the heat deposition pattern being generally consistent and displaying periodic variations. The toroidally uneven heat load distribution is attributed to the change in the magnetic field’s tilt angle. Additionally, the peak heat load is observed at the chamfer positions between cassettes, and an analysis is conducted on the relationship between peak heat load and the size of the chamfer. Smaller tilt angles result in lower surface heat load. Furthermore, a comparative analysis of peak heat loads at the chamfer under different plasma current conditions reveals that higher plasma currents generally result in increased peak heat loads. This investigation into the toroidal heat load distribution on the divertor surface contributes to understanding of the damage mechanism of W PFC as the target plate and provides valuable data for divertor design of future fusion devices.

Recent calculations of the lithium vapor box divertor using SOLPS have demonstrated that far Scrape-Off Layer (SOL) baffling and the corresponding far SOL lithium evaporation are unnecessary for divertor performance. The plasma, acting as a virtual baffle, is sufficient to block the neutral lithium from escaping the divertor region in a configuration known as the lithium vapor “cave.” Here, extensions of this PFC design simplification are presented which show that significantly less obtrusive PFCs can provide similar divertor performance. Differences in upstream concentration are predicted depending on the depth and size of the lithium emitting aperture, the shape of the private flux region (PFR) structure, and the location within the PFR of the deuterium puff. SOLPS predictions for (nLi/ne)LCFS vary between 0.020 and 0.049. For the first time, the effect of E×B drifts on predictions for a lithium vapor box divertor is also examined, with the resulting solutions having lower inner divertor heat flux but higher outer divertor heat flux for a given lithium evaporation rate. The differences between the PFC designs are minimized with drifts enabled, indicating design flexibility. The increased PFC design flexibility reported here, along with improved SOLPS modeling with drifts, are important steps towards a final design.

Divertor detachment is a promising method to solve the power exhaust problem in tokamak devices or even in future magnetic fusion reactors. In this work, a detachment experiment (HL-2A shot #38008) with mixed gas seeding (60% nitrogen and 40% deuterium) is simulated using the SD1D module in BOUT++. In the process from attachment to detachment, the target electron temperature and the target ion saturation current in simulations are found to be consistent with the experimental results measured by Langmuir probes on the target plate. In order to understand the underlying detachment mechanism on HL-2A, this work analyses the role of different particle species in the cases with different seeding rates. It shows that the plasma density varies little and the density of neutrals ( D and D 2 ) slightly decreases with the increase of nitrogen seeding rate, such that plasma–neutral interactions cannot effectively reduce plasma energy and plasma momentum in the divertor. The case with a high seeding rate predicts that increasing seeding rate cannot reach a target temperature lower than 2.5 eV , which is the required temperature for strong plasma–molecule interactions. Thus, the plasma–molecule interactions may not be important in the divertor during nitrogen seeding. This work also studies the parallel forces including (1) force due to the parallel electric field, (2) friction forces, (3) ion- and electron-thermal forces, and (4) collisional reactions (e.g. charge exchange recombination and ionisation). It is found that the friction force between nitrogen ions and other particle species (primarily D + ) is the dominant force towards the target, while ion- and electron-thermal forces are the dominant force pushing nitrogen ions back to upstream. Parallel forces determine the parallel distribution of nitrogen impurities, and therefore decide the region of nitrogen radiation. [ABSTRACT FROM AUTHOR]

Particle balance calculations are done for the seeding species N2, Ne, Ar, Kr as well as He with the aim of obtaining a realistic description of the divertor and core plasma impurity content. Experimental time traces of main plasma impurity densities are fitted by a single, time-independent parameter v z , in eff . This parameter represents the product of the impurity inward pinch in the pedestal, used for the description of gross fueling here, and the enrichment factor between the sub-divertor gas reservoir and the upstream separatrix. v z , in eff depends strongly on the first ionization energy as well as on the charge Z or mass of the impurity. The prevailing dependence of the enrichment values on the ionization energy suggests the importance of the relative impurity and deuterium ionization lengths in the divertor. Regression analysis of v z , in eff yields an expression for the impurity concentrations in the core and the divertor of ASDEX Upgrade which allows the prediction of the corresponding impurity densities and their divertor enrichment within a factor of 2 with only engineering parameters as input. A simple wall model has been introduced to take into account wall storage and release of impurities, e.g. for conditions of pre-loaded walls due to seeding in previous discharges. Wall effects are observed for all species considered, but wall storage turns out to be more important for N and He compared to Ne, Ar, Kr. Similar enrichment values are obtained for ELMy H-modes and EDA/QCE no-ELM regimes. A factor of approximately 1.4 reduction in enrichment is observed for divertor conditions for pronounced detachment with Ar and N2. The obtained analytical model for the core and sub-divertor impurity densities is well suited for integration into a discharge flight simulator or a real time state observer. [ABSTRACT FROM AUTHOR]

Calculations of the tungsten sputtering yields (the divertor material in the ITER tokamak) by He, Be, N, and O—impurity atoms in the plasma—were carried out at a collision energy of 0.010–100 keV using the Monte Carlo method. To calculate the trajectory of the incident particle, pair potentials obtained within the framework of density functional theory were used. These potentials were corrected for the parameters of the potential well obtained from spectroscopic measurements. The target consisted of tungsten randomly oriented crystals the size of one lattice constant. Next, the trajectories of the recoil particles were calculated using many-particle potentials calculated using density functional theory. Thermal vibrations of target atoms were taken into account. The vibration amplitude was taken to be 0.05 Å, which corresponded to room temperature. The strong dependence of the results on the shape of the surface potential barrier is shown, and the results are presented for two limiting cases of the surface state: a flat surface, when a planar surface potential barrier is realized, and a surface consisting of cones, when a spherical potential barrier occurs. In the experiment, the surface has some roughness, which depends on the experimental conditions. It is shown that the experimental results lie between the limiting cases we considered. Information was obtained on the average energy of sputtered atoms and angular distributions necessary for calculating the entry of impurities into the tokamak plasma. [ABSTRACT FROM AUTHOR]

Langmuir probe measurements in neutral beam injection (NBI) heated, low-confinement mode plasmas in JET ITER-like wall showed that the current to the divertor targets, Idiv, in helium (He) plasmas was up to 70% lower on the low-field side (LFS) than in otherwise identical deuterium (D) plasmas. The edge plasma density at which the rollover of Idiv occurred i.e. the onset of detachment, was 10% higher in He plasmas on both the LFS and high-field side (HFS). The density of Idiv rollover increases by 25% for He when the NBI power increases 1MW to 5MW. The total radiated power was similar in He and D plasmas for densities below the Idiv rollover. At densities above the Idiv rollover density, the total radiated power and power from within the separatrix are higher in He, reducing the power across the separatrix and subsequently Idiv,LFS. In He plasmas, the peak radiated power was observed within the confined region above the X-point in tomographic reconstructions from bolometry.

The magnetic system of an open trap usually includes expansion sections located between high-field magnetic mirrors and end surfaces that receive plasma. In the GOL-NB device, an arc plasma gun is located in one of the expanders, which creates a low-temperature starting plasma in the confinement area. The parameters of the surface plasma sheath affect the electrical connection of the confinement area with the walls and, thereby, affect the contribution of the line-tying effect to the plasma stability and the longitudinal energy losses from the trap. The experiments with additional hydrogen injection into the plasma gun were carried out at GOL-NB. We observed a radiating plasma formation detached from the surface, which visually corresponds to that in radiating divertors in tokamaks. In both standard and detached modes, decaying plasma existed near the receiving electrodes during the entire observation time after the discharge current was terminated. In the central trap of GOL-NB, some structures in the Fourier spectrogram of magnetic fluctuations manifest earlier in the detachment mode than in the standard mode and have lower frequencies. We associate these structures with the onset of interchange-like modes due to the loss of plasma stabilization by the line-tying to the conducting ends. The observed plasma response to the additional gas supply confirmed our understanding of the line-tying effect as the main factor stabilizing the plasma core in the initial phase of density accumulation in the central trap. [ABSTRACT FROM AUTHOR]

In divertor plasmas, atomic processes play a significant role in reducing the divertor heat load. In particular, the elastic scattering between ions and neutral particles can be characterized as a large‐angle scattering; this is in contrast with Coulomb scattering, which is dominated by small‐angle scattering. In a large‐angle scattering, a large fraction of the ion energy is transferred to the neutral particle, and the particle flight direction can be changed significantly. This process can generate additional particle transport perpendicular to magnetic field lines and affect plasma density profile in neutral‐rich divertor regions. However, in most edge plasma simulation codes, the elastic scattering with neutral particles is not taken into account in the direction perpendicular to the magnetic field lines. In this study, we performed a 2D orbital calculation without other drifts such as diamagnetic and E × B drifts to assess the effect of the elastic scattering on the density profile. The peak density is found to be decreased by 11.1 (6.3)% compared with the model that ignores the transport due to the elastic scattering, even at 3.0 (5.0) T magnetic field similar to that of the JT‐60SA (JA DEMO reactor) class. In addition, the ion transport can be expressed as a diffusion model by integrating the differential cross section over the large‐angle scattering range π/2−π$$ \pi /2-\pi $$. The proposed method may be easily introduced into any integrated codes to consider the elastic scattering. [ABSTRACT FROM AUTHOR]

Shielding of materials from the impact of powerful plasma flows is of undoubted interest for solving a wide range of fundamental and applied problems. For example, to protect the divertor from the effects of high-temperature plasma during current interruptions in tokamaks, nitrogen or neon gas injection is used directly in front of the surface of the divertor plates. Also, using a gas screen, it is possible to regulate the thermal load on materials when they are irradiated with powerful plasma flows in order to improve the performance characteristics of surface layers. The work investigated the influence of a nitrogen gas screen on the shielding of tungsten when exposed to powerful flows of hydrogen plasma and on the localization of tungsten vapor near the surface of the irradiated target. It has been established that a gas screen of nitrogen with an atomic density of 2 × 1017 cm–3 in front of a tungsten target reduces the thermal load on the target by 2‒2.5 times less than the maximum level of energy absorbed by tungsten under pulsed action of a powerful flow of hydrogen plasma without a gas screen. The dependence of the density of energy absorbed by the target on the density of particles in the screen was experimentally determined. It is shown that, under our experimental conditions, the presence of a nitrogen gas screen in front of the target prevents the propagation of tungsten over distances of more than 1 cm along the magnetic field lines toward the incident plasma flow. [ABSTRACT FROM AUTHOR]

This article describes the study of low pressure helium plasma with magnetic confinement at the experimental test setup at the Moscow Power Engineering Institute: plasma linear multicusp (PLM). This facility is intended for testing refractory materials and prototypes of elements of the first wall within the framework of development of the national fusion reactor (DEMO–FNS) and the International Thermonuclear Experimental Reactor (ITER). The facility provides the conditions of plasma impact on the surface of tested sample close to the parameters and regime of operation of tokamak divertor plates. The facility is a magnetic trap with minimum magnetic field on the axis, where the plasma is created by the flow of electrons moving from the directly heated tantalum cathode toward the anode. It is possible to create stationary helium plasma in the facility and to maintain it for several hours under constant discharge parameters: helium pressure in the chamber of 10–3–10–1 Torr, discharge current of 4–30 A, plasma column diameter of 35–40 mm, voltage drop across the discharge gap of 100–200 V. The thermal load on the surface of target introduced into the axial region of plasma column has reached 5 MW/m2. Optical emission spectroscopy is the main diagnostic tool in this work. The procedure for determining atomic concentrations from the data on the relative intensities of atomic spectral lines of metallic impurities is proposed in this work. [ABSTRACT FROM AUTHOR]

At the compact spherical Globus-M2 tokamak, a series of experiments was conducted to study the effect of the injection of nitrogen on the discharge parameters. The experiments were carried out in discharges in deuterium in the divertor configuration, and the auxiliary heating was performed by deuterium neutral beam injection. During the nitrogen seeding, a substantial decrease in electron temperature near the divertor was recorded as well as a sharp decrease of the heat flux onto the divertor plate, while the density and temperature of the main plasma changed insignificantly. Simulations by the SOLPS-ITER showed a satisfactory agreement with the experiment. [ABSTRACT FROM AUTHOR]

We present the first Thomson scattering (TS) measurements of electron temperature in the lower divertor of the Globus-M2 tokamak. The divertor TS diagnostics is designed for local measurements of the Te(z, t) in the range of 1–100 eV and ne(z, t) in the range of m–3. Parameters of the probing Nd:YAG laser are as follows 1064 nm/2 J/100 Hz/3 ns. The probing chord is launched vertically at R = 24 cm and covers areas of the inner leg, vicinity of separatrix and private flux region. Along probing chord of 110 mm, 9 spatial points were realized. Advanced filter polychromators were used to analyze Thomson scattering spectra. [ABSTRACT FROM AUTHOR]

When the local heat flux exceeds specified flux limit, tungsten PFC surfaces can be damaged, which is not acceptable for a reliable reactor operations. The divertor PFCs are typically designed for a specific heat flux limit usually assuming an average steady-state heat flux which is typically 5–10 MW/m2. However, in addition to steady-state heat flux, fusion reactor divertor PFCs could experience transient heat fluxes such as ELMs and/or other magnetic reconnection events which can deposit large transient heat fluxes onto the divertor PFCs. The transient divertor heat flux could be significantly larger than the steady-state heat flux which could damage the solid PFC surfaces. The divertor heat flux can be subjected to additional complications such as the uncertainties in the the divertor strike point heat flux projection. Moreover, there are additional experimental observations of non-axisymmetric power flux which can occur under non-axisymmetric magnetic perturbations. The liquid lithium (LL) PFCs is more resilient against such transient heat fluxes as they could evaporate LL as needed and the lost LL can be then replenished afterward. In this paper, we analyze a case for a transient divertor heat pulse of 1 MJ in 10 ms for a ITER-size reactor. This is a small perturbation (~ 0.1%) to the expected plasma stored energy compared to the previously analyzed case of 20 MJ heat pulse. Even with this relatively modest heat pulse, the LL surface undergoes ~ 100 °C temperature rise. However, the resulting LL surface heating without rapid cooldown mechanism could lead to excessive LL evaporation continuing well after the transient heat flux resulting in a significant Li injection of ~ 0.6 mol in about a 200 ms period. This amount of Li injection could cause plasma dilution and performance degradation. On the other hand, an active Li injection capability if optimized could prevent the LL surface temperature rise and thus reducing subsequent Li evaporation into the plasma by a factor of 7 compared to the passive LL PFC case. A crucial tool of active Li injection is a rapid response pellet injector which could inject light impurity pellets before the excessive heat flux could reach the divertor plate causing serious damage. A simple pellet ablation model suggests a favorable pellet deposition profile for smaller ~ 0.1 mm radius pellet with ~ 10–20 m/s velocity. Moreover, if it is possible to inject from the private flux region, the pellet injection efficiency into the high heat flux strike point region can be as high as 80% compared to ~ 50% for the injection from outer radius region. The pellet deposition efficiency can be further improved by designing a shell-pellet which can burst when a certain ablation fraction is reached. A possible implementation technique using an inductive pellet injector with a rapid time response of a few msec is proposed here which can be tested in NSTX-U. [ABSTRACT FROM AUTHOR]

Harsh heat load conditions on plasma-facing components (PFCs) in steady-state and transient phenomena (e.g., disruptions and ELMs) in DEMO fusion reactors question the feasibility of current approaches based on solid targets made of tungsten. This issue calls for the development of innovative plasma-facing components. Liquid metal PFCs with strong convection enhance heat removal capability and resilience after the transient phenomena. However, transporting liquid metal across magnetic fields gives rise to MHD drag. MHD drag for the case of uniform B, estimated analytically, is acceptable. Grad-B MHD drags with straight ducts could seriously drag the LM flow across non-uniform B. Expanding the duct along B and shrinking the duct in a perpendicular direction could make electromotive force |vBh| approximately constant along the duct and significantly reduces the grad-B MHD drag. Here v denotes the flow velocity along the duct, B is the magnetic field strength, and h is the vertical duct size. Three-dimensional simulations for internal and free surface thermo-MHD phenomena have demonstrated that the proposed duct design reduces the total pressure drop along the duct. [ABSTRACT FROM AUTHOR]

We report sputtering yields of Li+, H−, O−, and OHx− ion species from an Li–O–H surface for H, D, He, Ne, and Ar ion irradiation at 45° incidence in the energy range of 30–2000 eV. A Li film was deposited on a stainless steel target using Li evaporators in the LTX-β vessel, using the LTX-β Sample Exposure Probe (SEP), which includes an ultrahigh vacuum suitcase for transferring targets without significant contamination from air exposure. The SEP was used to transfer the Li-coated target from LTX-β to a separate Sample Exposure Station (SES) to perform ion exposure measurements. The SEP was also used for characterization of the Li-coated target utilizing X-ray photoelectron spectroscopy in a different chamber, showing that the lithium film surface was oxidized. Ion exposures were performed using an electron cyclotron resonance plasma source in the SES. Sputtered/ejected species were sampled by a quadrupole mass spectrometer with capabilities for detecting positive and negative ions, and an energy filter for determining the mean kinetic energy of the ejected ion species. All ion irradiations caused Li+ ions to be ejected, while causing impurity ions such as H+, H−, O− and OH− to be ejected. Measured ion energies of Li+ ions from a Li–O–H surface suggested that the typical sheath potential on the divertor surface can trap sputtered Li+ ions, which were previously reported as ~ 60% of total sputtered Li species from Li targets (Allain and Ruzic in Nucl Fusion 42:202, 2002). Hence, our results for the sputtering yields of ejected ion species and their associated ion energies from a Li–O–H surface indicates that lithium sputtering is suppressed and impurity removal is enhanced due to the sheath potential at the divertor surface for fusion reactor applications. [ABSTRACT FROM AUTHOR]

During reactor operation, the divertor must withstand unprecedented simultaneous high heat fluxes and high-energy neutron irradiation. The extremely severe service environment of the divertor imposes a huge challenge to the bonding quality of divertor joints, i.e., the joints must withstand thermal, mechanical and neutron loads, as well as cyclic mode of operation. In this paper, potassium-doped tungsten (KW) is selected as the plasma facing material (PFM), oxygen-free copper (OFC) as the interlayer, oxide dispersion strengthened copper (ODS-Cu) alloy as the heat sink material, and reduced activation ferritic/martensitic (RAFM) steel as the structural material. In this study, a vacuum brazing technology is proposed and optimized to bond Cu and ODS-Cu alloy with the silver-free brazing material CuSnTi. The most appropriate brazing parameters are a brazing temperature of 940 °C and a holding time of 15 min. High-quality bonding interfaces have been successfully obtained by vacuum brazing technology, and the average shear strength of the as-obtained KW/Cu and ODS-Cu alloy joints is ∼268 MPa. And a fabrication route for manufacturing the flat-type divertor target based on brazing technology is set. For evaluating the reliability of the fabrication technologies under the reactor relevant condition, the high heat flux test at 20 MW/m2 for the as-manufactured flat-type KW/Cu/ODS-Cu/RAFM mockup is carried out by using the Electron-beam Material testing Scenario (EMS-60) with water cooling. This paper reports the improved vacuum brazing technology to connect Cu to ODS-Cu alloy and summarizes the production route, high heat flux (HHF) test, the pre and post non-destructive examination, and the surface results of the flat-type KW/Cu/ODS-Cu/RAFM mockup after the HHF test. The test results demonstrate that the mockup manufactured according to the fabrication route still have structural and interfacial integrity under cyclic high heat loads.

The consequences of tungsten (W) damaging processes, such as cracking and melting, on divertor lifetime and plasma operation are high priority issues for ITER. A sustained melting experiment was conducted in WEST using a 2 mm deep groove geometry on the upstream mono-block (MB) to overexpose the sharp leading edge (LE) of the downstream MB. W-cracking has been evidenced for the first time with a very high spatial resolution infrared camera before tungsten melting was reached. These cracks develop when the monoblock temperature is about 2600 °C, thus higher than both ductile to brittle transition and softening threshold of tungsten, suggesting that these cracks are different from the ones observed in previous campaigns where brittle failure was involved, because of transient events on cold monoblock. Post-exposure analyses have been performed on the damaged monoblock, highlighting 12 main cracks on the LE, with a width varying from 33 µm to 77 µm, and an average spacing of 0.45 mm. Parallel heat flux about 90 MW/m2 has been derived from infrared temperature measurements, with a heat flux decay length on the target of 4 mm. The T-REX modelling code suggest here that with these thermal inputs, a crack can initiates due to thermal cycling without disruption, with a ductile failure, under 1 to 5 cycles for a tungsten DBTT varying from 400 °C to 500 °C.

An analysis of the results of measuring the erosion rates of W-fuzz samples with different fuzz thicknesses subject to arcing has been performed. Based on the examination of the structure of arc tracks, it has been concluded that a sharp increase in the erosion rate was accompanied by melting of the nanostructured layer. At the same time, numerous frozen droplets were observed at the boundary of the cathode spot track on the samples with a large fuzz thickness. Individual droplets were also found at some distance from the boundary. The experimental results have been interpreted theoretically using the explosive emission (ecton) model and the results of the analysis of thermal processes occurring in the cathode spot.

Using liquid metals confined in capillary porous structures (CPSs) as a plasma-facing component (PFC) could prolong the lifetime of the divertor in the high heat flux area. However, the high atomic number of tin (Sn) limits its acceptable fraction in the main plasma. Therefore, a crucial step in developing this concept is to test it in a tokamak environment, particularly in the diverted plasma region, e.g. ASDEX Upgrade (AUG). In this paper, the design of liquid tin module (LTM) is explained, and the testing in the high heat flux device GLADIS before its use in AUG is presented. The LTM was additively manufactured using selective laser melting, consisting of a 1.5mm porous layer tungsten (W) directly attached to a solid W bulk. The LTM has a plasma-facing area of 16×40mm2 and was filled with 1.54g of Sn. In GLADIS, the module was exposed to power loads between 2 and 8MWm−2 for 1 up to 10s, first unfilled and later filled with Sn. The surface temperature was monitored with infrared imaging and pyrometry. The thermal response was used to compare with simulations in Ansys Mechanical, enabling a determination of the module’s effective thermal properties. Sn droplets could be observed on the infrared camera, until a surface temperature of about a 1000°C was reached. The enhanced wetting of tin on the plasma-facing surface, which was observed by a visible camera, suggests that there is a conditioning of the surface, possibly due to the removal of impurities and oxides. Subsequent examinations of the adjacent tile revealed minor Sn leakages emanating from the module’s edge. Furthermore, the module showed no indication of mechanical failure. Therefore, these results indicated that the LTM qualifies for the heat fluxes expected in ASDEX Upgrade.

One of the ways to extend the lifetime of the divertor for DEMO could be to replace the solid tungsten plasma-facing components with liquid tin (Sn) confined in a tungsten capillary porous structure (CPS). Testing a CPS in a divertor plasma configuration is crucial for the development of a liquid metal divertor (LMD) to understand how the main plasma is affected. Only a limited Sn concentration is allowed in the plasma core, due to the high radiative losses associated with the high atomic number of Sn (50). Therefore, it is necessary to test a small-scale LMD filled with Sn in a tokamak environment, which has not previously been done. In ASDEX Upgrade, a liquid tin module (LTM) has been exposed by means of the divertor manipulator. During plasma flat-top, the outer strike point (OSP) was placed onto the pre-heated LTM and held there for a time interval between 2 and 3.4s over multiple discharges. Photographs of the LTM taken after each discharge, revealed macroscopic Sn leakage onto the adjacent tile. Simulations with the HeatLMD code predicted an acceptable tin erosion near the LTM with thermal sputtering dominating over evaporation. However, spectroscopic measurements revealed an order of magnitude higher erosion. Since this remained constant when the OSP was held on the LTM so that the surface temperature increased, evaporation could be excluded as the main source of Sn erosion. Comparison between discharges with different durations of OSP location on the LTM revealed an increase in core radiation up to 1.5MW due to Sn. The 1.5D-impurity transport code STRAHL was used to interpret this increase in total plasma radiation and revealed a Sn concentration in the main plasma of up to 1.4×10-4. Given that the LTM only covered about 1/650 of the outer divertor circumference, extrapolating to a full toroidal divertor implies erosion is above acceptable limits. The unexpectedly high Sn fraction in the main plasma is attributed to the ejection of Sn droplets reaching the main plasma, which may have originated from either the CPS or leaked tin. This conclusion is also supported by splashes of tin droplets, which were observed on the adjacent divertor tile and one ∼0.5m downstream. Therefore, to make a Sn-filled LMD a viable alternative to solid tungsten, the formation of droplets must be reduced by two orders of magnitude.

EAST is the first tokamak to feature fully actively water-cooled ITER-like W/Cu monoblocks on the divertor target. In 2014 and 2021, the graphite tiles in the upper and lower divertor were upgraded to W/Cu monoblocks, respectively. With the increase in plasma parameters, severe melting phenomena were inspected at the leading edges of upper divertor from 2017 to 2020. Through theoretical analysis and numerical simulation, two main factors, radial misalignment and chamfer structure have been identified to mainly impact the leading-edge induced melting of MBs. It was found that assembly misalignments increased gradually during the plasma discharges from 2015 to 2017 and the maximum misalignment can reach up to 3 mm, which play an important role that is responsible for the melting during 2018 to 2020. In 2020, engineering assembly improvements were implemented, resulting in a significant improvement compared to the previous situation, with all misalignments brought below 1 mm, and thus effectively prevent melting at the leading edge on upper divertor. In addition, the modification of chamfer structure to 1.5 × 17 mm at inter-CMs of the lower divertor further mitigated the issue of leading-edge-induced melting compared to upper divertor. Such experiences and lessons from the mitigation of leading edge induced melting provide important references for future fusion devices.

Wendelstein 7-X, the world’s largest superconducting stellarator in Greifswald (Germany), started plasma experiments with a water-cooled plasma-facing wall in 2022, allowing for long pulse operation. In parallel, a project was launched in 2021 to develop a W based divertor, replacing the current CFC divertor, to demonstrate plasma performance of a stellarator with a reactor relevant plasma facing materials with low tritium retention. The project consists of two tasks: Based on experience from the previous experimental campaigns and improved physics modelling, the geometry of the plasma-facing surface of the divertor and baffles is optimized to prevent overloads and to improve exhaust. In parallel, the manufacturing technology for a W based target module is qualified.This paper gives a status update of project. It focusses on the conceptual design of a W based target module, the manufacturing technology and its qualification, which is conducted in the framework of the EUROfusion funded WPDIV program. A flat tile design in which a target module is made of a single target element is pursued. The technology must allow for moderate curvatures of the plasma-facing surface to follow the magnetic field lines. The target element is designed for steady state heat loads of 10 MW/m2 (as for the CFC divertor). Target modules of a similar size and weight as for the CFC divertor are assumed (approx.

Divertor power load is a major challenge towards a burning plasma in a next-step tokamak. Here, the first results of a divertor power load characterisation in tritium plasmas in type-I ELMy H-mode, obtained in the JET deuterium-tritium campaign (DTE2) performed in 2021, are presented. It is demonstrated that both, transient loads due to type-I ELMs as well as the power fall-off length, do not exhibit an explicit ion mass dependence, with remarkably similar values in the tritium plasmas and in the deuterium references. This gives an improved credence to published scaling law predictions, solely based on deuterium plasma experiments. Moreover, the type-I ELM impact on the inner divertor target is studied in deuterium discharges. A slightly increased parallel energy fluence on the inner target with a factor of 1.08 compared to the outer target is observed. This is explained by the smaller major radius of the inner target. [ABSTRACT FROM AUTHOR]

Material issues pose a significant challenge for the design of future fusion reactors. These issues require new advanced materials to be developed. W‐fiber‐reinforced W‐composite material (Wf/W) incorporates extrinsic toughening mechanisms increasing the resistance against failure and thus granting steps toward application in a future fusion reactor. Wf/W can be produced based on chemical vapor deposition or powder metallurgical routes. In this contribution, the efforts of upscaling the production of Wf/W will be reviewed based on recent results. In addition, the activities related to enabling large‐scale production for new fusion applications are being studied. Herein, two main achievements are to be highlighted. First, an upscaled production is established to produce flat tile samples for joining tests on copper and steel, and second, a new method of joining Wf/W on copper is established and tested under high heat‐flux conditions. [ABSTRACT FROM AUTHOR]