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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.

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]

To reliably predict the distribution of heat and particle fluxes at the target plates of tokamaks, a comprehensive understanding of turbulence throughout the entire Scrape-Off-Layer (SOL) is imperative. This study examines divertor turbulence systematically across a broad parameter range on the TCV tokamak, including variations in magnetic field direction, plasma current $I_{\mathrm{p}} \in [140, 320]$ kA, edge safety factor $q_{95} \in [2.6,4.7]$ and Greenwald fraction $f_{\mathrm{G}}\in [0.18,0.6]$ . The TCV X-point Gas Puff Imaging (GPI) system is used to measure 2D filament properties in the inner and outer divertor region. The fluctuation levels in the divertor are found to strongly increase with density (to 80% over most of the SOL) while remaining insensitive to $I_{\mathrm{p}}$ . The previously identified divertor-localized filaments (DLF), located on the bad curvature side of the outer divertor leg, are found to be a common feature on TCV, while no filaments are observed in the PFR. DLFs are present over most of the parameter space and in both field directions. However, they are absent, or appear only closer to the target, for sufficiently large $\Lambda_{\mathrm{div}}\gtrsim 10$ or $q_{95}\gtrsim3.7$ . Across both $I_{\mathrm{p}}$ and $f_{\mathrm{G}}$ scans, some clear trends with $\Lambda_{\mathrm{div}}$ are found for divertor filament sizes and velocities, and with target fall-off lengths of density and heat flux profiles at the outer target. This study provides important experimental insights to turbulent transport in the divertor also for comparison with self-consistent, turbulence simulations and extrapolation to future reactor conditions.

The contributions in magnetic confinement fusion theory presented at the 29th Fusion Energy Conference (FEC 2003) are summarized here. This summary aims at providing an overview of the advances in the field and new directions in integrated modeling, computational physics, control design and application of artificial intelligence to discharge design and optimization. Given the increasing interest in bringing fusion to the grid and the focus on design and discharge optimization by means of self-consistent simulations that integrate physics and engineering with a hierarchy of fidelity models, the layout of this summary highlights applications over fundamental theory.

The predicted divertor conditions for the SPARC tokamak are calculated using SOLPS-ITER for a range of scrape-off-layer (SOL) heat flux widths λ _q , input powers, and particle fueling locations. Under H-mode scenario conditions with an upstream separatrix density of 1 $\times \mathrm{10}^{20}\:$ m ^−3 , the most conservative range of λ _q extrapolations ( ${\approx}$ 0.15 mm) results in extremely high unmitigated particle and energy fluxes to the divertor, both under full field (12.2 T) and power ( $P_{\mathrm{SOL}}$ = 29 MW) conditions, and 2/3 field with $P_{\mathrm{SOL}}$ = 10 MW. Increasing the cross-field SOL diffusivities by 2–10× reduces the magnitude of the mitigation challenge, however strategies such as impurity seeding or strike-point-sweeping will likely still be required. A combination of steady-state and time-dependent SOLPS-ITER simulations are used to map out phase space diagrams of upstream and divertor conditions. The simulations include parallel currents but neglect cross-field drifts. At low upstream density the inner and outer divertor conditions are highly asymmetric, with a large temperature difference and significant heat fluxes driven by parallel currents. The solution has sharp bifurcations with a region of hysteresis, depending on whether the initial state is at a low or high density. This behavior is observed even when the fueling location, cross-field diffusivity, and impurity level is changed, although the density window with asymmetry is reduced with increasing diffusivity. The addition of neon impurity seeding reduces the divertor heat fluxes, but also causes a drop in the upstream electron density with fixed particle throughput. This drop can be counteracted by increased main ion throughput, however too much neon results in a back transition into the asymmetric divertor regimes suggesting a need for control of both main ion and impurity seeding levels to achieve a desired divertor state.

Sequential SOLPS-ITER and DIVIMP simulations are carried out for tungsten (W) edge transport during neon (Ne) injection in EAST, elucidating the pathway of W impurity leakage from the divertor to the core plasma under different dissipative divertor conditions. Divertor conditions from low-recycling to detachment are generated by modulating the Ne injection rate in the SOLPS simulation. With the drift velocities incorporated in DIVIMP, W transport under the drift effect is investigated. For low-density L-mode plasma conditions with favorable B _t direction, the majority of W source originates from the outer divertor target and leaks upstream through the near-SOL E × B drift reversed region at both inner and outer divertors. The E × B drift is demonstrated to have a significant effect on W leakage and the effect diminishes with the increase of Ne injection rate. Compared to the D _2 injection cases, Ne injection helps to achieve a strong dissipative divertor condition at a much lower plasma density, resulting in a higher W leakage ability due to the smaller friction with the background plasma. For Ne injection cases under high-density H-mode conditions, W leaks mostly through the outer divertor region while the inner divertor is fully detached. Compared to the L-mode cases, the shorter power decay lengths in the scrap-off layer (SOL) of the H-mode cases result in smaller parallel E × B drift velocities especially in the middle and far SOL region. With the increase of the Ne injection rate, the decrease of the E × B drift in the middle SOL leads to a wider near-separatrix W leakage path. Therefore, a worse divertor W screening is expected, which is consistent with the previous published experimental observations (Ding et al 2022 Nucl. Mater. Energy 33 101250).

This study demonstrates that using spin-polarized deuterium-tritium (D-T) fuel with more deuterium than tritium can increase tritium burn efficiency (TBE) by at least an order of magnitude without compromising fusion power output, compared to unpolarized fuel. Although previous studies show that a low tritium fraction can enhance TBE, this strategy resulted in reduced fusion power density. The surprising improvement in TBE at fixed power reported here is due to the TBE increasing nonlinearly with decreasing tritium fraction but the fusion power density increasing roughly linearly with D-T cross section. A study is performed for an ARC-like tokamak producing 481 MW of fusion power with unpolarized 53:47 D-T fuel, finding the minimum startup tritium inventory ( $I_{\mathrm{startup,min}}$ ) is 0.69 kg. By spin-polarizing half of the fuel and using a 60:40 D-T mix, $I_{\mathrm{startup,min}}$ is reduced to 0.08 kg, and fully spin-polarizing the fuel with a 63:37 D-T mix further reduces $I_{\mathrm{startup,min}}$ to 0.03 kg. Some ARC-like scenarios are predicted to achieve plasma ignition with relatively modest spin polarization. These findings indicate that, with advancements in helium divertor pumping efficiency, TBE values of approximately 10%–40% could be achieved using low-tritium-fraction and spin-polarized fuel with minimal power loss. This would dramatically lower tritium startup inventory requirements and reduce the amount of on-site tritium. More generally than just for spin-polarized fuels, increased plasma performance can be used to increase TBE. This strongly motivates the development of spin-polarized fuels and low-tritium-fraction operation for burning plasmas.

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.

The ITER divertor design and performance assessment, primarily based on the SOLPS-4.3 burning plasma database (Pitts R. et al 2019 Nucl. Mater. Energy 20 100696), assumes the use of beryllium (Be) as the divertor surface material and the injection of gas from the main chamber top. However, the current ITER baseline favors gas injection from the more toroidally symmetric sub-divertor region. This paper evaluates the implications of these assumptions for divertor performance in the ITER fusion power operation phase. The impact of the divertor surface material and the gas injection location on the main ions mirrors the hydrogen only low power phase scenario shown in Park J.-S. et al (2020 Nucl. Fusion 61 016021). However, during burning plasma operation, extrinsic impurity seeding will be required. In the case of neon (Ne), studied here, impurity retention is influenced by both the divertor surface material and the fueling location. Neon leakage increases due to more energetic reflection from tungsten than beryllium, but equivalent divertor performance can be achieved by adjusting the neon seeding rate. While the impurity seeding location does not affect the distributions of impurity or radiation, the fueling location does. Top fueling provides local ionization sources mainly in the mid-SOL under detached conditions, enhancing divergences of the flux there (source-driven flow), bringing stagnation points close to the fueling location, and equilibrating flows towards both targets. In contrast, the global flow pattern (in the absence of fluid drifts) in the case of sub-divertor fueling is biased towards the inner target. Impurity flows, driven by force balance, largely mirror those of the main ion flow, including the stagnation point. The case with top fueling enhances Ne retention and corresponding radiation in the outer divertor, effectively reducing the total and peak target heat fluxes by 20%–40%, compared to the case with divertor fueling. Meanwhile, the case with outer target fueling also achieves similar reductions by enhancing plasma-neutral interactions. These results suggest the possibility that the selection of the fueling location and throughput can be used as an actuator to control impurity divertor retention and divertor radiation asymmetry.

In the Large Helical Device (LHD) a low temperature mode (LTM) of the helical divertor was discovered. It combines particle detachment and very large sub-divertor pressures up to 1.4 Pa. During the LTM, the electron temperature in the divertor was in the range from 0.25 to 0.42 eV so that volume recombination occurred. This result is remarkable because in the stellarators LHD and Wendelstein 7-X only low sub-divertor pressures (0.03–0.3 Pa) were expected and measured up to now due the loss of pressure conservation along flux tubes by an enhanced cross-field transport. It demonstrates that the more complex, three-dimensional divertors of stellarators can achieve a similar performance with respect to particle exhaust and detachment like the geometrically simpler poloidal divertors in tokamaks even if the favorable effect of flux amplification is absent. The LTM of the helical divertor depends, however, on the magnetic configuration, i.e. on geometry. It was only observed in the inward shifted configuration with $R_\mathrm{ax} = 3.55$ m, but not in the more frequently studied configuration with $R_\mathrm{ax} = 3.6$ m. A nuclear fusion reactor based on the heliotron concept (DEMO) would benefit from the LTM by the very compact divertor configuration and the excellent performance.

A new drift module has been developed in the DIVIMP code, enabling two-dimensional simulations of tungsten (W) transport in the edge plasma with full drifts included. By using the SOLPS-DIVIMP code package, the impact of drifts on W transport and screening has been investigated for various levels of dissipative divertor conditions and different divertor geometric configurations in EAST. Simulation results reveal that $\vec E \times \vec B$ drifts can enhance the W leakage by more than one order of magnitude. Under the favorable B _t direction, W erosion mainly occurs on the outer divertor target, making the W leakage from the outer divertor region the dominator. A leakage path from the near-scrape-off layer (SOL) region is revealed by the modeling results. In the leakage path, both the ion temperature gradient force and the reversed poloidal $\vec E \times \vec B$ drift are pointing upstream. With the radial $\vec E \times \vec B$ drift pushing W ions from the well-screened far-SOL region to the near-SOL region, the leakage from the near-SOL region becomes significant. As the divertor condition varies from the low-recycling regime to the deep detachment regime, the decrease of the ion temperature gradient velocity and poloidal $\vec E \times \vec B$ drift velocity narrows the width of the near-SOL leakage tunnel and thus enhances W screening. While under the unfavorable B _t , W erosion and leakage from the inner divertor target matters, the leakage mechanism especially the leakage path from the near-SOL region is similar as the favorable B _t cases. Furthermore, the effect of different divertor geometries on the W screening has been investigated. The configuration with the outer strike point (OSP) on the horizontal divertor plate is proved to narrow the near-SOL leakage tunnel, and thus the unreversed poloidal $\vec E \times \vec B$ drift pointing to the divertor target dominates and helps to enhance the divertor W screening. For the same D _2 puffing rate, the W leakage ability of cases with the OSP on the horizontal target can be more than 10 times weaker than the cases with the OSP on the vertical target, especially when the divertor is detached.

The Super-X Divertor (SXD) is an alternative divertor configuration leveraging total flux expansion at the Outer Strike Point (OSP). While the extended 2-Point Model (2PM) predicts facilitated detachment access and control in the SXD configuration, these attractive features are not always retrieved experimentally. These discrepancies are at least partially explained by the effect of parallel flows which, when self-consistently included in the 2PM, reveal the role of total flux expansion on the pressure balance and weaken the total flux expansion effect on detachment access and control, compared to the original predictions. This new model can partially explain the discrepancies between the 2PM and experiments performed on tokamak à configuration variable (TCV), in ohmic L-mode scenarios, which are particularly apparent when scanning the OSP major radius Rt. In core density ramps in lower Single-Null (SN) configuration, the impact of Rt on the CIII emission front movement in the divertor outer leg—used as a proxy for the plasma temperature in the divertor—is substantially weaker than 2PM predictions. Furthermore, in OSP radial sweeps in lower and upper SN configurations, in ohmic L-mode scenarios with a constant core density, the peak parallel particle flux density at the OSP is almost independent of Rt, while the 2PM predicts a linear dependence. Finally, analytical and numerical modeling of parallel flows in the divertor is presented. It is shown that an increase in total flux expansion can favour supersonic flows at the OSP. Parallel flows are also shown to be relevant by analysing SOLPS-ITER simulations of TCV.

Particle balance calculations are done for the seeding species N _2 , 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 $_{z,\textrm{in}}^\textrm{eff}$ . 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 $_{z,\textrm{in}}^\textrm{eff}$ 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 $_{z,\textrm{in}}^\textrm{eff}$ 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 N _2 . 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.

This paper presents a quantitative validation of Scrape-Off Layer Plasma Simulation-ITER (SOLPS-ITER) simulations against the TCV-X21 reference case and provides insights into the neutral dynamics and ionization source distribution in this scenario. TCV-X21 is a well-diagnosed diverted L-mode sheath-limited plasma scenario in both toroidal field directions, designed specifically for the validation of turbulence codes (Oliveira, Body et al 2022 Nucl. Fusion 62 096001). Five new, neutrals-related observables are added here to the extensive, publicly available TCV-X21 dataset. These are three deuterium Balmer lines in the divertor and neutral pressure measurements in the common and private flux regions. The quantitative agreement metric used in the validation is combined with the conjugate gradient method to approach the SOLPS-ITER input parameters that return the best overall agreement with the experiment. A proof-of-principle test of this method results in a modest improvement in the level-of-agreement; the shortcomings impacting the result and how to improve the methodology are discussed. Alternatively, a scan of the particle and heat diffusion coefficients shows an improvement of 10.4% in the level-of-agreement, approximately twice as high as that achieved by the gradient method. This result is found for an increased transport coefficient compared to what is usually used for TCV L-mode plasmas. The simulations further indicate that ∼65% of the total ionization occurs in the SOL, from which ∼70% in the divertor regions, despite being a sheath-limited regime, motivating the inclusion of self-consistent neutral models in future turbulence simulations on the path towards improved agreement with the experiment.

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 $H_{98-y2}\sim 1$ and reactor-relevant normalized pressures $\beta_\mathrm N \sim 2$ . 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 $\times \nabla$ B ) on access to detachment was demonstrated by the lower upstream density needed to access detachment with ion B $\times \nabla$ B drift directed outside of the active divertor (Greenwald fraction $f_{\mathrm{Gw}}\sim$ 0.9–1.0 vs $f_{\mathrm{Gw}}\sim$ 1.3). The upstream density at detachment onset was observed to increase linearly with plasma current with ion B $\times \nabla$ 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.

Experiments in DIII-D demonstrate that the upstream plasma density to detach an un-pumped slot divertor is similar for a V-shaped and a flat-end slot, despite significantly higher neutral pressure in the V-shaped slot and in contrast to SOLPS-ITER predictions. The detachment threshold can be reduced by using in-slot instead of main-chamber gas fuelling or by placing the strike point on the inner slanted slot baffle instead of the slot end, as described by simulations with full drift physics. When increasing the plasma line-averaged density (without extrinsic impurities), the transition to detachment in DIII-D slot divertor is sharp and requires a high value of plasma density with the ion $\overrightarrow {\mathbf{B}} {{ \times }}\nabla B$ drift into the slot, whereas it is smooth and requires a lower value of plasma density with the opposite drift direction, in accord with detachment experiments in the DIII-D open lower divertor. Unique experiments on DIII-D and comparison to advanced simulations expand the scientific understanding of slot-shaped divertors, considered highly desirable for next step fusion devices.

An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170 GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, X-divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

Commercial fusion reactors will be faced with extremely high divertor target heat fluxes that will require mitigation. Simulations of detachment in an NSTX-U scenario projected to have 92 MW m ^−2 unmitigated peak target heat flux are presented, which reaches sub-10 MW m ^−2 target heat flux using a highly dissipating lithium vapor box divertor design. The lithium vapor box is a detached divertor design which employs lithium vapor evaporation and condensation to contain lithium below the X-point. Previous SOLPS modeling has indicated a lithium vapor box can reduce the heat flux down to 10 MW m ^−2 via simultaneous evaporation from the Private Flux Region (PFR) and the Common Flux Region (CFR) sides of the vapor box. It is found here that PFR evaporation has improved access to the separatrix leading to significantly more efficient power dissipation than CFR evaporation. Simulations of target evaporation with an evaporation distribution that is self-consistent with the temperature of a Capillary Porous System with Fast flowing liquid lithium could reach $n_\textrm{Li}$ / $n_\textrm{e} \sim$ 0.025–0.030 at the Last Closed Flux Surface (LCFS) depending on the liquid metal flow speeds and lithium sputtering yield, while PFR-side evaporation can reach acceptable heat fluxes with $n_\textrm{Li}$ / $n_\textrm{e} \sim$ 0.038 at the LCFS. However, PFR evaporator performance can be improved if the target is allowed to be hot enough such that it reflects lithium, reaching $n_\textrm{Li}$ / $n_\textrm{e} \sim$ 0.028 and reducing required lithium evaporation. Ultimately PFR evaporation and target evaporation are found to have similar ability to produce acceptable heat flux solutions with minimal upstream concentration.

Introduction of cross-field drifts in SOLPS-ITER simulations of connected double-null plasmas in STEP with the ion $\nabla B \times B$ drift towards the upper divertors was found to enhance the detachment of the inner divertors with decreased target densities and ion and heat fluxes, while simultaneously complicating the access to detachment in the outer lower divertor by increasing the target temperature and heat loads to levels above the engineering limits. The $\nabla B \times B$ drift was observed to significantly affect the radial heat transport between the core and the scrape-off layer (SOL), altering the poloidal temperature and pressure profiles and, consequently, the poloidal conductive and convective heat transport in the SOL. As a result, up-down asymmetries of 52:48 and 58:42 biased towards the outer lower and upper inner divertors, respectively, were observed to arise in the unmitigated power entering the divertor regions, breaking the up-down symmetry seen in simulations without the drift terms and contradicting with earlier experimental observations on the low-field side. Moreover, the upstream electron density was found to decrease noticeably in the core and separatrix regions following the activation of the drifts due to an increased share of the neutrals arriving from D _2 injections near the upper and lower X-points ionizing already in the private flux regions.

A thermography inversion algorithm has been developed in the open-source Python-based computer code, HYPERION, to calculate the heat flux incident on plasma-facing components (PFCs) in axisymmetric tokamaks. The chosen mesh size at the surface significantly affects the calculated transient heat flux results. The calculated transient heat flux will exceed the real value when the mesh size tends to zero but will underestimate the real value when the mesh size is large. A criterion for determining the appropriate mesh size for the transient heat flux calculation will be discussed. The numerical scheme for HYPERION uses a 2D fully implicit finite-difference approach, allowing temperature-dependent thermal properties of PFC materials. The inversion algorithm is benchmarked against established heat flux calculation codes, TACO and THEODOR, based on thermography data from NSTX and DIII-D respectively. The primary benefits of HYPERION compared to TACO and THEODOR are that it is open-source and it allows for the optimization of mesh thickness along the substrate. The algorithm also accounts for the thermal properties of thin surface layers that characteristically form on PFCs due to plasma-material interactions. The agreement between HYPERION and THEODOR is excellent, as the percent difference between the codes is ∼5% on average in the case of the DIII-D data for moderate to high heat flux. Verification tests with TACO show slightly higher average percent differences of 8% and 12%. In using HYPERION to study filaments in heat flux, the initial results indicate that small ELMs filaments significantly broaden the divertor heat flux, and decrease divertor peak flux. Compared to the inter-ELM, the small ELM filaments decrease the divertor peak surface temperature. With intermittent divertor filaments, the divertor heat flux width is comparable with that found in L-mode.

The heat flux width ( λ _q ) 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 λ _q has not been well understood. Investigations of the influence of drifts on λ _q , 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 λ _q in the outer/inner side and decrease the λ _q 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 λ _q are different. The upstream λ _q ( λ _q _,u ) is directly affected by the drift-induced convective heat flux, while λ _q at the target ( λ _q _,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 λ _q _,t in the inner side at high density.

The micro-trench method is a new technique to measure polar and azimuthal incident ion angles, material erosion, and impurity deposition at plasma-facing surfaces, and has been applied in the DIII-D divertor. This article gives a tutorial of the micro-trench technique consisting of six steps: (1) micro-trench fabrication, (2) pre-exposure measurement of the fabricated micro-trench geometry, (3) tracer material deposition, (4) plasma exposure, (5) post-exposure observation of the tracer material, and (6) post-exposure measurement of the micro-trench geometry. Two criteria need to be satisfied to apply the micro-trench method successfully: (i) uniform impurity deposition on the micro-trench floor, and (ii) erosion dominated by the physical sputtering induced by the impinging ions. When those two criteria are satisfied, post-exposure analysis of the impurity deposition patterns on the micro-trench floor may be used to determine the polar and azimuthal incident ion directions (mean values of the ion angle directions), erosion rate, and impurity deposition rate during plasma exposure without computational interpretation. [ABSTRACT FROM AUTHOR]

• Advanced materials are selected. W-ZrC material exhibits high strength, high ductility, high thermal conductivity, low DBTT, excellent resistance to high thermal-loads and to high flux plasma etching as well as enhanced high-temperature stability. • A rarely occurred internal ablation failure mode emerged during the high heat flux (HHF) test to verify the feasibility of W-ZrC divertor. • According to the shape and size of the ablation pit, through thermal simulation analysis, it can be concluded that result of the suddenly and significantly increased heat load combining the reduction of water flow thermal ablation will be occurred inside of the heat sink and lead to serious consequences for divertor. As a plasma-facing material (PFM), W–ZrC is a promising candidate for the divertor in future fusion reactors. To verify the feasibility of W–ZrC, a high heat flux (HHF) test was conducted on a comprehensive experimental platform. A rare internal ablation failure mode was observed during the experiment. Considering the internal ablation phenomenon during the test, the measured data of the test platform were analyzed to predict an ablation time of 12 s. Based on the shape and size of the ablation pit, the cause and mechanism of ablation were determined using thermal simulation analysis. It can be concluded that a sudden and significant increase in heat flux combined with reduced water flow can cause thermal ablation to occur within the heat sink. Furthermore, dimensions of the ablation pit are related to the area in which the heat flux is imposed. In sum, internal ablation can have serious consequences for the divertor. [ABSTRACT FROM AUTHOR]

The feasibility of a radiatively cooled 3D-printable liquid metal heat pipe (HP) design is assessed. Using the design flexibility offered by 3D-printing, the design of the wick and geometry of the HP were optimised to meet the requirement of 20 MW / m 2 heat load for a HP placed in a 1.5 T magnetic field. COMSOL was used to assess the operational limits of the HP, the thermal stresses in the wall, the thermally radiated power, and various materials for the HP. The main parameters are the diameter and spacing of the screen wires and the emissivity, 200 μ m , 200 μ m and 0.86 respectively. Molybdenum was chosen as the wall material and lithium as the working fluid. The design was made in Siemens NX and then exported to COMSOL. From simulations it was concluded that a molybdenum HP with the final design was capable of handling a steady state heat load of 20 MW / m 2 . • The heat pipe design can withstand a peak steady state heat load of 20 MW/m2. • Heat pipe design possible due improvements in additive manufacturing of molybdenum. • The next step is to create and test the heat pipe in Magnum-PSI. • In the future, heat pipe divertor concepts can significantly improve replaceability. [ABSTRACT FROM AUTHOR]

• Vacuum pressure in the HL-3 vacuum vessel reached to 1.8 × 10-6 pa. • The temperature of the VV is baked by nitrogen to 176 °C. • A solution based on control variable method is used to adjust the temperature difference between the divertor and the VV. • A 100 % zero oil-free, high speed centrifugal fan is developed for the HL-3 baking system. HL-3 is a tokamak device constructed in 2020 and it has achieved remarkable results in a short period of time. Baking is an important way to obtain a high vacuum required for plasma discharge in tokamak devices. A nitrogen gas closed-circuit reheat baking system is utilized to concurrently bake the vacuum vessel (VV) and the lower divertor of HL-3. So far, three rounds of baking experiments have been carried out, with vacuum pressure up to 1.8 × 10–6 Pa and the results from the mass spectrometer indicated that the amount of compounds were greatly reduced in the residual gas. Besides the design description and experimental results of HL-3 baking system, this paper also introduces a simulation-based solution for addressing the issue of the lower divertor temperature being significantly lower than that of the VV after baking to the high temperature stage. [ABSTRACT FROM AUTHOR]

• JA-DEMO divertor concept with double-coolant circuits of 200°C and 290°C pressurized waters was developed. • Plasma Facing Unit support design for the coolant distribution was optimized to reduce variation of the flow velocity less than 5, 6 %. • A new cooling concept with two layers of puddles and fins at the side routes was proposed for CB. • Heat exhaust on fish-scale surface monobolock target and stress-strain of the CuCrZr pipe were investigated under assuming a DEMO divertor condition. Engineering design of the JA-DEMO divertor concept, i.e. double-coolant circuits of 200°C coolant (5MPa) for high heat load targets (CuCrZr-pipe) and 290°C coolant (15MPa) for high neutron load Plasma Facing Units (F82H-pipe) and cassette body (CB), has been developed. Computational fluid dynamics (CFD) calculation of the coolant distributions to targets, baffles, reflectors, dome and to CB was performed in order to determine the feasibility of the key concepts. (i) All PFU support designs for the coolant distribution to PFUs were optimized to reduce variation of the flow velocity (V cool) at the inlet of PFUs less than 5 – 6 %. Parallel coolant circuit design for the dome and reflectors was also developed, and mass flows were adjusted by orifices and the main mass flow rate. (ii) A new cooling concept with two layers of puddles and fins at the side routes was proposed for CB. The design provided V cool = 0.55 – 1.27 m⋅s−1 and average V cool = 1.04 m⋅s−1 with the fin transparent ratio of 0.1. It was enough to exhaust the nuclear heat, and the design issues were identified under the coolant condition. (iii) Heat exhaust on fish-scale surface target and stress-strain of the CuCrZr pipe were investigated under assuming a DEMO divertor condition. Steady-state heat load at wet region by plasma (q t wet) was restricted below 13.5 MWm−2 to avoid W-recrystallization. In the stress-strain cycle of higher q t wet ∼15.3 MWm−2, maximum tensile σ (120 – 200 MPa) and total change in strain during the heat load cycle (Δε ∼ 0.25 %) were relatively small. These evaluations suggested that such high heat load cycle may not be a critical lifetime issue, but reduction in T cool is preferable to handle ITER-like slow transients such as 15 – 20 MWm−2. [ABSTRACT FROM AUTHOR]

• Low detectability of high-frequency ultrasonic tests for original screw pipe. • Interference of the echoes from the interface and original screw pipe's inner surface. • Adjusting the height of the grooves in the screw pipe reduces echoes' interference. • Employing step-like grooves in the screw pipe improves the detectability of the tests. This study investigated the applicability of high-frequency ultrasonic testing to inspect the tungsten-copper bonded interface between a divertor's mono-block and a cooling pipe called the "screw pipe." Mock-up samples simulating the interface were fabricated by diffusion bonding of a 2.5 mm thick oxygen-free copper plate to a 12 mm thick tungsten block that satisfied the ITER material specifications. After making the bond, grooves were machined on the oxygen-free copper side of the samples. High-frequency ultrasonic tests were performed before and after machining the grooves using an ultrasonic scanner with a 30 MHz probe in the immersion pulse-echo setting. The testing results showed that the original screw pipe has a low capability for inspection with high-frequency ultrasonic testing. The visibility of a defect at the interface decreased because of the reflection of ultrasonic waves from the slope of the grooves and interference by the echoes from the interface and the grooves' surfaces. These issues were addressed by proposing geometrical modifications for the grooves and testing with high-frequency ultrasonic testing. The proposed modifications considerably increased the visibility of a defect at the interface and improved detectability by high-frequency ultrasonic tests. [ABSTRACT FROM AUTHOR]