Abstract Although actively cooled W/Cu monoblock targets are designed for high steady-state heat exhaust, local overheating and damage have been observed on these targets in EAST. These observations indicate that, beyond the nominal heat-removal capability, the practical thermal margin can be affected by engineering alignment, target geometry and plasma operating conditions, which modify the local surface-normal heat flux through magnetic incidence and magnetic shadowing. In this work, the toroidal heat-load distribution on EAST divertor targets is analysed by combining infrared thermography, finite-element heat-flux inversion and three-dimensional field-line tracing using PFCFlux. The analysis focuses on the effects of toroidal-field ripple, ripple-coupled cassette misalignment and divertor target geometry. The 16-fold toroidal-field ripple produces a 2π/16 modulation of the peak heat flux on the present upper outer target, with a peak fluctuation amplitude of approximately 17.75%. Under a peak heat-flux constraint of Qsafe = 10 MW m-2, the allowable average heat flux decreases from 4.28 MW m-2 without ripple to 3.94 MW m-2 with ripple. When ripple and relative cassette misalignment are considered together, it is further reduced to 3.11, 2.79 and 2.27 MW m-2 for Delta = 0.6, 1.0 and 3.0 mm, respectively. Comparison with the lower divertor shows that improved toroidal continuity reduces ripple-only peak amplification. Using the lower outer horizontal target as a reference, the new upper horizontal target reaches an allowable average heat flux of 4.70 MW m-2 under ripple and Δ = 1 mm misalignment. These results show that magnetic ripple, assembly tolerance and toroidal geometric continuity should be considered together in divertor thermal design.
Abstract Snowflake divertor (SFD) experiments show enhanced radial transport near the X-points, more than may be explained by standard two-dimensional edge transport models without drifts. Here we investigate transport mechanisms in SFDs through two complementary approaches: i) modelling of recent SFD experiments performed at MAST-U with the edge transport code UEDGE with ExB and ∇B drifts enabled, and ii) reduced modelling of transport and turbulence in the vicinity of the X-points with a model developed here and implemented in BOUT++. Through these two approaches we assess the two leading candidates for this transport: equilibrium ExB drifts driven by strong poloidal gradients near the primary X-point, and the 'churning mode' (CM), a toroidally-symmetric plasma vortex close to the X-points. UEDGE modelling shows that equilibrium ExB drifts do play a role in increasing particle and energy transport across the primary separatrix, increasing the proportion of heat flux delivered to secondary strike points (SPs) and reducing peak heat fluxes at the primary SPs. This mechanism is less effective in the upper divertor due the reversed drift directions in combination with the device geometry. Including drifts in the model is also shown to improve agreement with the experimental radiated power profiles. To self-consistently compare the impact of drifts and the CM we develop a reduced-MHD model which includes both effects, and which is implemented in BOUT++. We find that equilibrium ExB drifts are dominant when the ratio of plasma pressure in the null region to poloidal magnetic pressure at the midplane is low, β pm ≤ 0.5-1.0%, which were the conditions obtained in the MAST-U experiments. The CM was found to be dominant for β pm ≥ 1.0%, leading to increased transport across the null region and activation of secondary divertor legs.
Abstract Tungsten (W) is a preferred plasma-facing material for fusion reactor divertors. During service, high heat loads induce considerable thermal strain, while large amounts of free hydrogen (H) and helium (He) atoms accumulate in the W matrix. These factors promote the nucleation and coalescence of near-surface He bubbles, affecting the service safety of plasma-facing components and the transport behavior of hydrogen isotopes. However, the combined effects of strain and free H/He atoms on He bubble coalescence in W remain unclear. This work employs molecular dynamics (MD) simulations to systematically investigate the mechanisms by which uniaxial strain and free H/He atoms affect He bubble coalescence in W. The analysis is further combined with the thermal strain characteristics of divertor service conditions and the equilibrium pressure behavior of He bubbles. The results show that both tensile and compressive strains promote He bubble coalescence, while compressive strain has a stronger effect. Tensile strain facilitates bubble expansion by reducing the local atomic density, whereas compressive strain increases the internal He bubble pressure and accelerates He bubble connection. In H-containing systems, H concentration shows a dual effect: low H concentration promotes coalescence, whereas high H concentration suppresses it due to competition for defect sites around He bubbles. In H–He coexisting systems, free He atoms form He or H–He clusters, which redistribute H atoms and reduce H adsorption by pre-existing He bubbles. A modified equilibrium-pressure relation was further introduced to describe strain-assisted He bubble coalescence. These results provide atomic-scale insight into He bubble evolution in divertor W under complex service conditions.
Abstract While a tokamak has a plasma current disruption, a helical system is much robust because the current-less operation is, in principle, possible. However, a bootstrap (BS) current due to the pressure gradient could be concerns because it could change the rotational transform. We conducted experiments on the BS current in the Large Helical Device (LHD) in the 1/ν and plateau regimes using three magnetic axis configurations Rax = 3.75 m, 3.6 m and 3.53 m with completely balanced neutral beam injection (NBI). As a reference, we use the standard configuration of Rax = 3.75 m. We then examine the inward-shifted configuration of Rax = 3.6 m, and the strongly inward-shifted configuration of Rax = 3.53 m, which is referred to the neoclassically optimized configuration [1, 2]. It was demonstrated that a clear reduction of the BS current is observed for Rax = 3.53 m than that for Rax = 3.75 m, together with a confinement improvement. Based on these experimental results, the BS current coefficient has been derived, which helps to predict the BS current in a helical fusion reactors.
Abstract Tritium retention in plasma-facing components(PFCs) must be strictly controlled to ensure the safe operation of fusion devices. In ITER, compliance with the tritium inventory limit and assessment of fuel-removal strategies require in situ monitoring of fuel retention in PFCs. Laser-Induced Breakdown Spectroscopy(LIBS) is a diagnostic technique that can be implemented via remote handling to investigate fuel retention during maintenance shutdowns at ITER. In LIBS diagnostics, the selection of laser parameters plays a critical role in determining the ablation characteristics and thermal perturbation of the target. Used nanosecond(ns) laser pulses deliver energy on timescales that allow substantial thermal diffusion into the target, which could result in surface heating and melting. Under multi-pulse LIBS operation, these cumulative thermal effects can alter deuterium quantification, as deuterium may diffuse or outgas due to heating induced by earlier pulses. Such laser-induced perturbations can compromise accuracy of quantitative deuterium measurements if they are not properly accounted for in modelling. In this study, subsequent in situ ns-LIBS and in situ Nuclear Reaction Analysis(NRA) measurements were performed in Magnum-PSI on compact and porous tungsten coatings using a defocused, millimetre-scale laser spot(2.5×2.5mm). Deuterium retention and depth profiles were determined in and outside the LIBS crater. Under the investigated laser conditions, NRA measurements indicated a 66–79% reduction in retained deuterium within the sampled region of the LIBS crater, whereas pure material removal by ablation would account for only 21–42%, demonstrating additional laser-induced outgassing. Compact coatings further exhibited melting and intermixing with the underlying substrate. These findings highlight the importance of accounting for defocused ns-LIBS-induced perturbations in retention measurements. Further experimental and modelling studies are needed to assess the dependence of these effects on laser spot geometry and to determine whether picosecond(ps) LIBS, together with consideration of cumulative multi-pulse effects, could improve the reliability of fuel-retention diagnostics in future fusion devices.
