The heuristic drift (HD) model for the tokamak power scrape-off layer width provides remarkable agreement in both absolute magnitude and scalings with the measured width of the exponential component ...of the heat flux at divertors targets, in low gas-puff H-Mode tokamaks. This motivates further exploration of its theoretical aspects and practical implications. The HD model requires a small non-ambipolar electron particle diffusivity ∼10−2m2/s. It also implies large parallel heat flux in ITER and suggests that more radical approaches will be needed to handle the ∼20 GW/m2 parallel heat flux expected in Demo. Remarkably, the HD model is also in good agreement with recent near-SOL heat flux profiles measured in a number of limiter L-Mode experiments, implying ubiquity of the underlying mechanism. Finally, the HD model suggests that the H-Mode and more generally Greenwald density limit may be caused by MHD instability in the SOL, rather than originating in the core plasma or pedestal. If the SOL width in stellarators is set by magnetic topology rather than by drifts, this would be consistent with the absence of the Greenwald density limit in stellarators.
Experimental measurements of the SOL power decay length (λ(q)) estimated from analysis of fully attached divertor heat load profiles from two tokamaks, JET and ASDEX Upgrade, are presented. Data was ...measured by means of infrared thermography. An empirical scaling reveals parametric dependency λ(q) in mm = 0.73B(T)(-0.78)q(cyl)(1.2)P(SOL)(0.1)R(geo)(0), where B(T)(T) describes the toroidal magnetic field, q(cyl) the cylindrical safety factor, P(SOL)(MW) the power crossing the separatrix and R(geo)(m) the major radius of the device. A comparison of these measurements to a heuristic particle drift-based model shows satisfactory agreement in both absolute magnitude and scaling. Extrapolation to ITER gives λ(q) ≃ 1 mm.
Abstract
The lithium vapor box divertor is a proposed divertor design that will minimize contamination of the upstream plasma in a fusion device, while also ensuring protection of the target. In this ...design lithium is evaporated near the target by high temperature lithium surfaces, dissipating the plasma heat flux. The lithium vapor box has been predicted via the fluid-kinetic code scrape off layer plasma simulator (SOLPS-ITER) to achieve low (
n
L
i
/
n
e
∼
0.05) upstream concentrations of lithium and low target heat fluxes. Here we compare two choices of deuterium gas puff location using SOLPS-ITER, the private flux region (PFR) and the common flux region (CFR), and find significant differences in the contamination level required to reach an acceptable target heat flux (defined here as q
tar
max
⩽
10 MW m
−2
). Deuterium gas puffing from the PFR is seen to better reduce upstream lithium contamination. The difference in puffing location is seen to cause changes in the upstream flow of lithium ions. The PFR puff, having better access to the separatrix, can better reduce the upstream-directed flow of lithium near the separatrix which is the primary source of contamination due to a large thermal force in this region. Puffing from the CFR, partially due to inefficacy at reducing separatrix lithium flow, has higher lithium concentration within the plasma. Solutions that reduce the heat flux to below 10 MW m
−2
have a range of lithium concentrations between
n
L
i
/
n
e
∼
0.01–0.12 depending on puff intensity, location, evaporator temperature and recycling at the various plasma facing components. The efficacy of the puffs is tested for sensitivity to deuterium recycling coefficient at the target, evaporators, and main chamber walls. A CFR located puff is found to be more dependent on the recycling coefficients used than a PFR located puff. regardless of the set of recycling coefficients chosen, PFR puffing achieves lower lithium contamination than CFR puffing for a given heat flux.
A present topic of high interest in magnetic fusion is the “gap” between near-term and long-term concepts for high heat flux components (HHFC), and in particular for divertors. This paper focuses on ...this issue with the aim of characterizing the international status of current HHFC design concepts for ITER and describing the different technologies needed in the designs being developed for fusion power plants. Critical material and physics aspects are highlighted while evaluating the current readiness level of long-term concepts, identifying the design and R&D gaps, and discussing ways to bridge them.
The ITER first wall is designed for start-up and ramp-down in limiter configuration. The wall panels are toroidally shaped in order to spread the incident parallel power flux q|| uniformly, assuming ...a single decay length λq whose value is not known from first principles. In order to study the scaling of q|| with plasma parameters, infra-red viewing of specially-designed limiters has been used on the COMPASS tokamak in ∼100 discharges with scans in Ip, ne and for all combinations of magnetic field and Ip directions. The IR measurement clearly shows that in addition to the main SOL heat flux profile with λq>40mm, a steep gradient (λqnear=4±2mm) dominates q|| near separatrix. This appears independently of limiter shaping, insertion with respect to neighbors and incident field-line angles. Good agreement is found between the measured λqnear and the prediction of a heuristic drift-based model.
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