Island Divertor Experiment Research Articles

Application of three-dimensional MHD equilibrium calculation coupled with plasma response to island divertor experiments on J-TEXT

Three-dimensional (3D) equilibrium calculations, including the plasma rotation shielding effect to resonant magnetic perturbations (RMPs) produced by the island divertor (ID) coils, were carried out using the HINT and MARS-F codes on J-TEXT. Validation of 3D equilibrium calculations with experimental observations demonstrates that the shielding effect will prevent the penetration of the edge m/n = 3/1 mode component when the ID coil current is 4 kA, while change the size of magnetic islands once the current exceeds the penetration threshold. This indicates that equilibrium calculations including the plasma rotation shielding effect to RMPs can lead to better agreements with experimental observations compared to the vacuum approximation method. Additionally, the magnetic topology at the boundary undergoes changes, impacting the interaction between the plasma and the target plate. These results may be important in understanding RMP effects on edge transport and magnetohydrodynamic (MHD) instability control, as well as divertor heat and particle flux distribution control.

Local Island Divertor Experiment

To achieve an improvement of plasma confinement by an effective edge plasma control, the local island divertor (LID) was originally proposed in the National Institute for Fusion Science in the early 1980s. The LID is a kind of island divertor that utilizes the island separatrix as the channeling magnetic structure of the divertor, and it has the particular characteristic of localizing the particle recycling in very small areas. Thus, it is possible to construct a compact closed divertor configuration with efficient pumping capability, which results in the low-recycling condition in the edge region. In this paper the LID project is reviewed, from the physics design phase with numerical validation or estimation of the LID principle to a recent experimental result of the superdense core mode, which is a promising discharge for next-generation devices.

Carbon influx from the island divertor of the Wendelstein-7AS stellarator

Target plate fluxes (C,H) were measured in the island divertor experiment of W7-AS by CCD cameras equipped with interference filters. Strike point patterns depend strongly on the confinement regime. Carbon fluxes are given for attached divertor plasmas under HDH and normal confinement conditions. Our studies show that the basic picture of the radiation zones which has emerged from the investigation of tokamak divertors has to be completed for the island divertor by additional radiation zones in the X-point regions.

Overview of confinement and MHD stability in the Large Helical Device

The Large Helical Device is a heliotron device with L = 2 and M = 10 continuous helical coils with a major radius of 3.5–4.1 m, a minor radius of 0.6 m and a toroidal field of 0.5–3 T, which is a candidate among toroidal magnetic confinement systems for a steady state thermonuclear fusion reactor. There has been significant progress in extending the plasma operational regime in various plasma parameters by neutral beam injection with a power of 13 MW and electron cyclotron heating (ECH) with a power of 2 MW. The electron and ion temperatures have reached up to 10 keV in the collisionless regime, and the maximum electron density, the volume averaged beta value and stored energy are 2.4 × 1020 m−3, 4.1% and 1.3 MJ, respectively. In the last two years, intensive studies of the magnetohydrodynamics stability providing access to the high beta regime and of healing of the magnetic island in comparison with the neoclassical tearing mode in tokamaks have been conducted. Local island divertor experiments have also been performed to control the edge plasma aimed at confinement improvement. As for transport study, transient transport analysis was executed for a plasma with an internal transport barrier and a magnetic island. The high ion temperature plasma was obtained by adding impurities to the plasma to keep the power deposition to the ions reasonably high even at a very low density. By injecting 72 kW of ECH power, the plasma was sustained for 756 s without serious problems of impurities or recycling.

Formation and healing of n = 1 magnetic islands in LHD equilibrium

Magnetic islands with the toroidal mode number n = 1, e.g. m/n = 1/1 and 2/1 islands, in a Large Helical Device (LHD) equilibrium are studied using the three-dimensional MHD equilibrium code, HINT. In order to accomplish this purpose, the HINT code has been improved. The equilibrium analysis, in particular an analysis of the LHD equilibrium with an m/n = 1/1 island, is required for the local island divertor experiment, in order to understand the magnetic structures of field lines, i.e. flux surfaces, islands and ergodic field lines. We find that the m/n = 2/1 island can be healed for a finite equilibrium beta, while the m/n = 1/1 island is not healed and is surrounded with ergodic field lines for finite-β. From the latter result, we can conjecture that the island divertor concept is effective even for finite equilibrium beta-values, but the performance of the island divertor is deteriorated for finite-β because of the existence of the ergodic zone between the closed surfaces (i.e. the core region) and the m/n = 1/1 island. We also find that the width of the m/n = 1/1 island depends on the equilibrium beta value and that the island located at the inside of the torus has the advantage of retaining its width.

Local island divertor experiments on LHD

A local island divertor (LID) experiment has begun on LHD, with the aims of controlling edge recycling and improving the plasma confinement. The fundamental divertor functions of the LID have been demonstrated in the recent experiments. From the particle flux profile measurements on the LID head it was found that the particles diffusing out from the core region are well guided along the island separatrix to the LID head. Owing to the closed configuration around the LID head, evidence of the high efficient pumping was observed, together with a strong capacity to screen impurities. The first results of edge modeling using the EMC3-EIRENE code are also presented.

Physics of the geometry-related detachment stability in W7-AS

This paper presents a detailed analysis of the transport behaviour of the detached plasmas in W7-AS based on an extended numerical study using the EMC3-EIRENE code, aimed at understanding the underlying physics responsible for the geometry-dependent detachment stability observed in W7-AS island divertor experiments. Here, a stable detachment can only be established when the control coils are switched on to generate sufficiently large islands with relatively short connection lengths. Special attention will be paid to a discussion of the carbon radiation, location and dynamics of the radiation layer, the neutral screening efficiency specific to the island divertor geometry and its impact on the detachment stability. Based on the three-dimensional simulation results, a linear stability model is presented in order to obtain some insight into the mechanisms driving the instability. The radiation behaviour and the location and evolution of the radiation zone in the island divertor will be discussed with respect to those of tokamak-MARFEs.

Initial Results of Local Island Divertor Experiments in the Large Helical Device

A local island divertor (LID) experiment has begun in the Large Helical Device (LHD) to demonstrate improved plasma confinement, and fundamental LID functions were demonstrated in the sixth experimental campaign in 2002–2003. It was clearly shown that when an m/n = 1/1 island is generated by adding a resonant perturbation field to the LHD magnetic configuration, the particle flow is guided along the island separatrix to the backside of the island, where carbon plates are located on a divertor head. The particles recycled there are pumped out efficiently so that the line-averaged core plasma density is reduced by a factor of ~2 at the same gas puff rate, compared with non-LID discharges. Obvious improvement of the global plasma confinement was, however, not observed yet, because the discharge could not be optimized, due to a large amount of outgas from the divertor head to the core plasma. The size of the divertor head was found to be larger than the optimum one; hence, the core plasma impacted slightly on the core plasma-facing portion of the divertor head with which the core plasma was not expected to collide.

03/02650 Progress of local island divertor experiment: Morisaki, T. et al. Fusion Engineering and Design, 2003, 65, (3), 475–481

03/02650 Progress of local island divertor experiment: Morisaki, T. et al. Fusion Engineering and Design, 2003, 65, (3), 475–481

First island divertor experiments on the W7-AS stellarator

W7-AS has recently been equipped with ten open divertor modules in order to experimentally evaluate the island divertor concept. First results are reported in this paper. The new divertors enable access to a new NBI-heated, very high density (up to n̄e = 3.5 × 1020 m-3) operating regime with promising confinement properties. The energy confinement time increases steeply with density and then saturates. In contrast, the particle and impurity confinement times decrease with increasing density. This allows full density control and quasi-steady-state operation also under conditions of partial detachment from the divertor targets. Radiated power fractions are low to moderate in attached regimes and reach up to about 90% in detachment scenarios. The radiation always stays peaked at the edge. The extremely high densities necessitated the development of non-standard heating techniques for central heating. For the first time efficient heating of an NBI target plasma by electron Bernstein waves (140 GHz, second harmonic) is achieved. In addition, this heating scenario enables fine tuning of the upstream boundary conditions for divertor operation.

Local island divertor experiments on CHS

Local island divertor experiments on CHS

Local island divertor experiments on CHS

A local island divertor (LID) was installed on the Compact Helical System (CHS) to demonstrate the principle of LID and to study its effects on the edge plasma. The particle flow was observed to be guided to the back side of the externally created magnetic island along field lines, and pumped out by a cryogenic pump with high pumping efficiency. As a result, a factor of about 2 reduction in the average core density was seen compared with non-LID discharges at the same gas puff rate. In addition to the demonstration of these fundamental divertor functions, we observed a modest improvement of energy confinement, which could be due to the edge plasma control by the local island divertor.