In reduced recycling discharges in the Large Helical Device, a super dense core plasma develops when a series of pellets are injected. A core region with density as high as 4:5 10 20 m ÿ3 and temperature of 0.85 keV is maintained by an internal diffusion barrier with very high-density gradient. These results may extrapolate to a scenario for fusion ignition at very high density and relatively low temperature in helical devices. DOI: 10.1103/PhysRevLett.97.055002 PACS numbers: 52.55.HcImprovement of plasma particle and energy confinement is a major challenge for toroidal magnetic fusion research, and will be important in igniting burning plasmas in ITER [1]. Various confinement improvement modes have been discovered including edge transport barriers (ETBs, or H mode) [2] and internal transport barriers (ITBs) [3][4][5]. In this Letter, we describe improved confinement in super dense core (SDC) plasmas, in diverted discharges in the Large Helical Device (LHD), a heliotron configuration in which the rotational transform is provided by external magnetic coils. This operational regime may extrapolate to a high-density, relatively low temperature ignition scenario for these devices.LHD has an external helical field with poloidal winding number l 2 and M 10 toroidal field periods. The major radius of the magnetic axis, R ax 3:5-3:9 m, average plasma minor radius a 0:6 m, and toroidal magnetic field B 3:0 T [6]. Depending on the relative currents in the helical and auxiliary poloidal coils, the rotational transform on axis, 0 =2 0:3-0:6 and the edge transform, a =2 1-1:5. One of the major goals of the LHD program is the demonstration of a reactor-relevant, diverted helical plasma. Two different divertor systems are available in LHD: the Helical Divertor (HD) [7] and the Local Island Divertor (LID) [8][9][10]. The HD is an intrinsic helical double-null divertor with an open divertor geometry, essentially like a helically twisting double-null tokamak poloidal divertor. The LID uses an m 1, n 1 resonant magnetic island (poloidal and toroidal mode numbers m and n, respectively) to guide particle and heat fluxes to divertor plates.A SDC plasma develops spontaneously in LHD as a highly peaked density profile is created by injection of multiple pellets from the outside midplane as illustrated in Fig. 1(a). The density and temperature profiles are depicted for the standard (R ax 3:75 m, B 2:64 T, P 10 MW) discharge diverted by the LID in Fig. 1(b). These profiles are measured using a Thomson scattering diagnostic along R horiz , the major radius in the poloidal plane where the plasma is horizontally elongated [ Fig. 1(a)]. A core region with electron density 4:5 10 20 m ÿ3 and temperatures 0:85 keV is maintained by an internal diffusion barrier (IDB) located at normalized minor radius 0:5. The radial width of the IDB is 0:10 m ( 0:2). The density gradient at the IDB is extremely high (rn 2:5 10 21 m ÿ4 ). Inside the SDC region, the density and temperature gradients are nearly zero. The density gradient outside the IDB is we...

It is found that the remnant island structure created by n/m=1/1 resonant magnetic perturbation field in the stochastic magnetic boundary of the Large Helical Device (LHD) [A. Komori et al., Nucl. Fusion 49, 104015 (2009).] has a stabilizing effect on formation of radiating plasma, realizing stably sustained divertor detachment operation with the core plasma being unaffected. The data from the several diagnostics, (profiles of electron temperature & density, radiation and temporal evolution of divertor particle flux) indicate selective cooling around X-point of the island and thus peaked radiation there, which is stabilized outside of the last closed flux surface throughout the detachment phase. The VUV spectroscopy measurements of high Z impurity (iron) emission shows significant decrease during the detachment, indicating core plasma decontamination. The results from the 3D edge transport code EMC3 (Edge Monte-Carlo 3D) [Y. Feng et al., Contributions to Plasma Physics, 44, 57 (2004).]-EIRENE [D. Reiter et al., Fusion Sci. Technol., 47 172 (2005).] show similar tendency in the radiation pattern. The island size and its radial location are varied to investigate the magnetic topology effects on the detachment control. The divertor particle flux and neutral pressure exhibit intermittent oscillation as well as modification of recycling pattern during the detachment, which are found to reflect the island structure.

In the first four years of the LHD experiment, several encouraging results have emerged, the most significant of which is that MHD stability and good transport are compatible in the inward shifted axis configuration. The observed energy confinement at this optimal configuration is consistent with ISS95 scaling with an enhancement factor of 1.5. The confinement enhancement over the smaller heliotron devices is attributed to the high edge temperature. We find that the plasma with an average beta of 3% is stable in this configuration, even though the theoretical stability conditions of Mercier modes and pressure driven low-n modes are violated. In the low density discharges heated by NBI and ECR, internal transport barrier (ITB) and an associated high central temperature (>10 keV) are seen. The radial electric field measured in these discharges is positive (electron root) and expected to play a key role in the formation of the ITB. The positive electric field is also found to suppress the ion thermal diffusivity as predicted by neoclassical transport theory. The width of the externally imposed island is found to decrease when the plasma is collisionless with finite beta and increase when the plasma is collisional. The ICRF heating in LHD is successful and a high energy tail (up to 500 keV) has been detected for minority ion heating, demonstrating good confinement of the high energy particles. The magnetic field line structure unique to the heliotron edge configuration is confirmed by measuring the plasma density and temperature profiles on the divertor plate. A long pulse (2 min) discharge with an ICRF power of 0.4 MW has been demonstrated and the energy confinement characteristics are almost the same as those in short pulse discharges.

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 HTS prototype conductor sample achieved 100 kA for 1 hour  Mechanical lap-joint confirms low joint resistance (2 nW)  "Joint-winding" of helical coils technically feasible

As the finalization of the hydrogen experiment towards the deuterium phase, the exploration of the best performance of the hydrogen plasma was intensively performed in the Large Helical Device (LHD). High ion and electron temperatures, Ti, Te, of more than 6 keV were simultaneously achieved by superimposing the high power electron cyclotron resonance heating (ECH) on the neutral beam injection (NBI) heated plasma. Although flattening of the ion temperature profile in the core region was observed during the discharges, one could avoid the degradation by increasing the electron density. Another key parameter to present plasma performance is an averaged beta value . The high regime around 4 % was extended to an order of magnitude lower than the earlier collisional regime. Impurity behaviour in hydrogen discharges with NBI heating was also classified with the wide range of edge plasma parameters. Existence of no impurity accumulation regime where the high performance plasma is maintained with high power heating > 10 MW was identified. Wide parameter scan experiments suggest that the toroidal rotation and the turbulence are the candidates for expelling impurities from the core region.

An inter-machine dataset covering devices of different size and a variety of magnetic configurations is comprehensively analysed to assess the ranges of validity of neoclassical (NC) transport predictions in medium-to high density, high temperature discharges. A recently concluded benchmarking of calculations of transport coefficients from local NC theory [1] allows now a quantitative experimental energy transport study. While in earlier inter-machine studies of NC transport in 3D devices the electron energy transport at low densities has been investigated [2], this study focuses on the energy transport at medium to higher densities as anticipated when approaching reactor conditions. The validation approach as done here is to compare two fluxes: first, the 'NC flux' is determined with the NC transport coefficients and the gradients of the experimental density and temperature profiles. Second, the sources from deposition calculations considering heating and particle sources (the latter where available) yield the 'experimental flux'. Both fluxes are compared and the NC radial electric field E

The stochastic scrape-off layer (SOL) of the helical divertor configuration in LHD exhibits a rather complex field topology where remnant magnetic islands, thin edge surfaces and stochastic field lines coexist. Using the three-dimensional (3D) edge transport code package, EMC3-EIRENE, the paper presents a numerical study of the stochastic layer transport, aimed, first of all,