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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...

Divertor power load reduction is one crucial issue for magnetically confined fusion reactors. Increased edge radiation can dissipate power before reaching divertor plates. Stable sustainment of such enhanced radiation, i.e. radiative divertor (RD) or detached divertor, is, however, still an open issue. In this paper, we present experimental evidence of edge radiation stabilization effect and potential of divertor power load control with resonant magnetic perturbation (RMP). Figure 1 shows time evolution of peak divertor power load, radiation intensity (P rad ) and line averaged density ( e n ) during density ramp up experiments with and without RMP. RMP has m/n=1/1 mode, which has resonance layer in the edge stochastic region, and creates remnant island. The perturbation strength is kept constant at ≈ 0 / B b r 0.1% throughout the discharge in the case with RMP. The plasma is heated by neutral beam injection (NBI) with ~ 8MW of deposited power in both cases. The divertor power load is estimated with Langmuir probe. The radiation is obtained with photo diode array viewing almost entire plasma at specific toroidal location. Without RMP (gray lines), the radiation intensity gradually increases with increasing density. The rapid increase of radiation intensity at t~3.8 sec with concomitant density rise indicates onset of thermal instability. The instability grows so rapidly that it is difficult to stabilize the density rise, leading to discharge termination. With RMP (black lines), on the other hand, transition to enhanced radiation state occurs at t~3 sec, and it leads to divertor power load reduction by a factor of 3 ~ 10. The RD operation is successfully sustained by gas puff feedback control up to the end of NBI. The results show stabilization effect of RMP on the radiating edge plasma. The enhanced radiation with RMP is due to increased volume of low T e (~10 eV) region caused by temperature flattening at the Opoint. 3D edge transport simulation result, which is consistent with the radiation profile measurement, show that the radiation increases further around X-point of the island (1) , where the code predicts n e > 10 20 m -3 and T e ~ a few eV.The well structured edge radiation with RMP such as the selective cooling around X-point is considered to provide stabilization effect by holding the intense radiation there and thus avoids it penetrating inward. Fig.2 shows dependence of controllability of RMP assisted RD on radial location of the island X-point,

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.

Abstract. The Large Helical Device (LHD) and Wendelstein 7-X (W7-X, under construction) are experiments specially designed to demonstrate long pulse (quasi steady-state) operation, which is an intrinsic property of Stellarators and Heliotrons. Significant progress had been made in establishing high performance plasmas. A crucial point is the increasing impurity confinement at high density observed at several machines (TJ-II, W7-AS, LHD) which can lead to impurity accumulation and early pulse termination by radiation collapse. In addition, theoretical predictions for non-axisymmetric configurations predict the absence of impurity screening by ion temperature gradients in standard ion-root plasmas. Nevertheless, scenarios were found where impurity accumulation was successfully avoided in LHD and W7-AS due to the onset of friction forces in the (high density and low temperature) scrape-off-layer, the generation of magnetic islands at the plasma boundary and to a certain degree also by ELMs, flushing out impurities and reducing the net-impurity influx into the core. In both the W7-AS High Density H-mode (HDH) regime and in the case of application of sufficient ECRH heating power a reduction of impurity core confinement was observed. The exploration of such purification mechanisms is a demanding task for successful steady-state operation. Impurity transport at the plasma edge/SOL was identified to play a major role for the global impurity behaviour in addition to the core confinement.

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The EMC3‐Eirene code is improved in many aspects. Ad hoc boundary conditions for intrinsic impurities at the SOL‐core interface are removed by implicitly coupling to a 1D core model. Non‐uniform cross‐field transport coefficients are allowed in the new code version. A particle splitting technique is implemented for improving the Monte Carlo statistic in low‐temperature ranges of most interest. Domain splitting, which was possible for the toroidal direction only, is now feasible for all three directions, facilitating mesh optimization for any specific divertor configuration. Stellarator‐specific constraints on mesh construction have been relaxed. Axisymmetric neutral‐facing components have been moved to cylindrical coordinates. All these features have improved the code performance and capability significantly. (© 2014 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

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.