After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3 × 1019 m−3, central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre.
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...
OVERVIEW OF THE LARGE HELICAL DEVICE PROJECT. The Large Helical Device (LHD) has successfully started running plasma confinement experiments after a long construction period of eight years. During the construction and machine commissioning phases, a variety of milestones were attained in fusion engineering which successfully led to the first operation, and the first plasma was ignited on 31 March 1998. Two experimental campaigns are planned in 1998. In the first campaign, the magnetic flux mapping clearly demonstrated a nested structure of magnetic surfaces. The first plasma experiments were conducted with second harmonic 84 and 82.6 GHz ECH at a heating power input of 0.35 MW. The magnetic field was set at 1.5 T in these campaigns so as to accumulate operational experience with the superconducting coils. In the second campaign, auxiliary heating with NBI at 3 MW has been carried out. Averaged electron densities of up to 6 × 10 19 m-3 , central temperatures ranging from 1.4 IAEA-F1-CN-69/OV1/4 2 to 1.5 keV and stored energies of up to 0.22 MJ have been attained despite the fact that the impurity level has not yet been minimized. The obtained scarling of energy confinement time has been found to be consistent with the ISS95 scaling law with some enhancement.
Edge cooling experiments with a tracer-encapsulated solid pellet in the Large Helical Device (LHD) show a significant rise of core electron temperature (the maximum rise is around 1 keV) as well as in many tokamaks. This experimental result indicates the possible presence of the nonlocality of electron heat transport in plasmas where turbulence as a cause of anomalous transport is dominated. The nonlocal electron temperature rise in the LHD takes place in almost the same parametric domain (e.g. in a low density) as in the tokamaks. Meanwhile, the experimental results of LHD show some new aspects of nonlocal electron temperature rise, for example the delay of the nonlocal rise of core electron temperature relative to the pellet penetration time increases with the increase in collisionality in the core plasma and the decrease in electron temperature gradient scale length in the outer region of the plasma.
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