Search

The effect of the pitch modulation of the helical coils on the core plasma performance of the LHD-type helical fusion reactor has been examined. The analysis of the MHD stability and neoclassical transport for the pitch modulation α = 0.0 and 0.1 has been conducted based on the finite-beta equilibrium calculated by the HINT code. It was found that the MHD stability is clearly improved without deteriorating the energy transport property by changing the pitch modulation α from 0.1 to 0.0. The reachable operation region expands to the higher density and the expected fusion gain can increase from ∼10 to ∼20. Because the change of the pitch modulation α from 0.1 to 0.0 requires only a slight change in the shape of the helical coils, the engineering design including the maintenance method that has been examined for the reactor with α = 0.1 can be applied without a major modification.

The improved H-mode scenario (or high ? hybrid operations) is one of the main candidates for highfusion performance tokamak operation that offers a potential steady-state scenario. In this case, the normalized pressure ?N must be maximized and pressure-driven instabilities will limit the plasma performance. These instabilities could have either resistive ((m = 2, n = 1) and (3,2) neoclassical tearing modes (NTMs)) or ideal character (n = 1 ideal kink mode). In ASDEX Upgrade (AUG), the first limit for maximum achievable ?N is set by the NTMs. The application of pre-emptive electron cyclotron current drive at the q = 2 and q = 1.5 resonant surfaces reduces this problem, so that higher values of ?N can be reached. AUG experiments have shown that, in spite of the fact that hybrids are mainly limited by NTMs, the proximity to the no-wall limit leads to amplification of the external fields that strongly influence the plasma profiles. For example, rotation braking is observed throughout the plasma and peaks in the core. In this situation, even small external fields are amplified and their effect becomes visible. To quantify these effects, the plasma response to the magnetic fields produced by B-coils is measured as ?N approaches the no-wall limit. These experiments and corresponding modeling allow the identification of the main limiting factors, which depend on the stabilizing influence of the conducting components facing the plasma surface, the existence of external actuators, and the kinetic interaction between the plasma and the marginally stable ideal modes. Analysis of the plasma reaction to external perturbations allowed us to identify optimal correction currents for compensating the intrinsic error field in the device. Such correction, together with the analysis of kinetic effects, will help to increase ?N further in future experiments.

Integrated physics analysis of plasma operation scenario of the compact helical reactor FFHR-c1 has been conducted. The DPE method, which predicts radial profiles in a reactor by direct extrapolation from the reference experimental data, has been extended to implement the equipartition effect. Close investigation of the plasma operation regime has been conducted and a candidate plasma operation point of FFHR-c1 has been identified within the parameter regime that has already been confirmed in LHD experiment in view of MHD equilibrium, MHD stability and neoclassical transport.

The Fusion Advanced Study Torus (FAST) has been proposed as a possible European satellite, in view of ITER and DEMO, in order to: (a) explore plasma wall interaction in reactor relevant conditions, (b) test tools and scenarios for safe and reliable tokamak operation up to the border of stability, and (c) address fusion plasmas with a significant population of fast particles. A new FAST scenario has been designed focusing on low-q operation, at plasma current IP = 10 MA, toroidal field BT = 8.5 T, with a q 95 ≈ 2.3 that would correspond to IP ≈ 20 MA in ITER. The flat-top of the discharge can last a couple of seconds (i.e. half the diffusive resistive time and twice the energy confinement time), and is limited by the heating of the toroidal field coils. A preliminary evaluation of the end-of-pulse temperatures and of the electromagnetic forces acting on the central solenoid pack and poloidal field coils has been performed. Moreover, a VDE plasma disruption has been simulated and the maximum total vertical force applied on the vacuum vessel has been estimated. © 2013 Euratom-ENEA Association sulla Fusione.

Effects of poloidal shear flow on the stability of interchange modes in a Large Helical Device (LHD) configuration are numerically studied. Three-dimensional (3D) numerical codes are utilized for the equilibrium and stability calculations. A static equilibrium is employed and a model poloidal flow as a flux function is incorporated in the initial perturbation. The results show that the initially applied flow can suppress the growth of the interchange mode if the flow is sufficiently large.

Property of pressure driven modes in Large Helical Device (LHD) plasmas with a resonant magnetic perturbation (RMP) is numerically studied. Particularly, we analyze three-dimensional (3D) RMP effects on the linear magnetohydrodynamic (MHD) stability of the modes. For this purpose, 3D numerical codes are utilized for both calculations of an equilibrium including an RMP generating an m = 1/n = 1 magnetic island and the stability of the perturbations resonant at the ι = 1 surface. Here, m and n are the poloidal and the toroidal mode numbers, respectively, and ι denotes rotational transform. Owing to the RMP, the pressure driven mode is localized around the X-point of the island. The type of the mode structure changes from the interchange type to the ballooning type. This property is attributed to the fact that the equilibrium pressure gradient is larger at the X-point than at the O-point.

The use of ECRH has been investigated as a promising technique to avoid or postpone disruptions in dedicated experiments in FTU and ASDEX Upgrade. Disruptions have been produced by injecting Mo through laser blow-off (FTU) or by puffing deuterium gas above the Greenwald limit (FTU and ASDEX Upgrade). The toroidal magnetic field is kept fixed and the ECRH launching mirrors have been steered before every discharge in order to change the deposition radius. The loop voltage signal is used as disruption precursor to trigger the ECRH power before the plasma current quench. In the FTU experiments (Ip = 0.35–0.5 MA, Bt = 5. 3T ,PECRH = 0.4–1.2 MW) it is found that the application of ECRH modifies the current quench starting time depending on the power deposition location. A scan in deposition location has shown that the direct heating of one of the magnetic islands produced by magnetohydrodynamic (MHD) resistive instabilities (either m/n = 3/2, 2/1 or 3/1) prevents its further growth and also produces the stabilization of the other coupled modes and the delay of the current quench or its full avoidance. Disruption avoidance and complete discharge recovery are obtained when the ECRH power is applied on rational surfaces. The modes involved in the disruption are found to be tearing modes stabilized by a strong local ECRH heating. The Rutherford equation has been used to reproduce the evolution of the MHD modes. In the ASDEX Upgrade experiments L-mode plasmas (Ip = 0.6 MA, Bt = 2. 5T ,PECRH = 0. 6M W∼ POHM) the injection of ECRH close to q = 2 significantly delays the 2/1 onset and prolongs the duration of the discharge: during this phase the density continues to increase. No delay in the onset of the 2/1 mode is observed when the injected power is reduced to 0.35 MW.

Recent progress in plasma performance and the understanding of the related physics in the Large Helical Device is overviewed. The volume-averaged beta value is increased with an increase in the neutral beam injection (NBI) heating power, and it reached 5.0% of the reactor-relevant value. In high-β plasmas, the plasma aspect ratio should be controlled so that the Shafranov shift would be reduced, mainly to suppress transport degradation and the deterioration of the NBI heating efficiency. The operational regime of a high-density plasma with an internal diffusion barrier (IDB) has been extended, and the IDB, which was originally found using the local island divertor, has been realized in the helical divertor configuration. The central density was recorded as high as 1 × 1021 m-3, and the central pressure reached 130 kPa. Based on these high-density plasmas with the IDB, a new ignition scenario has been proposed. This should be a scenario specific to the helical fusion reactor, in which the helical ripple transport would be mitigated. A low-energy positive-NBI system was newly installed for an increase in the direct ion heating power. As a result, the ion temperature (Ti) exceeded 5.2 keV at a density of 1.2 × 1019 m-3 in a hydrogen plasma. Transport analysis shows improvement of ion transport, and the Ti-increase tends to be accompanied by a large toroidal rotation velocity of the order of 50 km/s in the core region. The plasma properties in the extended operational regime are discussed from the perspective of a steady-state helical fusion reactor.

A kinetic-magnetohydrodynamic model with kinetic pressure closure is derived from a consistent guiding-centre framework. Higher-order (gyroviscous) corrections to the pressure tensor are derived in complex geometry from a reduced kinetic equation. The proposed model allows for flows of the order of the thermal ion velocity, taking into account important centrifugal effects due to the E x B flow, as well as the effects of diamagnetic flows associated with finite Larmor radius corrections to both ion fluid inertia and long mean-free path contributions. Wave-particle interactions, such as toroidal drift-resonance, are retained. Furthermore, the linearised model includes a quasi-neutrality equation, allowing the effects of a parallel electric field to be studied in fast rotating tokamak plasmas.