External kink modes in a Large Helical Device (LHD) equilibrium with self-consistent bootstrap current

Averaged flux surfaces with respect to the toroidal angle in l =2 heliotron/torsatrons are deformed in D-shape or inverse D-shape by the hexapole field. D-shape deformation of vacuum flux surfaces enhances magnetic well or reduces magnetic hill in Heliotron-E and Large Helical Device (LHD) configuration. However, complete elimination of the magnetic hill region is not obtained by the D-shape deformation without substantial reduction of plasma radius. The inverse D-shape deformation degrades MHD stability against ideal interchange modes in the central region of both devices for fairly peaked pressure profiles. In the LHD configuration, a broad pressure profile proportional to (1-ψ 2 ) 2 degrades the MHD stability and the hexapole effect is almost irrelevant, where ψ is a poloidal flux function; however, second stability appears against both the local and non-local interchange modes.

Local helicity injection (LHI) is a non-inductive startup and current drive method via Taylor relaxation for the spherical torus. In achieving Taylor relaxation, it has been suggested that kink instability in 3D helical flux ropes plays an important role. However, the role and occurrence of kink instability during LHI have yet to be validated. Experimentally, determining the kink mode in a flux rope relies on measuring internal information using a probe. However, for LHI, the 3D geometry complicates this measurement process. Here, we propose a new approach for determining the kink modes of 3D helical flux ropes without any internal probe measurements. It is confirmed by this approach that flux ropes exhibit two different kink modes. With increasing plasma current in the flux ropes, a transition from the coherent internal kink mode to the external kink mode is observed. Kink mode properties such as rotating frequency calculated from the kink theory agree well with the magnetic signature driven by the kink mode. During the LHI experiment in the versatile experiment spherical torus, three distinguishable phases are confirmed by the approach, consistent with NIMROD simulation. Before driving the toroidal plasma current, the external kink mode is observed for 3D helical flux ropes. As the toroidal plasma current increases, the external kink mode disappears while generating broadband internal modes instead of coherent internal kink of flux ropes. Decoupling between the toroidal plasma and flux rope results in both decay of toroidal plasma current and re-appearance of the external kink mode in the flux ropes.

Recent progress of superconducting magnet system design for Large Helical Device (LHD) of National Institute for Fusion Science (NIFS) is described. LHD is one of large scale nuclear fusion devices which will be completed in the 20th century, and fully superconducting coil system to create magnetic field. LHD is composed of a plasma vacuum vessel, superconducting helical and poloidal coils, cryogenic supporting structures, and an outer vacuum vessel, and surrounded by utilities of coil power supply, helium refrigerator and nitrogen supply, and vacuum pumping system. Cryogenic mass of the system is estimated as 860tons including coil and supporting structure. Cooling method of the superconducting coil is decided as bath cooled of liquid helium for the helical coil, and force cooled of supercritical helium for the poloidal coil. Major and minor radii of the helical coil are 3.9m and 0.975m, respectively. The conductor material of Nb-Ti alloy is used for both helical and poloidal coils from the point of strain effect. Conductor maximum field will be 6.9T for the plasma center field of 3T at the first phase with saturated boiling helium and 9.2T for 4T at the second phase with superfluid helium, Electromagnetic force and electrical insulation of the coil are the most important items to design the coil. Research and development work is intensively continued now, including conductor stability, helical and poloidal winding, elasticity of the coil, and cryogenic handling.

The three-dimensional (3D) plasma response to the magnetic field structure is studied for high-β plasmas in the Large Helical Device (LHD). For the vacuum field in the LHD configuration, the boundary of the plasma is defined by the last closed flux surface. However, for high β, the magnetic field structure is changed by the 3D plasma response and the plasma boundary is shifted in the radial direction. The radial electric field, Er, to identify the plasma boundary, is measured in the peripheral region. The radial electric field shear from negative to positive appears in the region and this suggests the boundary between open and closed field lines. The position of the Er shear is always outside the vacuum boundary. A 3D magnetohydrodynamic modelling predicts the expansion of the stochastic region by the 3D plasma response. However, since the connection length of the stochastic field lines is larger than the electron mean free path, a confinement is expected in the region. In such a case, the edge of the stochastic region can be defined as the effective plasma boundary. The position of the appearance of the strong Er shear is almost comparable to the edge of the stochastic region of the modelling. That is, the 3D plasma response is identified in LHD experiments.

This paper reports on the results of an experimental investigation into electron cyclotron resonance heating (ECRH) using the fundamental O-mode in the large helical device (LHD). The aim of the experiment is to understand the mechanism of the selective excitation of the O-mode EC wave and, thereby, optimize the performance of the ECRH. The polarization angle and the ellipticity of the electric field of the incident high power millimetre waves were varied in the experiment using a pair of grating polarizers. The responses of the plasma stored energy and electron temperature were measured to evaluate the effects. The optimum values of the polarization angle and the ellipticity for the ECRH were identified experimentally. A simple model of power branching between the O- and X-modes is introduced in order to interpret the observed experimental results. Specific to the ECRH in the LHD configuration, the effect of magnetic shear in the plasma peripheral region is considered.

Progress in the understanding of magnetohydrodynamic (MHD) stability is summarized on JT-60U tokamak discharges with improved confinement such as the (hot-ion) H-mode, high-βp mode, high-βp H-mode, and reversed shear discharges. Transport barriers, which are essential for the improved confinement, play key roles in the local and global MHD stability owing to the local large pressure gradient and the related bootstrap current. Disruptive β limits of these discharges are consistent with theoretical ideal kink-ballooning stability limits with low toroidal mode numbers n. Achievable β limit is improved by broadening of the pressure profile with high plasma internal inductance, plasma shaping, and wall stabilization. Edge localized modes (ELMs) and barrier localized modes (BLMs), which are associated with edge and internal transport barriers, respectively, are analyzed carefully. Resistive interchange modes with n ≤ 3 are excited in the negative shear region in reversed shear discharges with the internal transport barrier and lead to major collapse occasionally through nonlinear coupling with a tearing mode in the positive shear region. MHD characteristics of low m/n (m: poloidal mode number) tearing modes, which are attributed to the neoclassical tearing mode, are investigated. Stabilization of tearing modes and control of sawtooth activity are demonstrated using the fundamental O-mode electron cyclotron wave injection. Resistive wall modes associated with current-driven and pressure-driven low n external kink modes are identified.

Review of SC magnet technologies developed in LHD project

Stability and quench experiments of a Nb-Ti cable-in-conduit superconductor with bare strands have been carried out to determine the performance of the poloidal coil for the Large Helical Device (LHD). The conductor was formed into a double-pancake coil named IV-S and was mounted on TOKI-PF-a previously tested R&D coil. In excitation tests at 7.4-8.3 K, the IV-S coil reached the critical currents without premature quenches. Stability tests at 7.5 K indicated that the limiting current exceeds 20 kA, which is the nominal operating current of the LHD poloidal coils. These results demonstrated that the stability of the chosen conductor with bare strands is high enough for the LHD.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

The Large Helical Device (LHD) is a plasma physics experimental device with a magnetic stored energy of 960 MJ, consisting of two superconducting (SC) helical coils and six SC poloidal coils. The trial operation and the first plasma discharge of the Phase I project for LHD were completed on 31 March 1998 as initially planned. The second experimental campaign was conducted with additional heating using two NBI devices. The third campaign started in June 1999 and was completed in January 2000. Many plasma heating tests with a plasma field of up to 2.90 T were carried out. Major test results on the SC magnet system for LHD are as follows: (1) The LHD cryogenic system was successfully operated for 13 400 h and proved its high reliability. (2) A central field of 2.91 T at a radius of 3.60 m was achieved at an inner helical coil current of 11.08 kA, a middle helical coil current of 11.83 kA and an outer helical coil current of 12.02 kA. (3) All six poloidal coils were excited stably. (4) Nine flexible SC bus lines with a total length of 497 m were operated stably and safely.

An explanation is provided for the disruptive instability in diverted tokamaks when the safety factor$q$at the 95 % poloidal flux surface,$q_{95}$, is driven below 2.0. The instability is a resistive kink counterpart to the current-driven ideal mode that traditionally explained the corresponding disruption in limited cross-sections (Shafranov,Sov. Phys. Tech. Phys., vol. 15, 1970, p. 175) when$q_{edge}$, the safety factor at the outermost closed flux surface, lies just below a rational value$m/n$. Experimentally, external kink modes are observed in limiter configurations as the current in a tokamak is ramped up and$q_{edge}$decreases through successive rational surfaces. For$q_{edge}&lt;2$, the instability is always encountered and is highly disruptive. However, diverted plasmas, in which$q_{edge}$is formally infinite in the magnetohydrodynamic (MHD) model, have presented a longstanding difficulty since the theory would predict stability, yet, the disruptive limit occurs in practice when$q_{95}$, reaches 2. It is shown from numerical calculations that a resistive kink mode is linearly destabilized by the rapidly increasing resistivity at the plasma edge when$q_{95}&lt;2$, but$q_{edge}\gg 2$. The resistive kink behaves much like the ideal kink with predominantly kink or interchange parity and no real sign of a tearing component. However, the growth rates scale with a fractional power of the resistivity near the$q=2$surface. The results have a direct bearing on the conventional edge cutoff procedures used in most ideal MHD codes, as well as implications for ITER and for future reactor options.

Performance of the superconducting helical coils of the Large Helical Device (LHD) during the past 12 cooling cycles is reviewed. The pair of helical coils are pool cooled by liquid helium and wound with aluminum-stabilized NbTi/Cu composite-type superconductors. Intensive efforts have been made to reliably carry out excitations, as more than 20 temporary normal transitions were observed. It was found that the minimum propagation current was about 10% lower than the nominal operation current. To improve the cryogenic stability, subcooled liquid helium has been supplied since 2006 using cold compressors, and the inlet temperature is lowered to be 3.2 K. The toroidal magnetic field has been raised by 5% and the plasma parameters are being enhanced. Pulse-height analysis is successfully applied on the balance voltage and acoustic emission signals to investigate the mechanical properties of the windings and their changes in years of operation. Short-duration normal transitions are automatically detected using a sophisticated monitoring system and careful operations are continued.

Helical coils of the Large Helical Device (LHD) are large scale superconducting magnets for fusion plasma experiments. The cooling system was upgraded in 2006 to improve the cryogenic stability of the coils by subcooling supplied helium. After the upgrade, the subcooled helium of designed mass flow rate of 50 g/s at designed temperature of 3.2 K has been supplied to the coils stably during steady state subcooling operations. The supplied helium is subcooled at a heat exchanger in a saturated helium bath. A series of two centrifugal cold compressors are utilized in the subcooling system to reduce the bath pressure and temperature. It is necessary to operate the cold compressors within a range of appropriate mass flow rate in order to escape from surge and choke. When the rotational speed of the cold compressors increased/decreased, the mass flow rate of the cold compressors varied. Thus, it is important to comprehend the transient performance of the cold compressors in those cases and to control the mass flow rate for the purpose of the safe operations. In the present study, the characteristics of the system were investigated and automatic operating methods with a heater in the bath and a bypass valve were developed during the change of the rotational speed of the cold compressors. Consequently, the reliable subcooling operations of the system have been achieved over 6,000 hours for three years.

Vertical profiles of edge impurity emissions have been measured in upper half region of elliptical plasmas at horizontally elongated plasma cross section in large helical device (LHD). The vertical profiles near upper O-point located just below helical coil are analyzed to study the plasma edge boundary of the ergodic layer consisting of stochastic magnetic field lines with connection lengths of 30 ≤ Lc ≤ 2000 m. As a result, C3+ ion emitting CIV spectrum is identified as the ion existing in the farthest edge of the ergodic layer. The peak position of CIV (312.4 Å: 1s23p 2P1/2,3/2-1s22s 2S1/2) vertical profile does not change at all in a wide temperature range of 150 ≤ Te(ρ = 1) ≤ 400 eV, whereas it moves inside the ergodic layer when Te(ρ = 1) is reduced below a threshold temperature, e.g., 130 eV at Rax = 3.75 m configuration. It is found that the C3+ ion exists at the boundary between ergodic layer and open magnetic filed layer at which the Lc distributes in lengths of 5 to 30 m. The result indicates that the edge boundary near the O-point in LHD is determined by a starting point of the open filed layer, where a tokamak-like steeper edge temperature gradient is formed, although the edge boundary is quite obscure at the X-point region. Any plasma does not exist between the edge boundary and the vacuum vessel. The CIV profile at the O-point is simulated using a three-dimensional edge transport code of EMC3-EIRENE in which the magnetic field structure in vacuum is used for the ergodic layer. A clear discrepancy of 8 mm is found in the peak positions of CIV between measurement and simulation for magnetic configurations with thick ergodic layer, i.e., Rax =3.90 m, while only a small discrepancy of 3 mm is observed for those with relatively thin ergodic layer, i.e., Rax = 3.75 m. It suggests that the discrepancy is caused by a modification of the magnetic filed due to the presence of plasma pressure.

Ideal magnetohydrodynamic stability analysis of local pressure-driven modes in an L=1 heliotron, Heliotron J [M. Wakatani et al., Nucl. Fusion 40, 569 (2000)], is investigated by means of three-dimensional (3D) ballooning formalism and the Mercier criterion. In 3D systems such as heliotrons, the ballooning modes are separated into two categories: One is tokamak-like ballooning modes which are localized only in the poloidal direction, and the other is modes inherent to 3D systems which are localized on the specific flux tubes. The tokamak-like ballooning modes change to the Mercier modes in the limit that the mode is sufficiently extended along the field line, but the nonaxisymmetric ballooning mode does not so. The L=1 Heliotron J equilibrium investigated here has weak global shear and the dominant Fourier amplitudes of magnetic-field strength is rather different from the conventional helical systems with L=2 helical coils. Since the weak global shear causes the reduction of integrated local shear along the field lines easily, which combines with strongly modulated destabilizing effects on the flux surface, the nonaxisymmetric ballooning modes localized on the specific flux tubes can become unstable. On the other hand, the Mercier modes are suppressed due to the deep magnetic well. The results obtained from the model equilibrium of L=2 Large Helical Device (LHD), for which several reports have already published [N. Nakajima, Phys. Plasmas 3, 4556 (1996), for example], are also shown and compared with the results of Heliotron J. The LHD equilibrium employed here has a magnetic hill region at the outer radius, and this tends to make the Mercier modes unstable. It is found that this difference of the Mercier stability property in two equilibria is concerned in the local ballooning stability, and the notable difference of local dispersion relations appears. It is also found from the comparison of two systems that the nonaxisymmetric ballooning modes have a similar property to the tokamak-like ballooning modes, in the sense of the s̄-ᾱ diagram where s̄ and ᾱ are shear and pressure gradient parameter.

Mechanical analysis and fabrication of the R&D forced-flow helical coil (TOKI-PF)