Jt Team

Characteristics of internal transport barriers in JT-60U reversed shear plasmas

The characteristics of internal transport barrier (ITB) structures are studied and active ITB control has been developed in JT-60U reversed shear plasmas. The following results are found. Outward propagation of ITBs with steep Ti gradients is limited to the minimum safety factor location ρqmin. However, ITBs with reduced Ti gradients can move to the outside of ρqmin. The lower boundary of the ITB width is proportional to the ion poloidal gyroradius at the ITB centre. Furthermore, active control of the ITB strength based on modification of the radial electric field shear profile is successfully demonstrated by toroidal momentum injection in different directions or an increase of heating power by neutral beams.

Neoclassical tearing mode control using electron cyclotron current drive and magnetic island evolution in JT-60U

The results of stabilizing neoclassical tearing modes (NTMs) with electron cyclotron current drive (ECCD) in JT-60U are described with emphasis on the effectiveness of the stabilization. The range of the minimum EC wave power needed for complete stabilization of an m/n = 2/1 NTM was experimentally identified for two regimes using unmodulated ECCD to clarify the NTM behaviours with different plasma parameters: 0.2 < jEC/jBS < 0.4 for Wsat/dEC ~ 3 and Wsat/Wmarg ~ 2, and 0.35 < jEC/jBS < 0.46 for Wsat/dEC ~ 1.5 and Wsat/Wmarg ~ 2. Here, m and n are the poloidal and toroidal mode numbers; jEC and jBS the EC-driven current density and bootstrap current density at the mode rational surface; Wsat, Wmarg and dEC the full island width at saturation, marginal island width and full-width at half maximum of the ECCD deposition profile, respectively. Stabilization of a 2/1 NTM using modulated ECCD synchronized with a mode rotation of about 5 kHz was performed, in which it was found that the stabilization effect degrades when the phase of the modulation deviates from that of the ECCD at the island O-point. The decay time of the magnetic perturbation amplitude due to the ECCD increases by 50% with a phase shift of ±50° from the O-point ECCD, thus revealing the importance of the phasing of modulated ECCD. For near X-point ECCD, the NTM amplitude increases, revealing a destabilization effect. It was also found that modulated ECCD at the island O-point has a stronger stabilization effect than unmodulated ECCD by a factor of more than 2.

Alfvén eigenmodes driven by Alfvénic beam ions in JT-60U

Instabilities with frequency chirping in the frequency range of Alfven eigenmodes have been found in the domain 0.1% < βh < 1% and vb||/vA ~1 with high energy neutral beam injection in JT-60U. One instability with a frequency inside the Alfven continuum spectrum appears and its frequency increases slowly to the toroidicity induced Alfven eigenmode (TAE) gap on the timescale of an equilibrium change ( ≈ 200 ms). Other instabilities appear with a frequency inside the TAE gap and their frequencies change very quickly by 10-20 kHz in 1-5 ms. During the period when these fast frequency sweeping (fast FS) modes occur, abrupt large amplitude events (ALEs) often appear with a drop of neutron emission rate and an increase in fast neutral particle fluxes. The loss of energetic ions increases with a peak fluctuation amplitude of θ/Bθ. An energy dependence of the loss ions is observed and suggests a resonant interaction between energetic ions and the mode.

High performance experiments in JT-60U reversed shear discharges

The operation of JT-60U reversed shear discharges has been extended to a high plasma current, low q regime keeping a large radius of the internal transport barrier (ITB), and a record value of equivalent fusion multiplication factor in JT-60U, QDTeq = 1.25, has been achieved at 2.6 MA. Operational schemes to reach the low q regime with good reproducibility have been developed. The reduction of Zeff was obtained in the newly installed W shaped pumped divertor. The β limit in the low qmin regime, which limited the performance of L mode edge discharges, has been improved in H mode edge discharges with a broader pressure profile, which was obtained by power flow control with ITB degradation. Sustainment of the ITB and improved confinement for 5.5 s has been demonstrated in an ELMy H mode reversed shear discharge.

Progress in confinement and stability with plasma shape and profile control for steady-state operation in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade

This paper describes the latest achievements of the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [H. Kimura and the JT-60 Team, Phys. Plasmas 3, 1943 (1996)] for the establishment of the physics and technology basis for steady-state, fully noninductive current-drive plasmas with high fusion performance. Recent results are highlighted by (1) high performance reversed-magnetic-shear discharges [an idealized equivalent QDT of 1.05, a confinement-enhancement factor (H factor) of 3.23, and a normalized beta value (βN) of 1.88 at Ip=2.8 MA/Pabs=17 MW]; the first observation of thermal transport barriers both for electrons and ions, (2) the first injection of negative-ion-based neutral deuterium beams in the world for studies of current drive, heating, and high-energy particle behavior (achieved so far: 2.5 MW/400 keV, design values: 10 MW/500 keV); high neutralization efficiency of 60% at 370 keV; high current drive efficiency of ηCD=8×1018 m−2 A/W, (3) improved giant edge-localized-mode (...

Electron cyclotron heating assisted startup in JT-60U

Electron cyclotron heating (ECH)-assisted startup experiments have been performed in JT-60U. The breakdown loop voltage was successfully reduced from 25 to 4 V (=0.26 V m−1) by 200 kW ECH. This is lower than the 0.3 V m−1, which corresponds to the maximum electric field in ITER. Parameter scans of ECH power, prefill pressure, resonance position and polarization were carried out. The sensitivity of the breakdown to polarization and resonance position was observed. A prefilling gas pressure scan showed that the initial breakdown density increases with prefill pressure when it is is lower than 8 × 10−5 Torr. Higher harmonic ECH was also attempted. The second harmonic ECH-assisted startup was possible with higher ECH power injection. However, the third harmonic ECH-assisted startup was not successful.

Stabilization effect of early ECCD on a neoclassical tearing mode in the JT-60U tokamak

Stabilization of an m = 3/n = 2 neoclassical tearing mode (NTM) has been studied experimentally by applying a local electron cyclotron current drive (ECCD) in the JT-60U tokamak. The EC power is injected before the mode onset, and its effect is compared with the ECCD applied at the saturation phase. The experimental results show that the ECCD applied at the growth phase is more effective than that applied at the saturation phase. The necessary EC power for the suppression is reduced and the mode onset is delayed, indicating the hysterisis characteristics of the NTM on the ECCD stabilization. The dependence on the EC power and injection angle is also shown.

JT-60U high performance regimes

High performance regimes of JT-60U plasmas are presented with an emphasis upon the results from the use of a semiclosed pumped divertor with a W shaped geometry. Plasma performance in transient and quasi-steady states has been significantly improved in reversed shear and high βp regimes. The reversed shear regime elevated an equivalent QeqDT transiently up to 1.25 (nD(0) τETi(0) = 8.6 × 1020m-3skeV) in a reactor relevant thermonuclear dominant regime. Long sustainment of enhanced confinement with internal transport barriers (ITBs) with fully non-inductive current drive in a reversed shear discharge was successfully demonstrated with LH wave injection. Performance sustainment has been extended in the high βp regime with high triangularity, achieving a long sustainment of plasma conditions equivalent to QeqDT ≈ 0.16 (nD(0) τETi(0) ≈ 1.4 × 1020 m-3skeV) for ~4.5 s with a large non-inductive current drive fraction of 60-70% of the plasma current. Thermal and particle transport analyses show significant reduction of thermal and particle diffusivities around the ITB, resulting in a strong Er shear in the ITB region. The W shaped divertor is effective for helium ash exhaust, demonstrating a steady exhaust capability of τ*He/τE ≈ 3-10, which is favourable for ITER. Suppression of neutral backflow and chemical sputtering effects has been observed, while MARFE onset density is rather decreased. Negative ion based neutral beam injection (N-NBI) experiments have created a clear H mode transition. An enhanced ionization cross-section due to multistep ionization processes was confirmed as theoretically predicted. A current density profile driven by N-NBI is measured in good agreement with the theoretical prediction. N-NBI induced TAEs characterized as persistent and bursting oscillations have been observed from a low hot β of βh ≈ 0.1-0.2% without a significant loss of fast ions.

Recent experimental and analytic progress in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade with W-shaped divertor configuration

The Japan Atomic Energy Research Institute (JAERI) completed the divertor modification of the JAERI Tokamak-60 Upgrade (JT-60U) [Y. Koide and the JT-60 Team, Phys. Plasmas 4, 1623 (1997)] in June, 1997, and experiments with the W-shaped pumped divertor have been carried out. The helium exhaust was demonstrated in the steady state by using helium neutral beam injection (NBI). The ratio of the effective particle confinement time of helium to the energy confinement time, τHe*/τE, became about 4 and the obtained enrichment factor of helium in the divertor was about 1. The high confinement mode (H-mode) plasma with edge localized mode (ELM) was sustained for 9 s with a confinement–enhancement factor (H factor) of ∼1.7. No increase of recycling and carbon impurity density was observed with the integrated NBI input energy of 203 MJ. In the reversed shear (RS) experiments, a quasi-steady-state ELMy H mode was achieved with an H factor ∼2.4. The stability of high-n toroidal drift modes was analyzed in the RS plasm...

Current ramps in tokamaks: from present experiments to ITER scenarios

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm–gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H96−L = 0.6 or HIPB98 = 0.4) has been validated on a multi-machine experimental dataset for predicting the li dynamics within ±0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi–Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than ±0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of Ip = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down.