Kiyotaka Hamamatsu

Ripple induced fast ion loss and related effects in JT-60U

Experiments have been carried out in JT-60U to verify the modelling of fast ion ripple transport. The ripple induced loss was estimated from the neutron decay following neutral beam pulse injection and the loss related heat load on the first wall. Comparison of the lost fraction and the hot spot positions between measurements and orbit following Monte Carlo calculations exhibited good agreement, indicating that the ripple transport governing fast ion losses is explained within the framework of existing theory. Neutral beam heating experiments in JT-60U also indicate that H modes free of ELMs are still obtainable for ripple amplitudes of up to 2.2%

Off-axis current drive and real-time control of current profile in JT-60U

Aiming at optimization of current profile in high-β plasmas for higher confinement and stability, a real-time control system of the minimum of the safety factor (qmin) using the off-axis current drive has been developed. The off-axis current drive can raise the safety factor in the centre and help to avoid instability that limits the performance of the plasma. The system controls the injection power of lower-hybrid waves, and hence its off-axis driven current in order to control qmin. The real-time control of qmin is demonstrated in a high-β plasma, where qmin follows the temporally changing reference qmin,ref from 1.3 to 1.7. Applying the control to another high-β discharge (βN = 1.7, βp = 1.5) with m/n = 2/1 neo-classical tearing mode (NTM), qmin was raised above 2 and the NTM was suppressed. The stored energy increased by 16% with the NTM suppressed, since the resonant rational surface was eliminated. For the future use for current profile control, current density profile for off-axis neutral beam current drive (NBCD) is for the first time measured, using the motional Stark effect diagnostic. Spatially localized NBCD profile was clearly observed at the normalized minor radius ρ of about 0.6–0.8. The location was also confirmed by multi-chordal neutron emission profile measurement. The total amount of the measured beam driven current was consistent with the theoretical calculation using the ACCOME code. The CD location in the calculation was inward shifted than the measurement.

Design optimization for plasma performance and assessment of operation regimes in JT-60SA

The design of the modification of JT-60U, JT-60SA has been optimized from the viewpoint of plasma performance, and operation regimes have been evaluated with the latest design. Upper and lower divertors with different geometries will be prepared for flexibility of the plasma shape, which will enable both low aspect ratio (A ~ 2.65) and ITER shape (A = 3.1) configurations. The beam lines of negative-ion neutral beam injection will be shifted downwards by ~0.6 m for the off-axis current drive (CD), in order to obtain a weak/reversed shear plasma, as well as having the capability of heating the central region. The feedback control coils along the openings in the stabilizing plate are found effective in suppressing the resistive wall mode and sustaining high βN close to the ideal wall limit. Sustainment of plasma current of 3–3.5 MA for 100 s will be possible in ELMy H-mode plasmas with moderate heating power, βN, and density within an available flux swing. It is also expected that higher βN, high-density ELMy H-mode plasmas will be maintained for 100 s with higher heating power. The expected regime of full CD operation has been extended with upgraded heating and CD power. Full CD operation for 100 s with reactor-relevant high values of normalized beta and bootstrap current fraction (Ip = 2.4 MA, βN = 4.3, fBS = 0.69, , HH98y2 = 1.3) is expected in a highly-shaped low-aspect-ratio configuration (A = 2.65).

Design study of National Centralized Tokamak facility for the demonstration of steady state high-β plasma operation

Design studies are shown on the National Centralized Tokamak facility, formerly called JT-60SC. The machine design is carried out to investigate the capability for flexibility in aspect ratio and shape controllability for the demonstration of the high-β steady state operation with nation-wide collaboration, in parallel with ITER towards DEMO. Two designs are proposed and assessed with respect to the physics requirements such as confinement, stability, current drive, divertor and energetic particle confinement. The operation range in the aspect ratio and the plasma shape is widely enhanced consistent with sufficient divertor pumping. Evaluations of the plasma performance towards the determination of the machine design are presented.

Overview of the National Centralized Tokamak programme

An overview is given of the National Centralized Tokamak (NCT) programme as a research programme for advanced tokamak research to succeed JT-60U. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility is pursued in aspect ratio and shape controllability for the demonstration of the high-β steady-state, feedback control of resistive wall modes, wide current and pressure profile control capability and also very long pulse steady-state operation. Existing JT-60 infrastructure such as the heating and current drive system, power supplies and cooling systems will be best utilized for this modification.

Ripple enhanced banana drift loss at the outboard wall during ICRF/NBI heating in JT-60 U

Ripple enhanced banana drift loss of fast ions is studied on the basis of heat load measurements at the first wall near the outer equatorial plane in JT-60 U. Hot spots in the ion cyclotron range of frequency (ICRF) and neutral beam injection (NBI) heating are observed at the outboard wall by using an infrared camera. In ICRF heating, the total heat load to the wall is around 5% of the injected ICRF power when the plasma-wall clearance gap is less than 0.1 m. The heat flux on the wall increases with the ICRF power and decreases strongly with the plasma density, which is consistent with Fokker-Planck calculations. In NBI heating, hot spots are observed at almost the same positions, and the heat flux and position agree with the banana drift loss as predicted by an orbit following Monte Carlo code

High-harmonic ICRF heating experiments in JT-60

A review of the results from JT-60 ICRF heating experiments is presented. The JT-60 ICRF experiments (120 and 131 MHz) are characterized by higher-harmonic heating with phase control of the compact 2*2 loop antenna array. Second-harmonic heating is most intensively investigated with antenna phase control for various cases where the resonant species (hydrogen) are either a majority component or a minority component, and with or without hydrogen HBI heating. It is found that antenna phase control plays a key role in optimizing antenna power injection characteristics as well as heating efficiency. The out-of-phase mode (( pi ,0) mode) can avoid RF sheath and parametric decay instabilities which cause harmful effects on antenna power injection capability with the in-phase mode ((0,0 ) mode) is some operating conditions. The incremental energy confinement time ( tau Einc) of the ( pi ,0) mode is about 50% better than that of the (0,0) mode. Minority hydrogen second-harmonic heating with ( pi ,0) phasing in helium discharges shows the best results over a wide range of electron densities (ne=2.5-6.5*1019 m-3) and plasma currents (Ip=1.5-2.4 MA (qa=4.2-2.6)). The best confinement enhancement factor ( tau E/ tau E(L-mode)) is 1.3, which is obtained with ohmic target plasmas. Higher-harmonic beam ion acceleration is observed up to fourth harmonics in a combined ICRF (( pi ,0) mode) and NBI heating scenario. Strong central-electron heating associated with beam ion acceleration is observed in combined third-harmonic ICRF and NBI heating.

Numerical Analysis of Ripple Loss in Reversed Shear Tokamak Operation.

Ripple loss of energetic alpha particles and neutral beam ions was calculated for reversed shear discharges in ITER-FDR(Final Design Report). The result indicates that, compared with normal operation with positive shear, reversed shear operation dramatically enhances the ripple loss. Ripple loss of alpha particles can reach 25 % with the maximum heat load of 3.7 MW/m2 on the outboard wall, while that of a 1 MeV negative ion source beam can be as high as 20 % as well. The heat load due to alpha particle loss is marginal in the light of a wall tolerance. The calculation suggests that the Toroidal Field (TF) ripple in a fusion reactor should be designed to be less than 0.6 % at the plasma surface so that ripple loss of alpha particles or beam ions can be acceptably low. Ferritic steel insert to the vacuum vessel, is a probable solution to reduce the ripple loss to an allowable level in the ITER-FDR design with the TF ripple reduction of by a factor of 0.4. With TF ripple reduction, ripple loss for alpha particles and beam ions is expected to decrease to 10 % and 4 %, respectively.

Effect of re-entering fast ions on NBI heating power in high-beta plasmas of the Large Helical Device

We calculate the heating power of the neutral beam injection (NBI) in a 〈β〉 = 4.8% high-beta discharge achieved in the Large Helical Device (LHD). We investigate the difference of the heating efficiency and the heating power profile with and without the re-entering fast ion effects. When the re-entering fast ion effects are taken into account, the heating efficiency of the co-injection of the NBI (co-NBI case) is improved and it is about 1.8 times larger than that without the re-entering effects. In contrast, the heating efficiency with the re-entering effects in the counter-injection of the NBI (ctr-NBI case) scarcely differs from that without the re-entering ones. We also study the re-entering fast ion effects on the transport properties in the LHD high-beta discharges. It is found that the tendency of the thermal conductivities on the beta value is not so much sensitive with and without the re-entering effects. In addition, we investigate the difference in the re-entering fast ion effects caused by the field strength and the magnetic configuration. In the co-NBI case, the heating efficiency with the re-entering effect was improved with a decrease in the field strength. In contrast, in the ctr-NBI case, it barely differs by changing the field strength.

Comparison between theoretical analyses and experimental results of an ICRF loading in JT-60

A cavity resonance of ICRF fast wave was observed in early experiments on the second harmonic ICRF heating in JT-60. The purpose of these experiments was to confirm an antenna-plasma coupling. We discuss the loading impedance of ICRF waves in a linear theory, since the loading condition was in a low-power regime. The experimental results are compared with a numerical analysis of one-dimensional ICRF wave model. Good agreement between the experimental results and the theoretical analysis is obtained qualitatively.