No abstract

used. Quantitative spectroscopic measurements of Z eff , impurity concentrations and radiated power losses were made for ohmically heated and neutral beam heated discharges with limiter and divertor configurations. In the first phase with metallic first wall material, oxygen, carbon and titanium were identified as the main plasma impurities. In neutral beam heated, diverted discharges, Z efr was 1.6 at n,. = 4xlO 1 9 m" 3 . The concentrations of oxygen, carbon and titanium were 1%, 0.1% and 0.006% of n e , respectively. In the second phase with graphite material, the metallic impurities were reduced, and the contribution of metallic impurities to the radiated power loss was less than 1%. However, Z eff increased up to 3 in neutral beam heated discharges. In limited plasmas, the concentrations of oxygen and carbon were 1% and 5%, respectively, at Sj = 4xlO 1 9 m" 3 ; in diverted plasmas, these concentrations were 2% and 0.4% at the same r^. The radiated power loss from the main plasma was 20-40% of the input power in neutral beam heated, limited discharges, and 7-25% in diverted discharges. The contributions of oxygen and carbon to the radiated power in limited discharges were comparable, and in diverted discharges the contribution of oxygen was dominant.

From June to October 1987, JT-60 achieved fusion product(ne(0) .r~*.Ti(Oj ) of 6 ~1 0 ' ~ m 3 . k e V . s with hydrogen plasma at plasma current of 2.8 to 3.1 MA with neutral beam power of -20 MW. The central electron density of 1 . 3 ~1 0 2 0 m-3 was obtained at plasma current of 3 MA with 13-20 MW neutral beam power and the confinement time reached 0.14-0.18 s. It is found that an offset linear scaling law like the Shimomura-Odajima scaling on confinement time will be able to reproduce experimental data better than that of the Goldston type scaling. With low beam energy injection, -40 keV, confinement degradation was found. Many short periods (0.05-0.1 s) of H-mode phase were found in outside X-point divertor discharges with NB or NB+RF(LH or IC) heating power above 16 MW. However, improvement in energy confinement time was limited to 10 %. The ballooning/interchange stability analysis were JT-60 TEAM also made for the outside X-point divertor equilibrium in connection with H-phase capability. Heating power of 9.5 MW and 1.9 MW was obtained by LHRF, ICRF injection, respectively. In combined LHRF and NB heating, the incremental energy confinement time of 0.064 s was obtained, which is the same level of that of NB heating only. In combined NB and on-axis ICRF heating of low ne discharge, an incremental energy confinement time of 0.21 s was obtained, which is three times as long as those of NB or ICRF heating only. It was also observed that high energy beam ions were accelerated by ICRF in the central region of the plasma.

The active beam scattering method was applied to ion temperature measurements in JT-60 plasmas. The ion temperature found was in reasonable agreement with that obtained from Doppler broadening of the Ti XXI and Ti XXII resonance lines in the temperature range of 1.5 to 10 keV. With the help of the scattering system, the central ion temperature of JT-60 plasmas during 40 keV and 70-75 keV NBI heating was measured. Higher ion temperature was obtained during 70 to 75 keV than during 40 keV NBI heating. These data and a possible explanation are presented. A numerical simulation of this diagnostic indicates that the ion temperature as deduced from the diagnostic is almost equal to the bulk temperature when the ratio of high energy ion component to the total ion stored energy is less than 0.3.

Emphases in recent JT-60 experiments are placed on 1)lower-hybrid (LH) current drive characteristics with a multi-junction type launcher and 2 ) the confinement study with combination of neutral beam injection, LH current drive and pellet injection. The new multi-junction LH launcher provides a 2 . 7 x 1 0 1 9 m -3 and Ip=1-1.7 MA.Volt-sec saving of -2volt-sec was demonstrated by 2 sec long, 1.6MW LHCD during the plasma current ramp of 0.4MA/s.A broad radial distribution of high energy electron current and -30% reduction in sawtooth inversion radius were obtained by high N I I (-2.5) LH injection. In order to fully suppress the sawtooth activity, low NII (-1.3) injection was found to be more effective, in which up to 1.8 sec sawtooth-free phase was obtained by 2MW LHCD for lOMW NB heating of Ip=I.SMA discharge, Improved energy confinement has been obtained with hydrogen pellet injection. Energy confinement time was enhanced up to 40% relative to usual gas fuelled discharges. The discharge has a strongly peaked electron density profile with ne(O)/-5 and n e ( 0 ) -2 . 0 ~ 1020,-3,The improved discharges are characterized by a strongly peaked pressure profile within the q=l magnetic surface, and degrades when a large sawtooth recovers or the pressure gradient may reach a critical value. When large (3mm,4mm) and fast (2.2km/s) pellets were injected, 30% energy confinement improvement was obtained even during the NB heating of 14MW. Further investigations of IDC characteristics have been made.The oxygen impurity lines from the main plasma and the main radiative loss drop first. Then the plasma stored energy starts to rise. The particle recycling is reduced around the main plasma, and is localized in the neighborhood of the X-point with a time lag of -0.2 sec. Eventually the discharge shows a significant remote radiative cooling power at the divertor region. Most of the IDC characteristics is totally different from that of H-mode, although there are some similarities i.e., gas puff degrades the confinement, both have the same favorable BT direction.