Two data points in the shaded zones of Figs. 3(a) and 3(b) (one point for each figure), were omitted in the published version of these figures. The correct Fig. 3 is shown below.FIG. 3. Radial profiles of (a) electron density, ( b) electron temperature, and (c) ion temperature at 6 s and (d) safety factor at 5.9 s of the discharge shown in Fig. 2. The volume-averaged plasma minor radius is 1.01 m.

The structure of the edge radial electric field E r inferred from the poloidal rotation velocity is compared with that of the particle and thermal transport barrier for //-mode plasmas in JFT-2M. Both E r and its gradient dE r /dr in the thermal transport barrier are found to become more negative at the L-H transition. On the other hand, dE r /dr is more positive outside of the separatrix. The shear of the radial electric field and poloidal rotation velocity in the H mode is localized within the order of an ion poloidal gyroradius near the separatrix, in the region of ion collisionality v*, ~ 20-40.PACS numbers: 52.55. Fa, 52.55.Pi, 52.70.Kz Since the //-mode plasma was discovered in ASDEX, it has been observed in many tokamaks. 2 " 0 Several theoretical models on the transition from L-mode to Hmode plasmas have been presented. 6 " 11 Recently, a radial electric field (E r ) near the plasma periphery has been found both experimentally and theoretically to play an important role in the L-H transition. 12 " 19 A more negative radial electric field was observed a few ms before the L-H transition in DIII-D (Ref. 12), and a decrease in particle transport was observed with negative E r , by driving a radial current, in the Continuous Current Tokamak. 13 Theoretical models associated with the radial electric field have been proposed to explain the L-H transition. 14 " 17 However, the predicted change of the gradient of the radial electric field (dE r /dr) is different between the models. In Shaing and Crume's model, 16 the poloidal flow velocity changes at the L-H transition and the corresponding radial electric field E r becomes more negative and dE r /dr becomes more positive, hence suppressing the fluctuations. On the other hand, Itoh and Itorfs model 17 predicts positive values of dE r /dr in the L mode and negative values of dE r /dr in the H mode, and that this negative dE r /dr reduces the banana width of the ions and the electron anomalous flux by the improved microstability. Thus it is crucial to measure the gradient or profile of the radial electric field for Land //-mode plasmas in tokamaks.In this paper we present the radial electric-field profile and temperature gradient profile a few cm inside the separatrix where the transport barrier is produced in Hmode plasmas in JFT-2M. 5 The radial electric-field profiles are inferred from poloidal and toroidal rotation velocity profiles and ion pressure profiles using the ionmomentum-balance equation, eZ { n, orwhere Z,, /?,, and n, are the ion charge, pressure, and density, B^ and B e are the toroidal and poloidal magnetic fields, and i> and v e are the toroidal and poloidal rotation velocities. The toroidal rotation velocity, ion temperature, and fully stripped carbon density profiles are measured using a multichannel charge-exchangespectroscopy technique 18,19 at Cvi 5292 A with toroidal arrays (two sets of 34 channels) with a spatial resolution of 1 cm. The poloidal rotation velocity and edge ion temperature profiles are measured using the intrinsic radiat...

Characteristics of internal transport barrier (ITB) structure are studied and the active ITB control has been developed in JT-60U reversed shear plasmas. The following results are found. Outward propagation of the ITB with steep T i gradient is limited to the minimum safety factor location (ρ qmin). However the ITB with reduced T i gradient can move to the outside of ρ qmin. Lower boundary of ITB width is proportional to the ion poloidal gyroradius at the ITB center. Furthermore the demonstration of the active control of the ITB strength based on the modification of the radial electric field shear profile is successfully performed by the toroidal momentum injection in different directions or the increase of heating power by neutral beams.

IAEA-CN-116/EX/2-1 _______________________________________________________________________________________ This is a preprint of a paper intended for presentation at a scientific meeting. Because of the provisional nature of its content and since changes of substance or detail may have to be made before publication, the preprint is made available on the understanding that it will not be cited in the literature or in any way be reproduced in its present form. The views expressed and the statements made remain the responsibility of the named author(s); the views do not necessarily reflect those of the government of the designating Member State(s) or of the designating organization(s). In particular, neither the IAEA nor any other organization or body sponsoring this meeting can be held responsible for any material reproduced in this preprint.

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma. The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.