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.

This paper describes the content of an L-mode database that has been compiled with data from Alcator C-Mod, ASDEX, DIII, DIII-D, FTU, JET, JFT-2M, JT-60, PBX-M, PDX, T-10, TEXTOR, TFTR, and Tore-Supra. The database consists of a total of 2938 entries, 1881 of which are in the L-phase while 922 are ohmically heated only (OH). Each entry contains up to 95 descriptive parameters, including global and kinetic information, machine conditioning, and configuration. The paper presents a description of the database and the variables contained therein, and it also presents global and thermal scalings along with predictions for ITER. The L-mode thermal confinement time scaling, determined from a subset of 1312 entries for which the T E ,~F , are provided, is

The article treats the recent development of quasi-steady ELMy high βp H mode discharges with enhanced confinement and high β stability, where long sustainment time, an increase in absolute fusion performance and extension of the discharge regime towards low q95 (∼3) are emphasized. After modification to the new W shaped pumped divertor, a long heating time (9 s) with a high total heating energy input of 203 MJ became possible without a harmful increase in impurity and particle recycling. In addition, optimization of the pressure profile characterized by the double transport barriers, optimum electron density and/or high triangularity δ made it possible to extend the performance in long pulses. The DT equivalent fusion gain Q eq DT ≈ 0.1 (δ = 0.16) was sustained for ∼9 s (∼50τE, ∼10τ * p ) and Q eq DT ≈ 0.16 (δ = 0.3) for 4.5 s at Ip = 1.5 MA. In the latter case with higher δ, an H factor (=τE/τ ITER89PL E ) of ∼2.2, βN ≈ 1.9 and βp ≈ 1.6 were sustained with 60-70% of the non-inductively driven current. In the low q95 (∼3) region, the β limit was improved by the high δ (∼0.46) shape, where βN ≈ 2.5-2.7 was sustained for ∼3.5 s with the collisionality close to that of ITER-FDR plasmas. The limit of the edge α parameter in the ELMy phase increases with δ, which is the main reason behind the improved β limit in a long pulse at high δ. The sustainable value of βN H also increases with δ. Sustainable βN is limited by the onset of low n resistive modes. Direct measurement of island width shows agreement with the neoclassical tearing mode theory.

Non-dimensional transport studies have been carried out for NBI heated ELMy H-mode plasmas in JT-60U. In a high-δ (triangularity) case, both the electron thermal diffusivity, χ e , and the ion thermal diffusivity, χ i , show gyroBohm-like diffusion, χ e /χ Bohm , χ i /χ Bohm ∝ (ρ * ) 0.7-0.9 , where ρ * is the normalized Larmor radius. The dependence of the energy confinement on the normalized collision frequency, ν * , was also studied. The weak ν * dependence of the energy confinement time, τ E /τ Bohm ∝ (1/ν * ) ∼0.3 , was also obtained in the high-δ case. On the assumption of the same ν * dependence of τ E /τ Bohm , the ρ * dependence of χ * in the low-δ case is compatible with the high-δ case. When the ELM activity is strong enough to affect the global energy confinement, however, the ion energy confinement was significantly degraded, although the electron energy confinement still follows the gyroBohm-like diffusion.

A new type of internal transport barrier (ITB) has been observed in JT-60U reversed shear discharges. It accompanies clear electron temperature and density pedestals and significant reduction of effective thermal diffusivities of electrons and ions (x eff e and x eff i ); x eff e sharply drops by a factor of 20 within 5 cm while x eff i is smaller than the conventional neoclassical value by a factor of 4 or more. The ratio of ion temperature to electron temperature was less than 1.5 inside the ITB. The ITB lies in the negative shear region and extended beyond 60% of the plasma minor radius. High density plasmas with high confinement have been obtained with the formation of ITB. [S0031-9007 (97)02736-1] PACS numbers: 52.55.Fa, 52.25.FiA reversed shear configuration, which has negative magnetic shear in the inner region and positive magnetic shear in the outer region, has been proposed as an advanced tokamak operation [1][2][3]. The magnetic shear is defined as ͑r͞q͒dq͞dr, where q is the safety factor and r is the volume-averaged minor radius. The reversed shear configuration has a possibility of economical steadystate tokamak reactor with high b, good confinement, and a large bootstrap current fraction. Here, b is the ratio of plasma pressure to magnetic field pressure. In experiments, reversed shear configurations were formed by pellet injection [4], initial current ramp with neutral beam (NB) heating [5-7], off-axis lower hybrid current drive (LHCD) [8], off-axis electron cyclotron heating (ECH) [9], and so on, and confinement improvement or formation of transport barrier in the negative shear region was observed. Furthermore, internal transport barriers were observed for the ion temperature in JT-60U high b p plasmas [10] (b p is the ratio of plasma pressure to poloidal magnetic field pressure), which had monotonic q profiles [11].In previous beam heating experiments [5,7,10], the reduction of the ion thermal diffusivity ͑x i ͒ or peaking of ion temperature ͑T i ͒ profile is reported but the reduction of electron thermal diffusivity ͑x e ͒ is not clear. In JET pellet enhanced performance mode, the reduction of x e is not known since the electron and ion power balances or x i and x e are not separated [12]. Clear reduction of x e in the negative shear region is reported in LHCD or ECH experiments [8,9]. However, it should be noted that the reduction of x e in these experiments is caused mainly by off-axis power deposition of lower hybrid or electron cyclotron waves rather than the spatial change of electron temperature ͑T e ͒ gradient. The inward electron heat flux (negative x e ) was observed in DIII-D off-axis ECH experiments [13], which is considered to be due to the profile resilience. This profile resilience effect may have some contribution to the reduction of x e in these experiments. Though the fast wave current drive experiments with reversed shear in indicates the reduction of x e with on-axis heating, x e decreases continuously toward the axis and the effect of negative magnetic shear on x e profile ...

The geometry effects of the W shaped divertor on the divertor plasma were investigated quantitatively. The ion flux was increased near the divertor strike point, which is effective for reducing the local electron temperature and decreasing the onset n̄e of divertor detachment. The plasma profile and parallel plasma flow in the scrape-off layer were systematically measured using reciprocating Mach probes installed at the midplane and the divertor X point. For the ion ∇B drift direction towards the divertor, `flow reversal' was observed at the midplane. A quantitative evaluation of the parallel plasma flow suggesting that the flow is produced in a torus to keep the pressure constant along the field lines was consistent with the measurements.

A compact formula for fully relativistic Thomson scattering spectrum including depolarization term is presented. By rational approximation, an analytic formula with high accuracy (relative error<0.1% at 100 keV) is obtained, which is applicable to a wide range of plasmas.

In JFT-2M, the ferritic steel plates (FPs) were installed inside the vacuum vessel all over the vacuum vessel, which is named Ferritic Inside Wall (FIW), as the third step of the Advanced Material Tokamak Experiment (AMTEX) program. A toroidal field ripple was reduced, however the magnetic field structure has become the complex ripple structure with a non-periodic feature in the toroidal direction because of the existence of other components and ports that limit the periodic installation of FPs. Under the complex magnetic ripple, we investigated its effect on the heat flux to the first wall due to the fast ion loss. The small heat flux was observed as the result of the reduced magnetic ripple by FIW. Additional FPs were also installed outside the vacuum vessel to produce the localized larger ripple. The small ripple trapped loss was observed when the shallow ripple well exist in the poloidal cross section, and the large ripple trapped loss was observed when the ripple well hollow out the plasma region deeply. The experimental results were almost consistent with the newly developed Fully three Dimensional magnetic field Orbit-Following Monte-Carlo (F3D OFMC) code including the three dimensional complex structure of the toroidal field ripple and the non-axisymmetric first wall geometry. By using F3D OFMC, we investigated the effect on the ripple trapped loss of the localized larger ripple produced by FPs in detail. The ripple well structure, e.g. the thickness of the ripple well, is important for ripple trapped loss in complex magnetic ripple rather than the value defined at one position in a poloidal cross section.