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...

An integrated divertor simulation code SONIC has been developed. The self-consistent coupling of an MC impurity code IMPMC to a divertor code SOLDOR/NEUT2D is succeeded by overcoming the intrinsic problems of Monte Carlo (MC) modelling for impurity transport. MC modelling for impurity transport is required in order to take into account the kinetic effect and the complex dissociation processes of hydrocarbons. The integrated divertor code SONIC enables us to investigate the details of impurity transport including erosion/redeposition processes on the divertor plates by further coupling of an MC code EDDY. The dynamic evolution of X-point MARFE observed in JT-60U is investigated. The simulation results indicate that the hydrocarbons sputtered from the dome contribute directly to the enhanced radiation near the X-point. Without the recycling, the kinetic effect of the thermal force improves the helium compression, compared with the conventional (fluid) evaluation. This effect is, however, masked by the recycling at the divertor targets.

Power exhaust for a 3 GW class fusion reactor with an ITER-sized plasma was investigated by enhancing the radiation loss from seeding impurity. The impurity transport and plasma detachment were simulated under the Demo divertor condition using an integrated divertor code SONIC, in which the impurity Monte-Carlo code, IMPMC, can handle most kinetic effects on the impurity ions in the original formula. The simulation results of impurity species from low Z (neon) to high Z (krypton) and divertor length with a plasma exhausted power of 500 MW and radiation loss of 460 MW, and a fixed core–edge boundary of 7 × 1019 m−3 were investigated at the first stage for the Demo divertor operation scenario and the geometry design. Results for the different seeding impurities showed that the total heat load, including the plasma transport and radiation , was reduced from 15–16 MW m−2 (Ne and Ar) to 11 MW m−2 for the higher Z (Kr), and extended over a wide area accompanied by increasing impurity recycling. The geometry effect of the long-leg divertor showed that full detachment was obtained, and the peak qtarget value was decreased to 12 MW m−2, where neutral heat load became comparable to and due to smaller flux expansion. Fuel dilution was reduced but was still at a high level. These results showed that a divertor design with a long leg with higher Z seeding such as Ar and Kr is not fulfilled, but will be appropriate to obtain the divertor scenario for the Demo divertor. Finally, influences of χ and D⊥ enhancement were seen significantly in the divertor, i.e. the radiation and density profiles became wider, leading to full detachment. Both qtarget near the separatrix and Te at the outer flux surfaces were decreased to a level for the conventional technology design. On the other hand, the problem of fuel dilution became worse. Extrapolation of the plasma transport coefficients to ITER and Demo, where density and temperature will be higher than ITER and edge-localized modes are mitigated, is a key issue for the divertor design.

Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor.

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