The measurements of the SOL flow and plasma profiles both at the high-field-side (HFS) and low-field-side (LFS), for the first time, identified the SOL flow pattern and its driving mechanism. 'Flow reversal' was found near the HFS and LFS separatrix of the main plasma for the ion ∇B drift direction towards the divertor. 'Flow reversal' at the main SOL was reproduced numerically using the UEDGE code with the plasma drifts included although Mach numbers in measurements were greater than those obtained numerically. Particle fluxes towards the HFS and LFS divertors produced by the parallel SOL flow and E r × B drift flow were evaluated from the measured profiles of Mach numbers, the density and the radial electric field. The drift flux in the private flux region was also evaluated, and it was found that its contribution to the HFS-enhanced asymmetry of the divertor particle flux was larger than the ion flux from the HFS SOL. The ion flux for the intense gas puff and divertor pump ('puff and pump') was investigated, and it was found that both the Mach number and density were enhanced, in particular, at the HFS. Ion flux at the HFS SOL can be enhanced so as to become larger than the drift flux in the private flux region.

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The design progress in a compact low aspect ratio (low A) DEMO reactor, ‘SlimCS’, and its design issues are reported. The design study focused mainly on the torus configuration including the blanket, divertor, materials and maintenance scheme. For continuity with the Japanese ITER-TBM, the blanket is based on a water-cooled solid breeder blanket. For vertical stability of the elongated plasma and high beta access, the blanket is segmented into replaceable and permanent blankets and a sector-wide conducting shell is arranged inbetween these blankets. A numerical calculation indicates that fuel self-sufficiency can be satisfied when the blanket interior is ideally fabricated. An allowable heat load to the divertor plate should be 8 MW m−2 or lower, which can be a critical constraint for determining a handling power of DEMO.

An integrated SOL/divertor code is being developed by the JAEA (Japan Atomic Energy Agency) for interpretation and prediction studies of the behavior of plasmas, neutrals, and impurities in the SOL/divertor region. A code system consists of the 2D fluid code for plasma (SOLDOR), the neutral Monte-Carlo code (NEUT2D), the impurity Monte-Carlo code (IMPMC), and the particle simulation code (PARASOL). The physical processes of neutrals and impurities are studied using the Monte Carlo (MC) code to accomplish highly accurate simulations. The so-called divertor code, SOLDOR/NEUT2D, has the following features: 1) a high-resolution oscillation-free scheme for solving fluid equations, 2) neutral transport calculation under the condition of fine meshes, 3) successful reduction of MC noise, and 4) optimization of the massive parallel computer. As a result, our code can obtain a steady state solution within 3 ∼ 4 hours even in the first run of a series of simulations, allowing the performance of an effective parameter survey. The simulation reproduces the X-point MARFE (multifaceted asymmetric radiation from edge) in the JT-60U. It is found that the chemically sputtered carbon at the dome causes radiation peaking near the X-point. The performance of divertor pumping in the JT-60U is evaluated based on particle balances. In regard to the divertor design of the next tokamak of JT-60U, the simulation indicates the dependencies of pumping efficiency on the divertor geometry and operational conditions. The efficiency is determined by the balance between the incident and back-flow fluxes into and from the exhaust chamber.

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.

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