[en] In LHD, edge radiation distribution and intensity control experiments have been conducted by changing thickness of the edge stochastic layer and by applying RMP field. Following experimental and modeling results were obtained. The RMP application enhances edge impurity radiation and induces stable detachment. This is considered due to increase in volume of radiating layer caused by edge magnetic island. The VUV spectroscopy measurement indicates that the enhancement occurs at both X- and O-point of the edge island. The edge visible spectroscopy measurement shows that sustainment of enhanced low charge state emission of carbon (C¹⁺) and enhanced volume recombination, are kept outside of the confinement region. Divertor particle and power load estimated by Langmuir probe is reduced significantly with increased cross-field transport towards private region, and with toroidal modulation in accordance with the mode number of the RMP. Without the RMP, these radiation layer penetrates into confinement region and leads to termination of discharge. Effects of the edge magnetic topology (O- and X-point of island) on radiation stabilization is conducted by using simple model.

[en] The compatibility of dissipative, detached divertor operation with high confinement core plasma scenarios is reviewed. The divertor and SOL plasma in future DEMO-scale tokamaks must simultaneously satisfy boundary conditions at the divertor target and at the upstream separatrix interface to the core plasma. While the requirements for the divertor target plasma to avoid excessive erosion and melting are well established, the upstream separatrix requirements are less well defined. The upstream parameters separatrix electron and impurity densities are particularly important for designing and assessing the viability of future divertor solutions.

[en] Japanese and European fusion research communities have developed the JT-60SA research plan, which summarizes strategies for JT-60SA, a superconducting large tokamak, for addressing issues for ITER and DEMO. In order to explore any possibilities of DEMO operational regime, JT-60SA will drive fusion research with a wide spectrum. For this, flexible upgrade plans for the hardware such as heating systems and divertor plasma facing components are considered. In addition, the research plan will enhance contribution to the studies for ITER support and DEMO steady-state operation, considering the latest ITER timeline.

[en] Spreading the heat load at stick point one of the most important works for increase the plasma performance. The Snowflake divertor(SFD) configuration could increase the flux expansion near the stick points, which fact for spreading the divertor heat load .Snowflake divertor configuration requires not only the first-order null at X-point as usual divertor configuration, but also the second-order derivative of the poloidal flux at X-point to be zero. Due to the Poloidal Field (PF) coils current limit, standard SFD configuration is difficult for EAST. One Quasi-SFD configuration can keeps the current within the allowable range of the coils.

[en] For the first wall of a fusion reactor unique challenges on materials in extreme environments require advanced mechanical and thermal properties. Tungsten (W) is the main candidate material for the first wall of a fusion reactor as it is resilient against erosion, has the highest melting point of any metal and shows rather benign behavior under neutron irradiation. However, the intrinsic brittleness of tungsten is a concern in respect with the fusion environment with high transient heat loads and neutron irradiation. Neutron induced effects e.g. transmutation add to embrittlement and are crucial to material performance. To overcome this drawback, tungsten fiber reinforced tungsten (Wf/W) composites are being developed relying on an extrinsic toughing principle. Accordingly, even in the brittle regime this material allows for a certain tolerance towards cracking and damage in general in comparison to conventional tungsten.

[en] Aim of this paper is to investigate a divertor power load in DEMO reactor with liquid tin divertor dependence on perpendicular scrape-off layer (SOL) transport coefficient and density at the separatrix. The simulation is performed with the COREDIV code, which self-consistently solves radial 1D energy and particle transport equations of plasma and impurities in the core region and 2D multifluid transport equations in the SOL. Influence of the sputtering, evaporation and prompt redeposition is taken into account. The evaporation rate depends on the energy flux to the divertor target and the amount of tin released is, thereby, coupled with the plasma in the divertor region. The heat deposition due to re-condensation of Sn is presently neglected as the Sn vapour distributes the heat over a large area or is absorbed in the pumps. Increased heat flux leads to more evaporation and hence more Sn in the SOL. Vapors radiate and redistribute significant part of the energy flux to the plate creating a vapor shield.

[en] A critical issue for ITER and future fusion devices is the handling of excessive heat load. It will require a high radiation level for power exhaust to avoid divertor heat overload and excessive surface erosion rates. Increasing divertor radiation by injecting impurities is a general and effective method to reduce scrape-off layer heat flux and to cool the divertor plasma to detachment. In recent years, argon (Ar) and neon (Ne) have been widely used in radiative divertor experiments on EAST and other devices. To compare effects of these two impurities and understand well the radiative divertor physics, both experiment and simulation work of Ar and Ne seeded plasma are carried out simultaneously in EAST.

[en] One of the conclusions of the DTT workshop held in Frascati on 19-20 June 2017 was that the design of a Divertor Tokamak Test facility [1-3] should offer sufficient flexibility to be able to incorporate the best candidate divertor concept (e.g. conventional, snowflake, super-X, double null, liquid limiter) even at a later stage of its realization. In this respect, four main points have been risen concerning the present status of the proposal and the possible directions to be explored: i) flexibility of the machine, as stated above; ii) up-down symmetry, so as to properly test double null configurations; iii) minimum amount of the additional power coupled to the plasma needed to guarantee significant results in view of DEMO; iv) the best additional power mix (coming from the various sources: ICRH, ECRH and NBI) in view of the flexibility requirements.

[en] Full text: The need for advanced divertor solutions to efficiently dissipate heat from fusion reactors is critical because the maximum steady-state power load is limited to q_t ≤5– 10 MWm −² to a plasma facing component (PFC) target surface, whether solid or liquid, while the undissipated power loads will be an order of magnitude higher. Previous work shows that the tightly closed divertor greatly improves trapping of recycling neutrals, thereby increasing radiative and charge exchange losses in the divertor and reducing the electron temperature Tet and deposited power density q_dep at the target plate. We studied the local closure effect of the divertor structure on the detachment density in CFETR by using the scrape-off layer plasma simulation (SOLPS) code. The divertor geometry near the striking point was set as a semicircle structure. The local closure of striking point was changed by reducing the radius of the semicircle step by step with all the other parameters fixed the same. The simulation results reveal that the heat flux show great differences in low density case and high density case. In low density case, the applied of semicircle structure (r~10cm) can hardly change the heat flux of the striking point. Yet, for the high density case, the semicircle structure will greatly reduce the temperature of the striking point, thus enhance the density of the neutral gas. In other words, the increased local closure effect will significantly reduce the heat flux of the striking point and help reduce the threshold density of the detachment. (author)

[en] We look into 3 questions. How does the confinement of low- or medium-Z impurities inside the divertor change with different divertor plasma conditions? How does the impurities radiation distribution look like meanwhile? How may transport affect the impurities cooling efficiency?