High-performance and long-pulse operation is a crucial goal of current magnetic fusion research. Here, we demonstrate a high-connement plasma regime known as an H-mode with a record pulse length of ...over 30 s in the Experimental Advanced Superconducting Tokamak sustained by lower hybrid wave current drive (LHCD) with advanced lithium wall conditioning. We nd that LHCD provides a exible boundary control for a ubiquitous edge instability in H-mode plasmas known as an edge-localized mode, which leads to a marked reduction in the heat load on the vessel wall compared with standard edge-localized modes. LHCD also induces edge plasma ergodization that broadens the heat deposition footprint. The heat transport caused by this ergodization can be actively controlled by regulating the edge plasma conditions. This potentially offers a new means for heat-ux control, which is a key issue for next-step fusion development. PUBLICATION ABSTRACT
In ITER, as in any tokamak, the first wall and divertor plasma-facing components (PFC) must provide adequate protection of in-vessel structures, sufficient heat exhaust capability and be compatible ...with the requirements of plasma purity. These functions take on new significance in ITER, which will combine long pulse, high power operation with severe restrictions on permitted core impurity concentrations and which, in addition, will produce transient energy loads on a scale unattainable in today’s devices. The current ITER PFC design has now reached a rather mature stage following the 2007 ITER Design Review. This paper presents the key elements of the design, reviews the physics drivers, essentially thermal load specifications, which have defined the concept and discusses a selection of material and design issues.
Plasma wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper, different aspects of the PWI are assessed ...in their importance for the initial wall materials choice: CFC for the strike point tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor/Be first wall and all-W or all-C. One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a
Q
=
10 ITER discharge G. Federici et al., J. Nucl. Mater. 290–293 (2001) 260 which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4
±
3 higher, this margin has been adopted as uncertainty of the scaling. With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated: It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5
MJ. Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated. For low-
Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms. For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO additional tritium trapping sites due to neutron-irradiation damage need to be taken into account. Finally, the expected values for erosion and tritium retention are compared to the ITER administrative limits for the lifetime, dust and tritium inventory.
A review of the main concepts, proposed solutions and results of R&D as well as outstanding issues on power exhaust in tokamaks is presented with specific emphasis on the expected issues that need to ...be resolved for ITER and future fusion reactors such as DEMO for both conventional and advanced divertor concepts.
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