The advantages of nuclear fusion as an energy source and research progress in this area are summarized. The current state of the art is described. Laser fusion, inertial confinement fusion, and magnetic fusion (the tokamak) are explained, the latter in some detail. Remaining problems and planned future reactors are considered. They are the Compact Ignition Tokamak (CIT), the International Thermonuclear Experimental Reactor (ITER), and a US design called TIBER II. The design of the latter is shown. >

http://ieeexplore.ieee.org/iel1/45/1370/00031587.pdf

Nuclear fusion

A blob-filament (or simply “blob”) is a magnetic-field-aligned plasma structure which is considerably denser than the surrounding background plasma and highly localized in the directions perpendicular to the equilibrium magnetic field B. In experiments and simulations, these intermittent filaments are often formed near the boundary between open and closed field lines, and seem to arise in theory from the saturation process for the dominant edge instabilities and turbulence. Blobs become charge-polarized under the action of an external force which causes unequal drifts on ions and electrons; the resulting polarization-induced E × B drift moves the blobs radially outwards across the scrape-off-layer (SOL). Since confined plasmas generally are subject to radial or outwards expansion forces (e.g., curvature and ∇B forces in toroidal plasmas), blob transport is a general phenomenon occurring in nearly all plasmas. This paper reviews the relationship between the experimental and theoretical results on blob form...

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Convective transport by intermittent blob-filaments: Comparison of theory and experiment

Fusion,四个来自上海的大男孩。2008年第一周,他们发行属于自己的第一张创作专辑《如果未来》。一点一点向梦想靠近的他们坦言:"一开始组乐队只是为了爱好,只是在玩,但是玩着玩着,发现已经不能不玩,可是如果要继续玩,我们需要更大的力气与努力。"这种勇气,我们坚信:音乐、如果,爱,就要,继续!

Fusion 如果,爱

Interactions between the plasma and the vessel walls constitute a major engineering problem for next step fusion devices, such as ITER, determining the choice of the plasma-facing materials. A prominent issue in this choice is the tritium inventory build-up in the vessel, which must be limited for safety reasons. The initial material selection, i.e. beryllium (Be) on the main vessel walls, tungsten (W) on the divertor upper baffle and dome, and carbon fibre composite around the strike points on the divertor plates, results both from the attempt to reduce the tritium inventory and to optimize the lifetime of the plasma-facing components.In the framework of the EU Task Force on Plasma?Wall Interaction (PWI TF), the many physics aspects governing the tritium inventory are brought together. Together with supporting information from international experts represented by the ITPA SOL/DIV section, this paper describes the present status of knowledge of the in-vessel tritium inventory build-up. Firstly, the main results from present fusion devices in this field are briefly reviewed. Then, the processes involved are discussed: implantation, trapping and diffusion in plasma-facing materials are considered as well as surface erosion and co-deposition of tritium with eroded material. The intermixing of the different materials and its influence on hydrogen retention and co-deposition is a major source of uncertainty on present estimates and is also addressed.Based on the previous considerations, estimates for the tritium inventory build-up are given for the initial choice of ITER materials, as well as for alternative options. Present estimates indicate a build-up of the tritium inventory to the administrative limit within a few hundred nominal full power D?:?T discharges, co-deposition with carbon being the dominant process. Therefore, tritium removal methods are also an active area of research within the EU PWI TF, and are discussed. An integrated operational scheme to slow the rate of tritium accumulation is presented, which includes plasma operation as well as conditioning procedures.

/pdf/tritium-inventory-in-iter-plasma-facing-materials-and-23mclmy3io.pdf

Tritium inventory in ITER plasma-facing materials and tritium removal procedures

A271 TRANSPORT BY INTERMITTENCY IN THE BOUNDARY OF THE DIII-D TOKAMAK. Intermittent plasma objectives (IPOs) featuring higher pressure than the surrounding plasma, are responsible for {approx} 50% of the E x B{sub T} radial transport in the scrape off layer (SOL) of the DIII-D tokamak in L- and H-mode discharges. Conditional averaging reveals that the IPOs are positively charged and feature internal poloidal electric fields of up to 4000 V/m. The IPOs move radially with E x B{sub T}/B{sup 2} velocities of {approx} 2600 m/s near the last closed flux surface (LCFS), and {approx} 330 m/s near the wall. The IPOs slow down as they shrink in radial size from 4 cm at the LCFS to 0.5 cm near the wall. The skewness (i.e. asymmetry of fluctuations from the average) of probe and beam emission spectroscopy (BES) data indicate IPO formation at or near the LCFS and the existence of positive and negative IPOs which move in opposite directions. The particle content of the IPOs at the LCFS is linearly dependent on the local density and decays over {approx} 3 cm into the SOL while their temperature decays much faster ({approx} 1 cm).

/pdf/transport-by-intermittency-in-the-boundary-of-the-diii-d-4unr1kj0ia.pdf

Transport by intermittency in the boundary of the diii-d tokamak

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ~10?20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITER's use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

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Y. Yang

Papers

Plasma-surface interaction, scrape-off layer and divertor physics:Implications for ITER

Plasma-surface interaction, scrape-off layer and divertor physics: Implications for ITER. Invited paper