Plasma-wall interactions fusion devices

The control of impurity release and transport requires a better understanding of the complex phenomena of plasma-wall interactions including the processes occuring in the scrape-off layer in the limiter shadow. In order to establish the feasibility of suggested solutions such as divertors or surface modifications, experiments have to be performed not only in the laboratory but also in-situ in fusion devices. The latter... 

A reliable determination of hydrocarbon fluxes is a vital task since the introduction of carbon as a PFC in fusion plasma devices. However, what looked promising after the work of [16,28] has turned out to be a real nightmare in plasma-wall interaction research. A taste of that is seen in Fig. 6.7 where the dependence of chemical erosion yields evaluated in different devices on ion flux density is plotted using the published D/XB-values for CD. Even stronger discrepancies from the general trend have been reported in [29,30]. 

Simulating erosion and re-deposition processes in fusion devices lead to a better understanding of the processes involved. The 3-dimensional Monte-Carlo code ERO-TEXTOR [35,36] has been developed to model the plasma-wall interaction and the transport of eroded particles in the vicinity of test limiters exposed to the edge plasma of TEXTOR. Important problems concerning the lifetime of various wall materials (high Z vs. low Z) under different plasma conditions and the transport of eroded impurities into the main plasma can be treated with the ERO-TEXTOR code. Recently, the divertor geometries have been implemented to carry out simulations for JET, ASDEX and ITER [37], In addition, first attempts have been made to simulate erosion and re-deposition processes in the linear plasma device PISCES to analyze the effect of beryllium. 

Plasma-Wall Interaction in Nuclear Fusion Devices. 2775... 

This section deals with plasma-wall-interaction (PWI) processes in magnetically confined fusion (MCF) devices only. In inertial confinement fusion (IGF), the interaction is... 

The requirements for long pulse operation in the next step fusion device ITER and beyond, like acceptable power exhaust, peak load for steady state, transient loads, sufficient target lifetime, limited long term tritium retention in wall surfaces, acceptable impurity contamination in central plasma and efficient helium exhaust, depend on complex processes. The input to the numerical codes, which are used for the optimization of divertor and wall components, relies to a large extend on our understanding of the major processes related to erosion and deposition, tritium retention, impurity sources and erosion processes. The reliability of predictions made with these codes depends crucially on the accuracy of the atomic and plasma-material interaction data available. 

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