Molybdenum sources and transport in Alcator C-Mod

Lipschultz, B.; Pappas, Dimitri; LaBombard, B.; Rice, J. E.; Smith, D.L.; Wukitch, S.J.

doi:10.1016/s0022-3115(00)00436-0

Cited by 18 publications

(20 citation statements)

References 23 publications

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“…A similar de-correlation of Mo sources and contamination was observed in C-MOD [106]. In JET [298] and ASDEX Upgrade [299], when the carbon divertor was replaced by a beryllium and tungsten divertor respectively, the change did not affect the carbon core concentration.…”

Section: Core Contaminationsupporting

confidence: 65%

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Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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Abstract.Progress, since the ITER Physics Basis publication, in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are : energy transport in the SOL in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main chamber material elements, ELM energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low and high Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas : refinement of the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a 2 major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral-neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma-materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed. Introduction.This chapter outlines the significant progress achieved since the ITER Physics Basis in understanding basic scrape-off layer (SOL) and divertor processes in a tokamak. The interaction of plasma with first-wall surfaces will have considerable impact on the performance of fusion plasmas, the lifetime of plasma facing components, and the retention of tritium in next step Burning Plasma E...

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Section: Core Contaminationsupporting

confidence: 65%

“…3) Increased sputtering of RF antennae and protection limiters by energetic ions due to acceleration in RF rectified sheaths [106,287].…”

Section: Intrinsic Impurity Sourcesmentioning

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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“…A lower level of the perpendicular transport, a higher prompt redeposition, and larger parallel flows at the highfield side might add to the asymmetry. In Alcator C-Mod, Mo [16] and low-Z impurities (N,C) [17,18] also showed a much higher penetration at the low-field. Penetration of W injected by laser-blow off did not show much variation between low-field and high-field side in ASDEX Upgrade [19].…”

Section: Main Chamber Tungsten Sources and Fuelingmentioning

confidence: 97%

Plasma-wall interaction and plasma behaviour in the non-boronised all tungsten ASDEX Upgrade

Dux

Bobkov

Herrmann

et al. 2009

Journal of Nuclear Materials

152

198

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ASDEX Upgrade has successfully started the second experimental campaign with a full tungsten coverage of the plasma facing components and without using a boronisation for machine conditioning. The tungsten erosion at all relevant positions in the main chamber and the divertor was investigated. The outer divertor is by far the strongest source region, especially in discharges with high divertor temperature in-between ELMs. In the main chamber, the central column is usually the first limiting structure and produces then larger W erosion fluxes than the outboard limiters. Nevertheless, the tungsten influx from the outboard limiters has a much stronger effect on the tungsten content in the confined plasma. An increase of the available power from the fly-wheel generator allowed for improved H-mode operation at 1 MA, and H factors in the range of 1.2 could be achieved at acceptable W concentrations of about 2¢10 .

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“…However, in RF heated discharges the simulations underestimated the observed influx by up to a factor of three which was attributed to the production of a small fraction of energetic plasma ions or a non thermal electron distribution. An attempt has been made to gain a consistent picture of Mo influxes and resulting Mo content [85,86]. It is concluded that the Mo source at the outer strike point zone dominates the Mo content during ohmic discharges and the influx from the inner wall, although large by number ( …”

Section: Alcator C-modmentioning

confidence: 99%

High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

Neu

2006

Phys. Scr.

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In present day fusion devices optimization of the performance and experimental freedom motivates the use of low-Z plasma facing materials. However in a future fusion reactor, for economic reasons a sufficient lifetime of the first wall components is essential. Additionally, tritium retention has to be small to meet safety requirements. Tungsten appears to be the most realistic material choice for reactor plasma facing components (PFCs) because it exhibits the lowest erosion. But besides this there are a lot of criteria which has to be fulfilled simultaineously in a reactor. Results from present day devices and from laboratory experiments confirm the advantages of high-Z PFM but also point to operational restrictions, when using them as PFCs. They are associated with the central impurity concentration, which is determined by the sputtering yield, the penetration of the impurities and their transport within the confined plasma. The restrictions could exclude successful operation of a reactor, but concomitantly there exist remedies to ameliorate their impact. Obviously some price has to be paid in terms of reduced performance but lacking of materials or concepts which could substitute high-Z PFCs, emphasis has to be put on the development and optimization of reactor relevant scenarios which incorporate the experiences and measures.

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Molybdenum sources and transport in Alcator C-Mod

Cited by 18 publications

References 23 publications

Chapter 4: Power and particle control

Chapter 4: Power and particle control

Plasma-wall interaction and plasma behaviour in the non-boronised all tungsten ASDEX Upgrade

High-Z plasma facing components in fusion devices: boundary conditions and operational experiences

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