Particle balance in diverted plasmas in TEXT-U

Rowan, W. L.; Bengtson, Roger D.; Bonnin, X.; Edmonds, P.H.; Hurwitz, P.; Solano, E. R.; Tsui, H.Y.W.; Uglum, J.R.; Wootton, A. J.

doi:10.1016/0022-3115(94)00562-1

“…Scrape-off layer profiles in many devices are often found to exhibit a two-zone structure: a steep gradient region in density and temperature near the separatrix (near SOL) and a flatter profile region (far SOL), characterized as a 'shoulder', extending from approximately one steep-gradient scale length outside the separatrix to the wall (e.g. ASDEX [43,44], ASDEX-Upgrade [45,46], TEXT-U [47], DIII-D [48,49], C-Mod [50,51], JT-60U [52], JET [53], TEXTOR [54], TCV [55]). The earliest hints that cross-field particle transport in the far SOL towards the main-chamber walls could be important for a tokamak reactor, competing with, or even exceeding, the parallel loss into the divertor, came from ASDEX-Upgrade [45,46].…”

Section: Particle Transport In the Solmentioning

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

Lipschultz

²

,

Kukushkin

³

et al. 2007

View full text Add to dashboard Cite

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...

show abstract

“…The gross features of SOL profiles discussed above are not new or unique to C-Mod. TEXT [248], ASDEX [249], T10 [250], DIII-D [251] and JT60-U [252,70] have reported SOL profiles with two distinct regions.…”

Section: Experimental Evidencementioning

confidence: 99%

Density limits in toroidal plasmas

Greenwald

¹

2002

Plasma Phys. Control. Fusion

View full text Add to dashboard Cite

In addition to the operational limits imposed by MHD stability on plasma current and pressure, an independent limit on plasma density is observed in confined toroidal plasmas. This review attempts to summarize recent work on the phenomenology and physics of the density limit. Perhaps the most surprising result is that all of the toroidal confinement devices considered operate in similar ranges of (suitably normalized) densities. The empirical scalings derived independently for tokamaks and reversed field pinches (RFP) are essentially identical, while stellarators appear to operate at somewhat higher densities with a different scaling. Dedicated density limit experiments have not been carried out for spheromaks and field-reversed configurations (FRC), however "optimized" discharges in these devices are also well characterized by the same empirical law. In tokamaks, where the most extensive studies have been conducted, there is strong evidence linking the limit to physics near the plasma boundary thus it is possible to extend the operational range for line-averaged density by operating with peaked density profiles. Additional particles in the plasma core apparently have no effect on density limit physics. While there is no widely accepted, first principles model for the density limit, research in this area has focussed on mechanisms which lead to strong edge cooling. Theoretical work has concentrated on the consequences of increased impurity radiation which may dominate power balance at high densities and low temperatures. These theories are not entirely satisfactory as they require assumptions about edge transport and make predictions for power and impurity scaling that may not be consistent with experimental results. A separate thread of research looks for the cause in collisionality enhanced turbulent transport. While there is experimental and theoretical support for this approach, understanding of the underlying mechanisms is only at a rudimentary stage and no predictive capability is yet available..

show abstract

“…Characterized as a 'shoulder', extending from approximately one steep-gradient scale length outside the separatrix to the wall (e.g., ASDEX [43,44], ASDEXUpgrade [45,46], TEXT-U [47], DIII-D [48,49], C-Mod [50,51], JT-60U [52], JET [53], TEXTOR [54], TCV [55]). The earliest hints that cross-field particle transport in the far SOL toward the main-chamber walls could be important for a tokamak reactor, competing with, or even exceeding, the parallel loss into the divertor, came from ASDEX-Upgrade [45,46].…”

Section: Particle Transport In the Solmentioning

confidence: 99%

Chapter 4: Power and particle control

Divertor¹,

Database²,

Editors³

1999

View full text Add to dashboard Cite

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...

show abstract

Particle balance in diverted plasmas in TEXT-U

Cited by 9 publications

References 3 publications

Chapter 4: Power and particle control

Chapter 4: Power and particle control

Density limits in toroidal plasmas

Chapter 4: Power and particle control

Contact Info

Product

Resources

About