Transport studies of high-<i><b>Z</b></i>elements in neon edge radiation cooled discharges in TEXTOR-94

Rapp, J.; Tokar, M. Z.; Könen, L.; Koslowski, H. R.; Bertschinger, G.; Brix, M.; Claaßen, H.A.; Jaspers, R.; Krämer‐Flecken, A.; Ohya, Kaoru; Philipps, V.; Pospieszczyk, A.; Samm, U.; Tanabe, T.; Telesca, G.; Unterberg, B.; Oost, G. Van

doi:10.1088/0741-3335/39/10/009

Cited by 55 publications

(33 citation statements)

References 26 publications

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“…Large high Z impurity accumulation is observed in conditions that are favourable for this from neoclassical transport expectations, as seen in TEXTOR [367] and FTU [368]. Central high-Z accumulation leads to strong central radiation, hollow temperature profiles and can, eventually, lead to the termination of the discharge.…”

Section: Behaviour Of High-z Plasma-facing Components In Tokamak Expementioning

confidence: 90%

Chapter 4: Power and particle control

Divertor¹,

Database²,

Editors³

1999

Nucl. Fusion

280

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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: Behaviour Of High-z Plasma-facing Components In Tokamak Expementioning

confidence: 90%

Chapter 4: Power and particle control

Divertor¹,

Database²,

Editors³

1999

Nucl. Fusion

280

View full text Add to dashboard Cite

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“…A schematic view of the most relevant processes as currently seen is presented in Fig.7. They can be described as an interplay of neoclassical effects with turbulent processes and macroscopic transport events as ELMs or sawteeth (see for example [39,40]). If the W content is high enough it can even influence the transport in the background plasma by changing the electron temperature profile through the radiation losses.…”

Section: W Transport and Contentmentioning

confidence: 99%

Overview on plasma operation with a full tungsten wall in ASDEX Upgrade

Neu¹,

Kallenbach²,

Balden³

et al. 2013

Journal of Nuclear Materials

110

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Operation with all tungsten plasma facing components has become routine in ASDEX Upgrade. The conditioning of the device is strongly simplified and short glow discharges are used only on a daily basis. The long term fuel retention was reduced by more than a factor of 5 as demonstrated in gas balance as well as in post mortem analyses. Injecting nitrogen Preprint submitted to Elsevier 23 November 2012for radiative cooling, discharges with additional heating power up to 23 MW have been achieved, providing good confinement (H98 y2 = 1), divertor power loads around 5 MW m −2 and divertor temperatures below 10 eV. ELM mitigation by pellet ELM pacemaking or magnetic perturbation coils reduces the deposited energy during ELMs, but also keeps the W density at the pedestal low. As a recipe to keep the central W concentration sufficiently low, central (wave) heating is well established and low density H-Modes could be re-established with the newly available ECRH power of up to 4 MW. The ICRH induced W sources could be strongly reduced by applying boron coatings to the poloidal guard limiters.

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“…At the same time the W concentration is strongly peaked, as can be seen from the ratio of the central to edge W concentration, which is about 20. This strong peaking -also denoted as accumulation -was already observed in early experiments using W injection in ASDEX Upgrade [22] and test-limiter experiments in TEXTOR [23]. It is usually explained by neoclassical inward drifts, which can be dominant in the absence of large turbulent transport or macroscopic instabilities as sawteeth [24].…”

Section: Ten Years Of W Programme In Asdex Upgrade -Challenges and Comentioning

confidence: 73%

Ten years of W programme in ASDEX Upgrade—challenges and conclusions

Neu¹,

Bobkov²,

Dux³

et al. 2009

Phys. Scr.

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Abstract. Since 1999 ASDEX Upgrade increased its tungsten plasma facing components and finally reached a full W coverage in 2007. Most of the initial goals of the investigations were successfully achieved. A highlight of the investigations were multiple start-ups and operation without any boronisation demonstrating that performance and confinement similar to boronised operation with carbon PFCs can be reached in high power, high density discharges. This also allowed the investigation of the hydrogen retention without disturbing effects from the low-Z coating. A strong reduction of hydrogen retention was found in gas balance measurements as well as in post mortem analyses. On the other hand, an almost complete suppression of low-Z divertor radiation was achieved after boronisation, providing valuable information on the control requirements of radiative cooling by artificially introduced impurities. Among the challenges remains the strong increase of the W source and W concentration resulting from ICRH. At the same time it helped to identify the underlying physics and may lead to solutions superior to the presently used ones.

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Transport studies of high-Zelements in neon edge radiation cooled discharges in TEXTOR-94

Cited by 55 publications

References 26 publications

Chapter 4: Power and particle control

Chapter 4: Power and particle control

Overview on plasma operation with a full tungsten wall in ASDEX Upgrade

Ten years of W programme in ASDEX Upgrade—challenges and conclusions

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