A study of molybdenum influxes and transport in Alcator C-Mod

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

doi:10.1088/0029-5515/41/5/311

Cited by 134 publications

(205 citation statements)

References 51 publications

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“…In general, high Z core concentrations decrease with increasing density [364,365,361], as expected from the reduction of physical sputtering with edge temperature, and are lower (by factors of 2.5 to 10) in divertor than in limiter geometries [290,364]. Experience in ASDEX-Upgrade has shown that it is possible to operate the device in a large range of conditions while maintaining the core W concentration in a range of 10 -6 to 3 10 -5 both with W as divertor and main chamber material [363].…”

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

confidence: 69%

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

confidence: 69%

“…In Alcator C-Mod the molybdenum level is most clearly linked to the confinement mode (ELM-free operation leads to much higher Mo concentrations that EDA mode [364]) and the quality of the boronised layer. As the boronisation layer wears off through operation, the Mo levels increase [290].…”

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

confidence: 99%

Chapter 4: Power and particle control

Loarte¹,

et al. 2007

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“…However, after having coated all poloidal low field side limiters with W it became evident that the beneficial influence on central particle transport from central ICRH could be overridden by the increased W influx leading to an equally radially enhanced W density. A similar be-haviour has been reported from the all-molybdenum device Alcator C-Mod when operating with Mo ICRH limiters [22,23]. Therefore, special emphasis was put on the investigation of the W sources, especially during ICRH [24,25].…”

Section: Tungsten Influxesmentioning

confidence: 81%

Final steps to an all tungsten divertor tokamak

Neu

Bobkov

Dux

et al. 2007

Journal of Nuclear Materials

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Currently 85% of the plasma facing components of ASDEX Upgrade are tungsten coated.Carbon influx from W PFCs is still observed but a reduction of the C content is found in plasma discharges and a lower C fraction is measured in deposited layers in agreement with modelling. W sputtering from the low field side guard and ICRF limiters is mainly due to fast particles from NBI as well as from ions accelerated in the rectified sheath during ICRF operation. The increase of the W source area is reflected in increased W concentrations.For medium to high density discharges the techniques developed so far, namely central heating and ELM pace-making, allow keeping the W concentration in the range of 10Boronisation strongly reduces the W influxes and similarly the W content especially during ICRF operation, but this reduction is only temporary and equilibrium is reached already after about 100 discharges.

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“…Typically, the integrated fluxes from the LFS and HFS are similar in size for similar distances, but it is more beneficial to have the LFS gap 1-2 centimetres larger than the HFS gap to reach the lowest W content [34]. Obviously, the low field side penetration of W is higher as also seen in Alcator C-Mod for molybdenum [36]. Comparing the absolute value of main chamber and divertor sources the time averaged divertor source is at least one order of magnitude larger in type I ELMy H-mode discharges without using ICRH [34,10].…”

Section: W Sourcesmentioning

confidence: 99%

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

Neu¹,

Kallenbach²,

Balden³

et al. 2013

Journal of Nuclear Materials

112

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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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A study of molybdenum influxes and transport in Alcator C-Mod

Cited by 134 publications

References 51 publications

Chapter 4: Power and particle control

Chapter 4: Power and particle control

Final steps to an all tungsten divertor tokamak

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

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