Analyses of erosion and re-deposition layers on graphite tiles used in the W-shaped divertor region of JT-60U

Gotoh, Y.; Yagyu, J.; Masaki, K.; Kizu, K.; Kaminaga, A.; Kodama, K.; Arai, Takashi; Tanabe, T.; Miya, N.

doi:10.1016/s0022-3115(02)01354-5

Cited by 51 publications

(82 citation statements)

References 6 publications

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“…deposition in the inner divertor area is exceeding erosion mainly in the outer divertor area, is lacking, as clearly seen in the erosion/deposition profiles of JT-60U divertor region given in Fig. 2 [24,25], for example. Hence the first wall of the main chamber is assigned to be erosion dominated, though it is not confirmed yet because of poor diagnostics to monitor the main chamber.…”

Section: Physical and Chemical Sputteringmentioning

confidence: 86%

“…Since sputtered carbon neutrals are immediately ionized and gyrating, most of them return to the sputtered surface with a little distance (prompt redeposition). Owing to the magnetic filed line in the normal filed geometry, the sputtered atom are promptly redeposited at the poloidally inner side (or vertically upper side), as clearly appeared in the deposition profiles of JT-60U [25,31] (see Fig. 2).…”

Section: Erosion and Deposition In Tokamaks Including Geometry Effectmentioning

confidence: 90%

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On the possibility of ITER starting with full carbon

Tanabe

2006

Fusion Engineering and Design

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Section: Physical and Chemical Sputteringmentioning

confidence: 86%

Section: Erosion and Deposition In Tokamaks Including Geometry Effectmentioning

confidence: 90%

On the possibility of ITER starting with full carbon

Tanabe

2006

Fusion Engineering and Design

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“…In JT-60U, carbon layers build up at the inner divertor [311], as in JET and ASDEX Upgrade and have reached thicknesses up to 100 µm after the operation period from 1997-2000 in which ~ 8500 shots were performed, leading to a total carbon deposition of 320 g at the divertor, similar to the JET values [425]. The deposition on shadowed areas (mainly on areas in the private flux region viewing the vertical tiles) is much lower than on the target and only ~ 9g C were deposited in these areas during this operation period.…”

Section: Database On Fuel Retention In Present Fusion Devicesmentioning

confidence: 99%

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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“…for carbon materials used for divertor target tiles with the density of 1.7 g/cm 3 . In the inner divertor region, thick re-deposition layer up to 200 μm was observed [11,15]. Considering that the density of the re-deposition layer was 0.91 g/cm 3 [16], the depth of 14 × 10 23 carbon atoms/m 2 corresponds to a linear depth of 30.7 μm for re-deposition layers on the ID sample.…”

Section: Nuclear Reaction Analysis (Nra)mentioning

confidence: 93%

Deuterium depth profiling in graphite tiles not exposed to hydrogen discharges before air ventilation of JT-60U

Hayashi

Sugiyama

Mayer

et al. 2009

Journal of Nuclear Materials

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Depth profiles of deuterium trapped in graphite tiles not exposed to hydrogen discharges just before air ventilation of JT-60U were determined based on the D( 3 He,p) found at the surface of the inner divertor area, which was significantly higher than that exposed to H discharges at the same location. However in the deeper region, the deuterium concentrations without H discharges were lower than those with H discharges in 3 of 4 samples. These results indicate that hydrogen discharges can remove deuterium trapped only in the shallow region.

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Analyses of erosion and re-deposition layers on graphite tiles used in the W-shaped divertor region of JT-60U

Cited by 51 publications

References 6 publications

On the possibility of ITER starting with full carbon

On the possibility of ITER starting with full carbon

Chapter 4: Power and particle control

Deuterium depth profiling in graphite tiles not exposed to hydrogen discharges before air ventilation of JT-60U

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