Design, R&amp;D and commissioning of EAST tungsten divertor

Yao, Damao; Luo, Guang–Nan; Zhou, Zhiping; Cao, Lei; Li, Q; Wang, W J; Li, L; Qin, Sigui; Shi, Yingli; Liu, G H; Li, J G

doi:10.1088/0031-8949/2015/t167/014003

Cited by 33 publications

(17 citation statements)

References 9 publications

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“…A potential solution to overcome the relatively large erosion and succeeding redeposition of carbon is the use of tungsten as plasma-facing material [16]: at detached plasma conditions and low plasma impurity contents, the erosion of tungsten by hydrogen ions is extremely small, and tungsten is successfully used in a number of tokamaks worldwide, such as ASDEX Upgrade [17], JET [18], WEST [19] and EAST [20]. However, the use of high-Z materials in stellarators is more challenging due to the higher contribution of neoclassical transport in stellarators compared to tokamaks.…”

Section: Introductionmentioning

confidence: 99%

Erosion of tungsten marker layers in W7-X

et al. 2021

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In order to get first insight into net tungsten erosion in W7-X, tungsten (W) marker layers were exposed during the operational phase OP 1.2b at one position of the Test Divertor Unit (TDU), at 21 different positions of the inner heat shield, and at two scraper elements. The maximum tungsten erosion rate at the TDU strike line was 0.13 nm s−1 averaged over the whole campaign. The erosion was inhomogeneous on a microscopic scale, with higher erosion on ridges of the rough surface inclined towards the plasma and deposition of hydrocarbon layers in the recessed areas of the rough surface. The W erosion at the inner heat shield was below the detection limit of 3–6 × 1012 W-atoms/cm2s, and all inner heat shield tiles were covered with a thin B/C/O layer with thickness in the range 2 × 1017–1018 B + C atoms/cm2 (about 20–100 nm B + C). W-erosion of the marker layers on the scraper elements was also below the detection limit.

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Section: Introductionmentioning

confidence: 99%

Erosion of tungsten marker layers in W7-X

et al. 2021

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“…[17] There are flexible B t direction and different divertor configurations (double null (DN), upper single null (USN) and lower single null (LSN)) on EAST, [18] and equipped with the upper ITER-like tungsten divertor since 2014. [19] The key diagnostic systems in the neon seeding experiments are shown in Fig. 1.…”

Section: Methodsmentioning

confidence: 99%

Evolution of the high-field-side radiation belts during the neon seeding plasma discharge in EAST tokamak

Xu¹,

Wang²,

Xu³

et al. 2022

Chinese Phys. B

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Divertor detachment achieved by injecting impurities or increasing density is always accompanied with various local radiation phenomena in the boundary or core plasma. This paper presents the formation and evolution of the high-field-side (HFS) radiation belts during the neon seeding plasma discharge in upper single null configuration with two directions of toroidal magnetic field in EAST tokamak. The neon mixed with deuterium seeding can induce the divertor detachment with strong radiation belts in the HFS scrape-off layer (SOL) region. With the increase of the radiation power, the plasma discharge will transit from H-mode to L-mode, meanwhile the radiation belts move away from the near X-point to HFS SOL. When the radiation power is high enough, the radiation belts begin to move further to the other X-point along the HFS SOL, and even cause plasma disruption. The results indicate that the behavior of the radiation belts is related to the radiation power, plasma confinement performance and state of divertor detachment, which is useful for developing better feedback control methods to achieve high-performance radiative divertor operation mode.

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“…ASIPP developed a batch manufacturing technology for ITER-like water-cooled full W-PFC which employs copper alloy (CuCrZr) as a heat sink material and the hot isostatic pressing (HIP) method to join the W-PFM and the CuCrZr heat sink together [6]. Therewith ASIPP upgraded the EAST upper divertor from the bolted graphite tile PFC [7] into the ITER-like watercooled full W-PFC in 2014, and the W divertor came into service successfully [8][9][10][11]. Fully non-inductive steadystate H-mode plasmas (H 98, y2 ~ 1.1) were achieved with a duration of more than 60 s in 2016 experimental campaign [10] and further extended to ~ 100 s in 2017 campaign with a good control of impurity and heat exhaust benefiting from the upper W divertor [11].…”

Section: Application Of Tungsten As Plasma-facing Materials In Eastmentioning

confidence: 99%

Plasma–tungsten interactions in experimental advanced superconducting tokamak (EAST)

et al. 2019

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Tungsten (W) is used as the armor material of the International Thermonuclear Experimental Reactor (ITER) divertor and is regarded as the potential first wall material of future fusion reactors. One of the key challenges for the successful application of W in fusion devices is effective control of W at an extremely low concentration in plasma. Understanding and control of W erosion are not only a prerequisite for W impurity control, but also vital concerns to plasma-facing component (PFC) lifetime. Since the application of ITER-like water-cooled full W divertor in EAST in 2014, great efforts were made to investigate W erosion by experiment and simulation. A spectroscopic system was developed to provide a real-time measurement of W sputtering source. Both experiment and simulation results indicate that carbon (C) is the dominant impurity causing W sputtering in L-mode plasmas, which comes from the erosion of C plasma-facing material (PFM) in the lower divertor and the main chamber limiters. The mixture layer on the surface of W PFCs formed through redeposition or the wall coating can effectively suppress W erosion. Increasing the plasma density and radiation can reduce incident ion energy, thus alleviating W sputtering. In H-mode plasmas, control of edge localized mode (ELM) via resonant magnetic perturbation (RMP) proves to be capable of suppressing intra-ELM W erosion. The experiences and lessons from the EAST W divertor are beneficial to the design, manufacturing and operation of ITER and beyond.

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Design, R&D and commissioning of EAST tungsten divertor

Cited by 33 publications

References 9 publications

Erosion of tungsten marker layers in W7-X

Erosion of tungsten marker layers in W7-X

Evolution of the high-field-side radiation belts during the neon seeding plasma discharge in EAST tokamak

Plasma–tungsten interactions in experimental advanced superconducting tokamak (EAST)

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