Damage and melting of ITER-like flat-type tungsten castellated blocks exposed to long pulse H-mode plasmas

Hong, Suk–Ho; Kim, Kyungmin; Kim, Hong-Tack; Bang, Eunnam; Choi, Heekyung; Kim, Hyung-Chan; Pitts, R.A.

doi:10.1016/j.fusengdes.2018.05.046

Cited by 7 publications

(5 citation statements)

References 16 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The blocks were exposed to various plasmas during a campaign and observed by a high-resolution infrared (IR) camera (FLIR ® , resolution: 640 × 320, 125 Hz or 2.7 kHz acquisition speed) [21]. Figure 8 (right) shows images of heat load patterns on the leading edge of a height of 0.3 mm under the edge localized mode (ELM) and inter-ELM phase (#16279), obtained by the divertor IR camera with 2.7 kHz acquisition speed.…”

Section: Damage and Melting Test Of Tungsten Blocks By Power Loading mentioning

confidence: 99%

“…9 Visible camera image of the 2 mm leading edge after 2016 KSTAR campaign Reproduced with permission from Ref. [21] Copyright 2018 Elsevier Fig. 10 Top and side views of W castellated blocks exposed to long-pulse H-mode plasmas in a 2016 and b 2017 KSTAR campaign a Reproduced with permission from Ref.…”

Section: Damage and Melting Test Of Tungsten Blocks By Power Loading mentioning

confidence: 99%

Section: Damage and Melting Test Of Tungsten Blocks By Power Loading mentioning

confidence: 99%

mentioning

confidence: 99%

See 3 more Smart Citations

A brief summary of tungsten technology development in Korea

Hong

2020

Tungsten

Self Cite

View full text Add to dashboard Cite

A review of tungsten technology development in Korea is briefly given. According to the upgrade plan of Korea superconducting tokamak advanced research (KSTAR) tokamak associated with Korean fusion energy R&D program, graphite plasma-facing components will be replaced with tungsten-based ones to handle the high-peak heat load caused by an increase of heating power up to 26 MW. Brazing technique to bond tungsten was developed and tungsten blocks were manufactured. Blocks were installed at the central divertor in KSTAR and exposed to high heat flux. Under high heat flux and long-pulse discharge, tungsten blocks were severely damaged. Molten tungsten materials show movements towards the high field side, which is j×B direction. The COMSOL ® modeling described the melting event quantitatively well. The failure of a water cooling system with a metal wall environment during a long-pulse plasma operation is very critical.

show abstract

Section: Damage and Melting Test Of Tungsten Blocks By Power Loading mentioning

confidence: 99%

Section: Damage and Melting Test Of Tungsten Blocks By Power Loading mentioning

confidence: 99%

Section: Damage and Melting Test Of Tungsten Blocks By Power Loading mentioning

confidence: 99%

mentioning

confidence: 99%

See 2 more Smart Citations

A brief summary of tungsten technology development in Korea

Hong

2020

Tungsten

Self Cite

View full text Add to dashboard Cite

show abstract

“…Therefore, leading-edge-induced thermal issues are one of the major concerns for divertor operation of ITER and future fusion reactors [7]. In order to investigate the leading-edge-induced thermal effect as well as the possible damage, various experiments have been performed on castellated samples by setting dedicated leading edges to be exposed to the plasma heat fluxes in different tokamaks such as JET [8,9], ASDEX-Upgrade [10,11], KSTAR [12], DIII-D [13]. In JET experiments, the melting of protruded sharp leading edges was induced by transient heat flux during edge localized modes (ELMs), and the heat flux deposition profile on various sized leading edges was also investigated and discussed.…”

Section: Introductionmentioning

confidence: 99%

In situ leading-edge-induced damages of melting and cracking W/Cu monoblocks as divertor target during long-term operations in EAST

Zhu¹,

Li²,

Gao³

et al. 2022

Nucl. Fusion

View full text Add to dashboard Cite

Leading-edge-induced thermal loading effect due to assembly tolerance between neighboring castellated plasma facing components is a critical issue in fusion devices. Actively-cooled ITER-like W/Cu monoblocks were successfully installed for upper divertor target in EAST which significantly increases the ability of divertor power exhaust. The misalignment between neighboring monoblocks was formed inevitably during manufacturing and assembly processes, providing a possibility to demonstrate the leading-edge-induced thermal damages. Indeed, the leading-edge-induced melting phenomena of W/Cu monoblocks on upper divertor targets were observed during plasma discharges with a large number of droplets ejected from divertor target using CCD camera, which were also identified at the leading edges of W/Cu monoblocks. Not only that, but also many macro cracks with width of ~70 m and depth of < 5 mm along radial and toroidal directions were also found universally at the leading edges of W/Cu monoblocks by post-mortem inspection after plasma campaigns. Thermal-mechanical analysis by means of the finite element simulation demonstrated that the maximum temperature could reach W melting point under current projected heat load of ~3 MWm-2 on flat top surface with large misalignment up to 3 mm at the leading edges. Meanwhile, the high temperature also induced high thermal stress and strain concentration at the center of leading edges, at which the thermal fatigue cracking could be initially generated. Such kind of cracks at leading edges on W/Cu monoblocks may be unavoidable due to the long-term pulsed fatigue effects. However, the influence of these cracks seems to be acceptable thanks to the limited propagated distance by self-castellation effect, which still need long-term tracking. The in-situ leading-edge-induced damages of melting and cracking on W/Cu monoblocks of EAST upper divertor target provide significant reference to understand the leading-edge-induced thermal effect in ITER and future fusion devices.

show abstract