Conceptual design and heat transfer performance of a flat-tile water-cooled divertor target

Li, Lei; Han, Lijun; Zi, Pengfei; Cao, Lei; Xu, Tianheng; Mou, Nanyu; Wang, Zhaoliang; Yin, Lei; Yao, Damao

doi:10.1088/2058-6272/ac0689

Cited by 15 publications

(3 citation statements)

References 31 publications

(35 reference statements)

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“…The incident heat flux is about 16.5±0.8 MW m −2 (according to formula 1) and the absorbed heat flux is about 10.1±0.5 MW m −2 (according to formula 2). Thus, the heat absorption ratio is about 0.61, which is comparable to the results reported by others [20,22]. After the test, the interface of OFC/CuCrZr is examined by ultrasonic testing (UT), and the surface morphology is observed using a digital microscope.…”

Section: Experimental Procedures Of Steady-state Hhf Testsupporting

confidence: 86%

Microchannel cooling technique for dissipating high heat flux on W/Cu flat-type mock-up for EAST divertor

Han

Zhao

et al. 2022

Plasma Sci. Technol.

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As an important component of the tokamak, divertor is mainly responsible for extracting heat and helium ash, and the targets of the divertor need to withstand high heat flux of 10 MW/m2 for steady-state operation. In this work, we proposed a new strategy, using microchannel cooling technology to remove high heat load on the targets of the divertor. The results demonstrated that the microchannel-based W/Cu flat-type mock-up survived successfully the thermal fatigue test of 1000 cycles at 10 MW/m2 with the cooling water of 26 L/min, 30 °C (inlet), 0.8 MPa (inlet), 15 s power on and 15 s dwell time; the maximum temperature on the heat loaded surface (W surface) of the mock-up was 493 °C, which is much lower than the recrystallization temperature of W (1200 °C). Moreover, no occurrence of macrocrack and ‘hot spot’ at the W surface, as well as no detachment of W/Cu tiles were observed during the thermal fatigue testing. These results indicate that microchannel cooling technology is an efficient method for removing the heat load of the divertor at a low flow rate. The present study offers a promising solution to replace the monoblock design for the EAST divertor.

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Section: Experimental Procedures Of Steady-state Hhf Testsupporting

confidence: 86%

Microchannel cooling technique for dissipating high heat flux on W/Cu flat-type mock-up for EAST divertor

Han

Zhao

et al. 2022

Plasma Sci. Technol.

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“…The other type of CM covering the rest area employs W/Cu flat-type PFCs for each target, as shown in figure 2(c). W/Cu flat-type PFCs based on new explosive welding and brazing technologies are expected to withstand high heat loads of up to 20 MW m −2 on the top surface [39], and were thus installed in EAST to test the reliability and capacity of the new structure and technologies. For a typical W/Cu flat-type PFC, several pure W plates with 2 mm thickness were brazed on the copper-stainless steel composite plate joined by the explosive welding method.…”

Section: New W/cu Pfcs For the Lower Divertor Target In Eastmentioning

confidence: 99%

In situ melting phenomena on W plasma-facing components for lower divertor during long-pulse plasma operations in EAST

et al. 2023

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Tungsten (W) is one of the most promising plasma-facing materials (PFMs) for future fusion device. Although its melting point is the highest among all the metals, it still has great risk of melting under extremely high plasma heat fluxes, which is a big concern for the ITER and future reactors. Actively-cooled W plasma-facing components (PFCs) with both monoblocks and flat-type structure have been successfully installed in the lower divertor of the EAST tokamak since 2021, which provide a good opportunity for direct comparison of damage mechanism for the two types of PFCs. Various in-situ melting phenomena on the lower divertor have been observed by CCD cameras, which are further verified by Post-mortem inspections. Severe melting even exfoliation of the edge beveled W plates on some W/Cu flat-type components at horizontal outer targets were observed. A large number of droplets were ejected during long-pulse operations, which induced the significant increase of W impurity and total irradiation in the core plasma, and thus greatly deteriorated the plasma performance and even cause disruptions. Two different shaping structures of flat-type PFCs show different position of melting and corresponding mechanisms. Few slight melting has been found on the sharp leading edges of W/Cu monoblocks at the inter-cassette modules for horizontal targets with small droplets ejection, which is much improved compared to that occurred on the upper W divertor, illustrating that the application of large-sized bevel chamfer at inter-cassette modules was generally effective. In addition, an unexpected melting phenomenon on the dome plate was attributed to the extreme transient heat flux during disruption with runaway electrons. The application of both types of W/Cu PFCs for divertor provides important experiences and lessons for engineering design and optimization of divertor PFCs in future fusion devices.

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“…Until April 2021, the EAST lower divertor has completed the upgrade from graphite divertor to tungsten divertor. The lower divertor has totally 48 modules, of which three-quarters use the ITER-like mono-block structure as the target plate, and the other quarter use the flat-type structure [72] as the target plate as comparison of different technologies, shown in figure 18. The installation accuracy of all divertor modules reaches ±0.5 mm, and all target plates have large chamfers to avoid leading-edge.…”

Section: Summary and Future Plansmentioning

confidence: 99%

Advances in the long-pulse steady-state high beta H-mode scenario with active controls of divertor heat and particle fluxes in EAST

et al. 2022

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Since the last IAEA-Fusion Energy Conference, the Experimental Advanced Superconducting Tokamak (EAST) research program has been, in support of ITER and CFETR, focused on development in terms of the long-pulse steady-state (fully noninductive) high beta H-mode scenario with active controls of the stationary and transient divertor heat and particle fluxes. The operational domain of the steady-state H-mode plasma scenario has been significantly extended with ITER-like tungsten mono-block divertor, plasma control and heating schemes. EAST has achieved several important milestones in the development of high β p H-mode scenario and its key physics and technologies. A 60 s-scale long-pulse steady-state high β p H-mode discharge with the major normalized plasma parameters similar to the designed performance of the CFETR 1 GW fusion power operation scenario has been successfully established and sustained by pure RF heating and current drive. Several feedback control schemes have been developed for a sustained detachment with good core confinement. This includes control of the total radiation power, target electron temperature, and particle flux measured using divertor Langmuir probes or a combination of the control of target electron temperature and AXUV radiation near the X point. The detachment feedback control schemes have been integrated with small-ELM regimes and high β p scenario via neon seeding, enabling a core and edge compatible integrated high-beta scenario applicable to long-pulse operations. ELM suppression has been achieved using various methods, including resonant magnetic perturbations and impurity seeding. Full suppression of ELMs by using n = 4 RMPs has been demonstrated for ITER for the first time in low input torque plasmas in EAST. EAST has been operated with helium to support the ITER research requirements for the first time. For a long-pulse, high bootstrap current fraction operation, a new lower tungsten divertor with active water-cooling has been installed, along with improvements in the heating and current drive capability.

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Conceptual design and heat transfer performance of a flat-tile water-cooled divertor target

Cited by 15 publications

References 31 publications

Microchannel cooling technique for dissipating high heat flux on W/Cu flat-type mock-up for EAST divertor

Microchannel cooling technique for dissipating high heat flux on W/Cu flat-type mock-up for EAST divertor

In situ melting phenomena on W plasma-facing components for lower divertor during long-pulse plasma operations in EAST

Advances in the long-pulse steady-state high beta H-mode scenario with active controls of divertor heat and particle fluxes in EAST

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