Research Progress on Heat Load Damage Behavior of Tungsten-Based Materials for Divertor

doi:10.11900/0412.1961.2024.00309

Acta Metall Sin

2025, Vol. 61

Issue (7): 961-978 DOI: 10.11900/0412.1961.2024.00309

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Research Progress on Heat Load Damage Behavior of Tungsten-Based Materials for Divertor

LUO Laima¹^,²^,³, CHEN Yu¹, YAO Gang⁴, ZHU Xiaoyong¹^,²^,³(

), ZHU Dahuan⁵, WU Yucheng¹^,²^,³

1 School of Materials Science and Engineering, Hefei University of Technology, Hefei 230009, China
2 Engineering Research Center of High-Performance Copper Alloy Materials and Processing, Ministry of Education, Hefei University of Technology, Hefei 230009, China
3 National-Local Joint Engineering Research Centre of Nonferrous Metals and Processing Technology, Hefei University of Technology, Hefei 230009, China
4 School of Materials Science and Engineering, Inner Mongolia University of Science and Technology, Baotou 014010, China
5 Institute of Plasma Physics, Hefei Institutes of Physical Science, Chinese Academy of Sciences, Hefei 230031, China

Cite this article:

LUO Laima, CHEN Yu, YAO Gang, ZHU Xiaoyong, ZHU Dahuan, WU Yucheng. Research Progress on Heat Load Damage Behavior of Tungsten-Based Materials for Divertor. Acta Metall Sin, 2025, 61(7): 961-978.

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Abstract

Limited energy resources cannot meet the long-term developmental needs of human society. As such, nuclear fusion energy is considered a key solution for environmental protection and meeting future energy demands. However, to ensure the reliable operation of fusion reactors, addressing heat load damage to the divertor facing plasma in tokamak devices is crucial. The divertor, an indispensable core component of fusion devices, plays essential roles in these devices, including the removal of heat load generated via scraping layers and radiation and protection of the main vacuum chamber, auxiliary heating systems, and diagnostic systems, thereby ensuring the safe and stable operation of nuclear fusion reactors. Nevertheless, due to harsh operational conditions, the divertor is prone to damage, limiting the stable operation of long-pulse, high-parameter plasmas. W is critical in the divertor of fusion reactors, primarily owing to its high melting point, low physical sputtering rate, low deuterium retention, and excellent mechanical properties, allowing it to perform stably under extreme conditions. However, tungsten materials have several limitations, including a high ductile-brittle transition temperature, a low recrystallization temperature, and susceptibility to activation. Therefore, it is necessary to regulate, modify, and optimize these materials to enhance the performance of plasma-facing materials (PFMs). Such improvements aim to increase their resilience under extreme environments, minimize damage risks under high heat loads, and enhance heat load resistance, thereby ensuring the long-term stable operation of the divertor in fusion reactors and to meet future energy challenges. The working conditions of fusion reactors are extremely harsh, with the divertor region experiencing continuous heat load damage. It typically faces steady-state heat loads with peak values as high as 5-20 MW/m² and transient heat loads of up to ~2 GW/m². These heat loads can cause melting and cracking on both sides of the divertor cassette, posing a risk of reactor failure. Consequently, the study of the heat load damage behavior in tungsten-based PFMs as well as development of damage mitigation strategies have become hot topics in fusion research. This paper reviews current research efforts, both domestic and international, related to the damage behavior of pure tungsten, tungsten alloys, and dispersed phase-strengthened tungsten under heat load conditions. Additionally, it summarizes and forecasts the evolution of heat load damage in tungsten-based materials and presents strategies for damage mitigation, thereby providing a reference for future research endeavors.

Key words: divertor tungsten-based material heat load damage behavior strategies to inhibit the damage

Received: 03 September 2024

ZTFLH:

TG146.1

Fund: National Key Research and Development Program of China(2019YFE03120002);National Key Research and Development Program of China(2022YFE03140001);National Key Research and Development Program of China(2022YFE03030003);Fundamental Research Funds for the Central Universities(JZ2023HGQB0164);Postdoctoral Fellowship Program of CPSF(GZC20230656);Natural Science Foundation of Anhui Province(2108085J21);Natural Science Foundation of Anhui Province(2308085QE154);Key Research and Development Program of Anhui Province(202104A05020045)

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	WU Yucheng

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https://www.ams.org.cn/EN/10.11900/0412.1961.2024.00309 OR https://www.ams.org.cn/EN/Y2025/V61/I7/961

Fig.1 Schematics of magnetic confinement nuclear fusion Tokamak international thermonuclear experimental reactor (ITER)^[2] (a) and divertor structure design^[3] (b); and synergistic multi-field coupling loading scenarios of plasma-facing materials (PFMs) in future thermonuclear fusion reactors^[13] (DEMO—demonstration plant, ELM—edge localized mode, VED—vertical displacement) (c) and sketch maps showing the general mechanism of crack formation in heat load tests^[26] (d1, d2)

Fig.2 Surface SEM images of cracked and melted zones in W after 3 GW/m² heat loads^[27]
(a) 10⁴ pulse of 1 ms at 20 Hz (b1, b2) high magnified SEM images of Fig.2a (c, d) 1 pulse of 13 ms (c) and 14 ms (d), respectively

Fig.3 SEM images showing typical distribution of primary and secondary cracks (a) and cracking on the pure W surface under different heat loads with 1 pulse (b-d)^[34]
(b) 0.5 GW/m² (c) 0.85 GW/m² (d) 1.25 GW/m²

Fig.4 Low (a, b) and high (a1, b1, b2) magnified surface morphologies of W monoblock in small-scale W monoblock divertor mock-up (SSMU) after high heat loading test (HHFT)^[37] (ALMT—allied material, NSCM—nidec sankyo cmi; red arrows represent macro-crack; green, red, and blue rectangles are the high magnified images of the heat-loaded surface)
(a, a1) ALMT-W (b, b1, b2) NSCM-W

Fig.5 Surface SEM images showing recrystallized W (W-Recry.) (a, b) and W prepared by selective electron beam melting (SEBM) with the best parameter set (c, d) before (a, c) and after (b, d) high heat loading test^[39]

Table 1 Statistical analyses of the heat load resistance of pure ultra-fine grained W^[51,52]

Fig.6 Low and high (insets) magnifed SEM images of surface of pure W (a, c) and W-3%Ta alloy (b, d) after thermal shock tests with background steady state D-plasma exposure under 700 ^oC base temperature and 0.38 GW/m² laser power density^[58] (RD—rolling direction, TD—transverse direction)
(a, b) 10³ pulse (c, d) 10⁵ pulse

Fig.7 Low and high (insets) magnified SEM images of the loaded surfaces under different high heat flux (HHF) densities^[61]
(a-c) W-1V alloy (d-f) W-5V alloy (g-i) W-10V alloy

Fig.8 SEM images of pure W (a, a1, c, c1) and W-K alloy (b, b1, d, d1) after heat loading tests without (a, a1, b, b1) and with (c, c1, d, d1) irradiation^[68]

Table 2 Comparisons of heat load resistances between W-K-xY alloys and pure W on a single thermal shock test^[71-73]

Fig.9 Low (a-c) and high (a1-c1) magnified surface FESEM images of pure W (a, a1), W-1.0%TaC alloy (b, b1), and W-1.0%TiC alloy (c, c1) after heat loading test^[87] (Inset in Fig.9c1 is the corresponding high magnified image)

Fig.10 Low (a-d) and high (a1-d1) magnified SEM images showing surface morphologies of the W-2.0%Y₂O₃ composite material after 100 pulses thermal shock experiment with different power densities^[88] (ND—normal direction)
(a, a1, b, b1) TD-ND surfaces (c, c1, d, d1) RD-TD surfaces

Fig.11 SEM images showing surface (a, b, d-f) and cross-section (c) morphologies of the samples after heat loading tests
(a-c) W-0.3Y₂O₃ (a) and W-1.1Y₂O₃ (b, c) at 0.6 GW/m²^[92] (d-f) W-0.5Y₂O₃-1Mo (d, e) and W-0.5Y₂O₃-1Ti (f) after high heat load test applying a single shot at 0.9 GW/m²^[102]

Table 3 Comparisons of pure W and tungsten-based materials after transient heat load tests^{[83,85,88,90,92,97,102,103,106-112]}

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