Engineering evaluation of the upgrade KSTAR divertor system

IF 1.9 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY
Sungjin Kwon , Soo-Hyeon Park , Hong-Tack Kim , Yong Bok Chang , Nak Hyong Song , Sang Woo Kwag , Hyung Ho Lee , Jong Man Lee , Hwnag Rae Cho , Do Yoon Kim , Hyeongseok Seo , Soocheol Shin , Sangmin Kim , Junyoung Jeong , Henri Greuner , Bernd Boeswirth
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引用次数: 0

Abstract

KSTAR (Korea Superconducting Tokamak Advanced Research) plans to upgrade the external heating power to 24 MW by improving heating systems such as NBI, ECH, ICRF, and so on. The upgrade divertor system should guarantee resistance to high heat flux and cooling capacity for the exhausting power in the scrape-off layer domain. The activity for the upgrade of the KSTAR divertor was initiated in 2019, and the upgrade divertor was successfully manufactured and installed in October 2023.The upgraded KSTAR divertor system employs the water-cooled tungsten monoblock, ITER-like divertor type. Tungsten is the most robust and promising plasma-facing material under high heat flux plasma circumstances, and the combination of CuCrZr heat sink and pressurized water coolant is the effective cooling method. The upgraded KSTAR divertor system has a single null configuration and 64 cassette divertor modules placed at the bottom of the vacuum vessel. A divertor module consists of the inner target, the central target, the outer target, and the cassette body, with supports to connect each part. CFD analysis was carried out in the previous study to confirm the thermal stability of a whole divertor module. The result showed the design could be operated within a thermal allowable range in 10 MW/m2 heat flux. The temperature distribution from CFD analysis is applied to the thermo-mechanical analysis. Based on the ASME code, the upgrade KSTAR divertor was estimated for plastic collapse, ratcheting, fatigue, and buckling. The result showed the upgraded KSTAR divertor is reliable from the thermal and mechanical points of view. In the KSTAR divertor, tungsten monoblocks were used only in the straight section due to space constraints. However, in anticipation of a new Korean fusion device following KSTAR, we also explored the production technology for a divertor target that includes a curved section. While the hot isostatic pressing (HIP) process was utilized for dissimilar metal bonding in the KSTAR divertor, we applied the hot radial pressing (HRP) process for manufacturing the curved section target to diversify our production methods. A small mock-up sample of the curved section, manufactured using the HIP and HRP, underwent testing for 20 MW/m² during 1,000 high-heat flux tests. Both the HIP and HRP samples were completed without any issues. The result confirms the design and quality of the KSTAR divertor target were reliable enough to withstand the heat load, although the recrystallization of tungsten occurred.
KSTAR改进型导流器系统工程评价
韩国超导托卡马克先进研究院(KSTAR)计划通过改进NBI、ECH、ICRF等加热系统,将外部加热功率提高到24兆瓦。升级的导流器系统应保证在刮擦层区域的高热流密度和排气冷却能力。KSTAR转向器的升级工作于2019年启动,升级转向器于2023年10月成功制造并安装。升级后的KSTAR转喷器系统采用水冷钨单块,类似iter型转喷器。钨是高热流密度等离子体环境下最坚固、最有前途的等离子体面材料,CuCrZr散热器与压水冷剂组合是有效的冷却方法。升级后的KSTAR转向器系统具有一个单一的零配置和64个盒式转向器模块,放置在真空容器的底部。导流器模块由内靶、中心靶、外靶和盒体组成,各部分用支架连接。在之前的研究中,我们通过CFD分析来确认整个导流器模块的热稳定性。结果表明,该设计可以在10 MW/m2热流密度的热允许范围内运行。将CFD分析得到的温度分布应用于热-力学分析。根据ASME规范,升级后的KSTAR分流器对塑性坍塌、棘轮、疲劳和屈曲进行了评估。结果表明,改进后的KSTAR转流器在热性能和力学性能上都是可靠的。在KSTAR转喷器中,由于空间限制,只在直线段使用了钨块。然而,为了期待KSTAR之后新的韩国聚变装置,我们还探索了包括弯曲截面的分流靶的生产技术。KSTAR导向器采用热等静压(HIP)工艺进行异种金属粘合,而我们采用热径向压(HRP)工艺制造弯曲截面靶材,以使生产方法多样化。使用HIP和HRP制造的弧形部分的小型模型样品在1,000次高热流密度测试中进行了20 MW/m²的测试。HIP和HRP样品都完成了,没有任何问题。结果表明,KSTAR转流靶的设计和质量是可靠的,可以承受热负荷,但钨会发生再结晶。
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来源期刊
Fusion Engineering and Design
Fusion Engineering and Design 工程技术-核科学技术
CiteScore
3.50
自引率
23.50%
发文量
275
审稿时长
3.8 months
期刊介绍: The journal accepts papers about experiments (both plasma and technology), theory, models, methods, and designs in areas relating to technology, engineering, and applied science aspects of magnetic and inertial fusion energy. Specific areas of interest include: MFE and IFE design studies for experiments and reactors; fusion nuclear technologies and materials, including blankets and shields; analysis of reactor plasmas; plasma heating, fuelling, and vacuum systems; drivers, targets, and special technologies for IFE, controls and diagnostics; fuel cycle analysis and tritium reprocessing and handling; operations and remote maintenance of reactors; safety, decommissioning, and waste management; economic and environmental analysis of components and systems.
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