Thermal and structural analysis of W-7X first wall graphite tiles under direct NBI loads

IF 1.9 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY
Vojtěch Smolík , Mikhail Khokhlov , Axel Lorenz , Samuel Lazerson , Victor Bykov , Paul McNeely , W7-X team
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Abstract

Thermo-mechanical analysis of the Wendelstein 7-X neutral beam dump is presented focusing on failure phenomena resulting from strong impulsive thermal loading of the graphite tiles. One of the plasma heating systems of the Wendelstein 7-X stellarator is Neutral Beam Injection (NBI). There are two NBI boxes with currently a total injected hydrogen beam power of 7 MW and a pulse length up to 5 s. Each NBI box can be operated with four positive ion neutral injectors (PINI). A total of three PINIs were used simultaneously in the last operation campaign in 2022. NBI is highly important for the W7-X physics program, thus increasing NBI power is planned in the next operation phase. A portion of the NBI power passes through the plasma, creating a region of the first wall (FW) with a significantly increased heat load. This region, called the beam dump, could be a limiting factor for the NBI pulse length and thus it requires a precise remodeling of all constituents before the 2024/2025 operation phases to prevent repeated structural damage to this area. Since the graphite tiles are attached by multiple bolts to the CuCrZr heat sinks, the thermal expansion of graphite can induce significant stress in the graphite. The safety limiting factor is the induced thermal stress in the graphite tile due to the NBI beam power load. ANSYS Mechanical model was developed, including the bolted connections, and analyzed to examine the effect of the heat load on the FW components and to evaluate the maximal allowed duration of the NBI beam. A coupled transient simulation of the thermal and structural analysis is performed. The calculation used the NBI heat loads on the FW surfaces estimated by the BEAMS3D code. This paper presents the results of the numerical analysis and the related operating limits.
W-7X初壁石墨瓦在直接NBI载荷下的热结构分析
对Wendelstein 7-X中性束流堆进行了热力学分析,重点研究了石墨瓦在强脉冲热载荷作用下的破坏现象。中性束注入(NBI)是Wendelstein 7-X仿星器的等离子体加热系统之一。目前有两个NBI箱,总注入氢束功率为7兆瓦,脉冲长度可达5秒。每个NBI盒可与四个正离子中性注射器(PINI)一起操作。在2022年的最后一次行动中,总共使用了三个pi。NBI对W7-X物理计划非常重要,因此计划在下一个操作阶段增加NBI功率。一部分NBI功率通过等离子体,在第一壁(FW)形成一个热负荷显著增加的区域。该区域被称为束流堆,可能是NBI脉冲长度的限制因素,因此需要在2024/2025年运行阶段之前对所有成分进行精确的重塑,以防止该区域的重复结构损坏。由于石墨瓦通过多个螺栓连接到CuCrZr散热器上,石墨的热膨胀会在石墨中引起显著的应力。石墨瓦的安全极限系数为NBI梁功率载荷引起的热应力。建立了包括螺栓连接在内的ANSYS力学模型,分析了热负荷对FW构件的影响,并评估了NBI梁的最大允许持续时间。进行了热学和结构分析的耦合瞬态模拟。计算采用BEAMS3D规范估算的FW表面NBI热负荷。本文给出了数值分析的结果和相应的运行限值。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
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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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