C. Marsden , X. Zhang , M. Moscheni , T.K. Gray , E. Vekshina , A. Rengle , A. Scarabosio , M. Sertoli , M. Romanelli , ST40 team
{"title":"Inferring the scrape-off layer heat flux width in a divertor with a low degree of axisymmetry","authors":"C. Marsden , X. Zhang , M. Moscheni , T.K. Gray , E. Vekshina , A. Rengle , A. Scarabosio , M. Sertoli , M. Romanelli , ST40 team","doi":"10.1016/j.nme.2024.101773","DOIUrl":null,"url":null,"abstract":"<div><div>Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MW/m<sup>2</sup>. The magnitude of these heat loads is set by the width of the channel, the ‘scrape-off layer’ (SOL), into which heat is exhausted, and can be characterised by an e-folding length scale for the decay of heat flux across the channel. It is expected this channel will narrow as tokamaks move towards reactor relevant conditions. Understanding the processes involved in setting the SOL heat flux width is imperative to be able to predict the heat loads PFCs must handle in future devices. Measurements of the SOL width are performed on the high-field spherical tokamak, ST40, using a newly commissioned infrared thermography system. With its high on-axis toroidal magnetic field (<span><math><mo>≥</mo></math></span>1.5 T) ST40 is uniquely positioned to investigate the influence of toroidal field on the heat flux width in spherical tokamaks, whilst also extending measurements of the SOL width in spherical tokamaks to increased poloidal field (<span><math><mo>≥</mo></math></span>0.3 T). Due to the divertor on ST40 having a low degree of axisymmetry, it is necessary for a set of radial measurements of the heat flux to be taken across the divertor, made possible using an automated toolchain that fully incorporates its 3D geometry. These radial profiles are combined with the magnetic geometry of the plasma to infer the width of the SOL, with both Eich and double exponential profiles of heat flux observed. A reduction in the heat flux is observed toroidally across part of the divertor, along with increased heat loads observed locally around the edges of the tiles. Future work in characterising the impact of tile misalignment and uncertainties in the reconstructed divertor magnetic geometry is required in order to further understand the observed heat flux patterns, as are additional investigations into the role potentially being played by an inhomogeneous sheath electric field.</div></div>","PeriodicalId":56004,"journal":{"name":"Nuclear Materials and Energy","volume":"41 ","pages":"Article 101773"},"PeriodicalIF":2.3000,"publicationDate":"2024-10-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Nuclear Materials and Energy","FirstCategoryId":"101","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S2352179124001960","RegionNum":2,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q1","JCRName":"NUCLEAR SCIENCE & TECHNOLOGY","Score":null,"Total":0}
引用次数: 0
Abstract
Plasma facing components (PFCs) in the next generation of tokamak devices will operate in challenging environments, with heat loads predicted to exceed 10 MW/m2. The magnitude of these heat loads is set by the width of the channel, the ‘scrape-off layer’ (SOL), into which heat is exhausted, and can be characterised by an e-folding length scale for the decay of heat flux across the channel. It is expected this channel will narrow as tokamaks move towards reactor relevant conditions. Understanding the processes involved in setting the SOL heat flux width is imperative to be able to predict the heat loads PFCs must handle in future devices. Measurements of the SOL width are performed on the high-field spherical tokamak, ST40, using a newly commissioned infrared thermography system. With its high on-axis toroidal magnetic field (1.5 T) ST40 is uniquely positioned to investigate the influence of toroidal field on the heat flux width in spherical tokamaks, whilst also extending measurements of the SOL width in spherical tokamaks to increased poloidal field (0.3 T). Due to the divertor on ST40 having a low degree of axisymmetry, it is necessary for a set of radial measurements of the heat flux to be taken across the divertor, made possible using an automated toolchain that fully incorporates its 3D geometry. These radial profiles are combined with the magnetic geometry of the plasma to infer the width of the SOL, with both Eich and double exponential profiles of heat flux observed. A reduction in the heat flux is observed toroidally across part of the divertor, along with increased heat loads observed locally around the edges of the tiles. Future work in characterising the impact of tile misalignment and uncertainties in the reconstructed divertor magnetic geometry is required in order to further understand the observed heat flux patterns, as are additional investigations into the role potentially being played by an inhomogeneous sheath electric field.
下一代托卡马克设备中的等离子体面组件(PFC)将在极具挑战性的环境中运行,其热负荷预计将超过 10 MW/m2。这些热负荷的大小取决于排出热量的通道--"刮除层"(SOL)的宽度,并可通过通道上热通量衰减的电子折叠长度尺度来描述。随着托卡马克向反应堆相关条件发展,预计这一通道将逐渐变窄。要预测 PFC 在未来设备中必须处理的热负荷,就必须了解设置 SOL 热通量宽度所涉及的过程。在高场球形托卡马克 ST40 上使用新投入使用的红外热成像系统对 SOL 宽度进行了测量。ST40 具有高轴环形磁场(≥1.5 T),在研究环形磁场对球形托卡马克热通量宽度的影响方面具有得天独厚的优势,同时还能将球形托卡马克中的 SOL 宽度测量扩展到更高的极性磁场(≥0.3 T)。由于 ST40 上的分流器轴对称程度较低,因此有必要对整个分流器的热通量进行径向测量。这些径向剖面图与等离子体的磁性几何形状相结合,推断出 SOL 的宽度,并观察到热通量的艾希曲线和双指数曲线。在分流器的部分环形区域观察到热通量的减少,同时在瓦片边缘的局部区域观察到热负荷的增加。为了进一步了解所观测到的热通量模式,需要在今后的工作中确定瓦片错位的影响以及重建的岔道磁几何形状的不确定性,还需要对不均匀鞘电场可能发挥的作用进行更多的研究。
期刊介绍:
The open-access journal Nuclear Materials and Energy is devoted to the growing field of research for material application in the production of nuclear energy. Nuclear Materials and Energy publishes original research articles of up to 6 pages in length.