Pub Date : 2024-07-16DOI: 10.1088/1741-4326/ad63b7
H. D. He, Yueqiang Liu, G. Hao, Jinxia Zhu, Yong Shen, Guoyao Zheng
Drift-kinetic effects of fusion-born alpha particles on the n=1 (n is the toroidal mode number) resistive wall mode (RWM) is numerically investigated for a recent design of the ITER 10 MA steady state plasma scenario, utilizing a magneto-hydrodynamic (MHD)-kinetic hybrid toroidal model. While the fluid theory predicts unstable RWM as the normalized plasma pressure βN exceeds the no-wall Troyon limit and with the mode growth rate monotonically increasing with βN, inclusion of the drift-kinetic contribution of trapped alphas qualitatively modifies the behavior by stabilizing the mode at high βN. In fact, a complete stabilization of the n=1 RWM up to the ideal-wall Troyon limit is found. On the other hand, another unstable branch - the alpha-driven n=1 fishbone mode (FB) – is identified in the high-βN regime, with the mode frequency matching that of the toroidal precession frequency of trapped alphas. Fast plasma toroidal flow however helps mitigate the FB instability. Kinetic stabilization of the RWM and flow stabilization of the (alpha-triggered) FB result in an enhancement of βN from the design value of 3.22 to 3.52 for the ITER scenario considered, while still maintaining stable plasma operation against the aforementioned MHD instabilities.
{"title":"Resistive wall mode and fishbone mode in ITER steady state scenario: roles of fusion-born alphas and plasma flow","authors":"H. D. He, Yueqiang Liu, G. Hao, Jinxia Zhu, Yong Shen, Guoyao Zheng","doi":"10.1088/1741-4326/ad63b7","DOIUrl":"https://doi.org/10.1088/1741-4326/ad63b7","url":null,"abstract":"\u0000 Drift-kinetic effects of fusion-born alpha particles on the n=1 (n is the toroidal mode number) resistive wall mode (RWM) is numerically investigated for a recent design of the ITER 10 MA steady state plasma scenario, utilizing a magneto-hydrodynamic (MHD)-kinetic hybrid toroidal model. While the fluid theory predicts unstable RWM as the normalized plasma pressure βN exceeds the no-wall Troyon limit and with the mode growth rate monotonically increasing with βN, inclusion of the drift-kinetic contribution of trapped alphas qualitatively modifies the behavior by stabilizing the mode at high βN. In fact, a complete stabilization of the n=1 RWM up to the ideal-wall Troyon limit is found. On the other hand, another unstable branch - the alpha-driven n=1 fishbone mode (FB) – is identified in the high-βN regime, with the mode frequency matching that of the toroidal precession frequency of trapped alphas. Fast plasma toroidal flow however helps mitigate the FB instability. Kinetic stabilization of the RWM and flow stabilization of the (alpha-triggered) FB result in an enhancement of βN from the design value of 3.22 to 3.52 for the ITER scenario considered, while still maintaining stable plasma operation against the aforementioned MHD instabilities.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"28 2","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141643876","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-16DOI: 10.1088/1741-4326/ad63ba
J. Karhunen, S. Henderson, A. Järvinen, David Moulton, Sarah L Newton, Ryoko Tatsumi Osawa
Introduction of cross-field drifts in SOLPS-ITER simulations of connected double-null plasmas in STEP with the ion ∇B×B drift towards the upper divertors was found to enhance the detachment of the inner divertors with decreased target densities and ion and heat fluxes, while simultaneously complicating the access to detachment in the outer lower divertor by increasing the target temperature and heat loads to levels above the engineering limits.The ∇B×B drift was observed to significantly affect the radial heat transport between the core and the scrape-off layer (SOL), altering the poloidal temperature and pressure profiles and, consequently, the poloidal conductive and convective heat transport in the SOL. As a result, up-down asymmetries of 52:48 and 58:42 biased towards the outer lower and upper inner divertors, respectively, were observed to arise in the unmitigated power entering the divertor regions, breaking the up-down symmetry seen in simulations without the drift terms and contradicting with earlier experimental observations on the low-field side. Moreover, the upstream electron density was found to decrease noticeably in the core and separatrix regions following the activation of the drifts due to an increased share of the neutrals arriving from D2 injections near the upper and lower X-points ionizing already in the private flux regions.
{"title":"First SOLPS-ITER predictions of the impact of cross-field drifts on divertor and scrape-off layer conditions in double-null configuration of STEP","authors":"J. Karhunen, S. Henderson, A. Järvinen, David Moulton, Sarah L Newton, Ryoko Tatsumi Osawa","doi":"10.1088/1741-4326/ad63ba","DOIUrl":"https://doi.org/10.1088/1741-4326/ad63ba","url":null,"abstract":"\u0000 Introduction of cross-field drifts in SOLPS-ITER simulations of connected double-null plasmas in STEP with the ion ∇B×B drift towards the upper divertors was found to enhance the detachment of the inner divertors with decreased target densities and ion and heat fluxes, while simultaneously complicating the access to detachment in the outer lower divertor by increasing the target temperature and heat loads to levels above the engineering limits.The ∇B×B drift was observed to significantly affect the radial heat transport between the core and the scrape-off layer (SOL), altering the poloidal temperature and pressure profiles and, consequently, the poloidal conductive and convective heat transport in the SOL. As a result, up-down asymmetries of 52:48 and 58:42 biased towards the outer lower and upper inner divertors, respectively, were observed to arise in the unmitigated power entering the divertor regions, breaking the up-down symmetry seen in simulations without the drift terms and contradicting with earlier experimental observations on the low-field side. Moreover, the upstream electron density was found to decrease noticeably in the core and separatrix regions following the activation of the drifts due to an increased share of the neutrals arriving from D2 injections near the upper and lower X-points ionizing already in the private flux regions.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"6 5","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141644259","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-16DOI: 10.1088/1741-4326/ad63b8
Hua Yang, Youwen Sun, M. Jia, A. Loarte, P. Xie, Q. Ma, Xuemin Wu, Cheng Ye, Yueqiang Liu, Jiale Chen, Ruirong Liang, Zhendong Yang, Gaoting Chen, Bin Zhang, Qing Zang, Kaixuan Ye
The experiment in EAST demonstrates effective modulation of the stationary heat flux to the secondary lobes of the magnetic footprint induced by the resonant magnetic perturbations (RMPs) by slightly varying the equilibrium q95, consistent with prior numerical modeling. During the small q95 variation, the edge localized mode (ELM) control is well maintained, and the position of the secondary heat flux peak is effectively shifted, thus avoiding a specific location heat flux accumulation. As the divertor heat load is one of the significant concerns in tokamak, these results provide a promising choice, varying magnetic equilibrium periodically to shift stationary heat load deposition position during static n = 4 ( n is the toroidal mode number) RMP condition, for further fusion devices. In this respect the use of this technique for n =4 RMPs is advantageous because the q95 range that needs to be covered to spread the divertor heat load is reduced because of the smaller toroidal extent of off-separatrix heat deposition zones compared to lower n’s.
EAST 中的实验表明，通过稍微改变平衡 q95，可以有效调节共振磁扰动 (RMP) 诱导的磁足迹次叶的静态热通量，这与之前的数值建模一致。在 q95 的微小变化过程中，边缘局部模式 (ELM) 控制得到了很好的维持，二次热通量峰值的位置得到了有效的移动，从而避免了特定位置热通量的积累。在静态 n = 4（n 为环模数）RMP 条件下，由于岔道热负荷是托卡马克的主要问题之一，这些结果为进一步的核聚变装置提供了一个很有前途的选择，即周期性地改变磁平衡来移动静态热负荷沉积位置。在这方面，对 n = 4 RMP 使用这种技术是有优势的，因为与较低的 n 相比，偏离分离矩阵热沉积区的环形范围较小，因此分散分流器热负荷所需要覆盖的 q95 范围也就减小了。
{"title":"Dynamic control of divertor heat flux during n = 4 RMP ELM suppression by small variation of q95 in EAST","authors":"Hua Yang, Youwen Sun, M. Jia, A. Loarte, P. Xie, Q. Ma, Xuemin Wu, Cheng Ye, Yueqiang Liu, Jiale Chen, Ruirong Liang, Zhendong Yang, Gaoting Chen, Bin Zhang, Qing Zang, Kaixuan Ye","doi":"10.1088/1741-4326/ad63b8","DOIUrl":"https://doi.org/10.1088/1741-4326/ad63b8","url":null,"abstract":"\u0000 The experiment in EAST demonstrates effective modulation of the stationary heat flux to the secondary lobes of the magnetic footprint induced by the resonant magnetic perturbations (RMPs) by slightly varying the equilibrium q95, consistent with prior numerical modeling. During the small q95 variation, the edge localized mode (ELM) control is well maintained, and the position of the secondary heat flux peak is effectively shifted, thus avoiding a specific location heat flux accumulation. As the divertor heat load is one of the significant concerns in tokamak, these results provide a promising choice, varying magnetic equilibrium periodically to shift stationary heat load deposition position during static n = 4 ( n is the toroidal mode number) RMP condition, for further fusion devices. In this respect the use of this technique for n =4 RMPs is advantageous because the q95 range that needs to be covered to spread the divertor heat load is reduced because of the smaller toroidal extent of off-separatrix heat deposition zones compared to lower n’s.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"2 12","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141641736","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-16DOI: 10.1088/1741-4326/ad63b9
Christian Hopf, N. den Harder, B. Heinemann, C. Angioni, U. Plank, M. Weiland
In early 2022 one source of ASDEX Upgrade's (AUG) neutral beam injector 2 was equipped with a first-of-its-kind beam extraction grid system with in-situ variable extraction/acceleration gap that allows one to choose beam energy and beam power independently in a wide operational space, greatly enhancing experimental flexibility. The gap can be changed from one AUG discharge to the next. The extended operational space makes it possible to reduce beam energy and shine through while maintaining high heating power, or to reduce NBI power at high beam energy, e.g. to stay in L mode. Furthermore, the feature opens the door for advanced control of heat and torque deposition, such as to change torque and ion-to-electron heating ratio at constant power. The prototype system was successfully tested in 2022 and already found first applications in AUG's physics programme. The remaining three sources of the same injector will also be equipped with variable gaps during the 2022-24 opening of AUG and installation of this new system on all sources of NBI 1 is also under discussion in order to exploit the full potential of the new feature.
2022 年初，ASDEX 升级版（AUG）中性束注入器 2 的一个光源配备了首创的光束提取网格系统，该系统具有原位可变提取/加速间隙，可在宽广的操作空间内独立选择光束能量和光束功率，大大提高了实验的灵活性。间隙可以从一次 AUG 放电改变到下一次。扩展的操作空间使得在保持高加热功率的同时降低光束能量和穿透力，或在高光束能量下降低 NBI 功率（如保持 L 模式）成为可能。此外，该功能还为热量和扭矩沉积的高级控制打开了大门，例如在恒定功率下改变扭矩和离子-电子加热比。原型系统已于 2022 年成功通过测试，并在 AUG 的物理项目中首次得到应用。在 2022-24 年 AUG 启用期间，同一喷射器的其余三个源也将配备可变间隙，并且正在讨论在 NBI 1 的所有源上安装这一新系统，以充分挖掘新功能的潜力。
{"title":"Decoupling beam power and beam energy on ASDEX Upgrade NBI with an in-situ variable extraction gap system","authors":"Christian Hopf, N. den Harder, B. Heinemann, C. Angioni, U. Plank, M. Weiland","doi":"10.1088/1741-4326/ad63b9","DOIUrl":"https://doi.org/10.1088/1741-4326/ad63b9","url":null,"abstract":"\u0000 In early 2022 one source of ASDEX Upgrade's (AUG) neutral beam injector 2 was equipped with a first-of-its-kind beam extraction grid system with in-situ variable extraction/acceleration gap that allows one to choose beam energy and beam power independently in a wide operational space, greatly enhancing experimental flexibility. The gap can be changed from one AUG discharge to the next. The extended operational space makes it possible to reduce beam energy and shine through while maintaining high heating power, or to reduce NBI power at high beam energy, e.g. to stay in L mode. Furthermore, the feature opens the door for advanced control of heat and torque deposition, such as to change torque and ion-to-electron heating ratio at constant power. The prototype system was successfully tested in 2022 and already found first applications in AUG's physics programme. The remaining three sources of the same injector will also be equipped with variable gaps during the 2022-24 opening of AUG and installation of this new system on all sources of NBI 1 is also under discussion in order to exploit the full potential of the new feature.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"8 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141641329","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-16DOI: 10.1088/1741-4326/ad63bb
Wei Zhang, Lunan Liu, Xinjun Zhang, Chengming Qin, Hua Yang
Efficient ion cyclotron range of frequencies (ICRF) wave heating requires good wave coupling at the plasma edge and good radio frequency power absorption in the plasma core. This study reviews recent progress in improving these two aspects of ICRF heating with the new two-strap antennas through various experiments and simulations on the Experimental Advanced Superconducting Tokamak (EAST). Our study shows that the ICRF coupling can be significantly improved by decreasing the parallel wave number, increasing the local scrape-off layer (SOL) density by midplane gas puffing, and increasing the global SOL density by decreasing the separatrix–antenna distance. It can also be improved by increasing the core plasma density, changing the divertor strike point position, and optimizing the antenna phasing. The core ICRF power absorption can be increased by optimizing the cyclotron resonance position and minority ion concentration and by applying new heating schemes such as three-ion heating. Although some of the methods have been previously studied on other devices, improving ICRF coupling by shifting the divertor strike point was tested on EAST for the first time. Quantitative characterization of these methods and the conclusions drawn from this study can provide important insights for achieving more efficient ICRF heating in current and future fusion machines.
高效离子回旋频率范围（ICRF）波加热需要等离子体边缘良好的波耦合和等离子体核心良好的射频功率吸收。本研究通过在实验性先进超导托卡马克（EAST）上进行的各种实验和模拟，回顾了利用新型双带天线改善离子回旋频率范围波加热的这两个方面的最新进展。我们的研究表明，通过减少平行波数、通过中面气体膨化增加局部刮除层（SOL）密度以及通过减少分离矩阵-天线距离增加全局 SOL 密度，可以显著改善 ICRF 耦合。此外，还可以通过增加核心等离子体密度、改变分流器打击点位置和优化天线相位来改善功率吸收。通过优化回旋共振位置和少数离子浓度，以及采用新的加热方案（如三离子加热），可以提高核心 ICRF 功率吸收。虽然其中一些方法之前已在其他设备上进行过研究，但通过转移分流器打击点来改善ICRF耦合的方法是首次在EAST上进行测试。对这些方法的定量表征以及从这项研究中得出的结论，可以为在当前和未来的聚变机器中实现更高效的ICRF加热提供重要的启示。
{"title":"Recent progress in improvement in ion cyclotron range of frequencies coupling and power absorption with new antennas of Experimental Advanced Superconducting Tokamak (EAST)","authors":"Wei Zhang, Lunan Liu, Xinjun Zhang, Chengming Qin, Hua Yang","doi":"10.1088/1741-4326/ad63bb","DOIUrl":"https://doi.org/10.1088/1741-4326/ad63bb","url":null,"abstract":"\u0000 Efficient ion cyclotron range of frequencies (ICRF) wave heating requires good wave coupling at the plasma edge and good radio frequency power absorption in the plasma core. This study reviews recent progress in improving these two aspects of ICRF heating with the new two-strap antennas through various experiments and simulations on the Experimental Advanced Superconducting Tokamak (EAST). Our study shows that the ICRF coupling can be significantly improved by decreasing the parallel wave number, increasing the local scrape-off layer (SOL) density by midplane gas puffing, and increasing the global SOL density by decreasing the separatrix–antenna distance. It can also be improved by increasing the core plasma density, changing the divertor strike point position, and optimizing the antenna phasing. The core ICRF power absorption can be increased by optimizing the cyclotron resonance position and minority ion concentration and by applying new heating schemes such as three-ion heating. Although some of the methods have been previously studied on other devices, improving ICRF coupling by shifting the divertor strike point was tested on EAST for the first time. Quantitative characterization of these methods and the conclusions drawn from this study can provide important insights for achieving more efficient ICRF heating in current and future fusion machines.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"4 10","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141642283","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-15DOI: 10.1088/1741-4326/ad6335
A. Hakola, M. Balden, Matteo Baruzzo, R. Bisson, S. Brezinsek, T. Dittmar, D. Douai, Michael G Dunne, L. Garzotti, M. Groth, Rafael Henriques, Laszlo Horvath, I. Jepu, E. Joffrin, A. Kappatou, D. Keeling, K. Krieger, Benoit Labit, Morten Lennholm, J. Likonen, A. Loarte, Peter Lomas, C. G. Lowry, M. Maslov, D. Matveev, R. Pitts, U. Plank, M. Rasinski, D. Ryan, S. Saarelma, S. Silburn, E. Solano, W. Suttrop, Tuomas J J Tala, Emmanuelle Tsitrone, N. Vianello, T. Wauters, A. Widdowson, M. Wischmeier
For its initial operational phase, ITER has until recently considered using non-nuclear hydrogen (H) or helium (He) plasmas to keep nuclear activation at low levels. To this end, the Tokamak Exploitation Task Force of the EUROfusion Consortium carried out dedicated experimental campaigns in He on the ASDEX Upgrade (AUG) and JET tokamaks in 2022, with particular emphasis put on the ELMy H-mode operation and plasma-wall interaction processes as well as comparison to H or deuterium (D) plasmas. Both in pure He and mixed He+H plasmas, H-mode operation could be reached but more effort was needed to obtain a stable plasma scenario than in H or D. Even if the power threshold for the LH transition was lower in He, entering the type-I ELMy regime appeared to require equally much or even more heating power than in H. Suppression of ELMs by resonant magnetic perturbations was studied on AUG but was only possible in plasmas with a He content below 19%; the reason for this unexpected behaviour remains still unclear and various theoretical approaches are being pursued to properly understand the physics behind ELM suppression. The erosion rates of tungsten (W) plasma-facing components (PFCs) were an order of magnitude larger than what has been reported in hydrogenic plasmas, which can be attributed to the prominent role of He2+ ions in the plasma. For the first time, the formation of nanoscale structures (W fuzz) was unambiguously demonstrated in H-mode He plasmas on AUG. However, no direct evidence of fuzz creation on JET was obtained despite the main conditions for its occurrence being met. The reason could be a delicate balance between W erosion by ELMs, competition between the growth and annealing of the fuzz, and coverage of the surface with co-deposits.
直到最近，热核实验堆还在考虑在其初始运行阶段使用非核氢（H）或氦（He）等离子体，以将核激活保持在较低水平。为此，EUROfusion 联合会托卡马克利用特别工作组于 2022 年在 ASDEX 升级（AUG）和 JET 托卡马克上开展了专门的 He 实验活动，重点是 ELMy H 模式运行和等离子体壁相互作用过程，以及与氢或氘等离子体的比较。在纯 He 和 He+H 混合等离子体中，都可以实现 H 模式运行，但与在 H 或 D 等离子体中相比，需要付出更多努力才能获得稳定的等离子体情况。即使在 He 中 LH 转变的功率阈值较低，进入 I 型 ELMy 状态似乎需要同样多甚至比在 H 中更多的加热功率。在 AUG 上研究了共振磁扰动对 ELM 的抑制，但只有在 He 含量低于 19% 的等离子体中才有可能发生；这种意外行为的原因仍不清楚，目前正在寻求各种理论方法，以正确理解 ELM 抑制背后的物理学原理。钨（W）等离子体面元件（PFC）的侵蚀率比氢等离子体中的侵蚀率大一个数量级，这可以归因于等离子体中 He2+ 离子的突出作用。在 AUG 上的 H 模式 He 等离子体中，首次明确证明了纳米级结构（W 绒毛）的形成。然而，尽管发生模糊的主要条件已经满足，但在 JET 上却没有获得模糊产生的直接证据。原因可能是 ELM 对 W 的侵蚀、模糊生长和退火之间的竞争以及共沉积物覆盖表面之间的微妙平衡。
{"title":"Helium plasma operations on ASDEX Upgrade and JET in support of the non-nuclear phases of ITER","authors":"A. Hakola, M. Balden, Matteo Baruzzo, R. Bisson, S. Brezinsek, T. Dittmar, D. Douai, Michael G Dunne, L. Garzotti, M. Groth, Rafael Henriques, Laszlo Horvath, I. Jepu, E. Joffrin, A. Kappatou, D. Keeling, K. Krieger, Benoit Labit, Morten Lennholm, J. Likonen, A. Loarte, Peter Lomas, C. G. Lowry, M. Maslov, D. Matveev, R. Pitts, U. Plank, M. Rasinski, D. Ryan, S. Saarelma, S. Silburn, E. Solano, W. Suttrop, Tuomas J J Tala, Emmanuelle Tsitrone, N. Vianello, T. Wauters, A. Widdowson, M. Wischmeier","doi":"10.1088/1741-4326/ad6335","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6335","url":null,"abstract":"\u0000 For its initial operational phase, ITER has until recently considered using non-nuclear hydrogen (H) or helium (He) plasmas to keep nuclear activation at low levels. To this end, the Tokamak Exploitation Task Force of the EUROfusion Consortium carried out dedicated experimental campaigns in He on the ASDEX Upgrade (AUG) and JET tokamaks in 2022, with particular emphasis put on the ELMy H-mode operation and plasma-wall interaction processes as well as comparison to H or deuterium (D) plasmas. Both in pure He and mixed He+H plasmas, H-mode operation could be reached but more effort was needed to obtain a stable plasma scenario than in H or D. Even if the power threshold for the LH transition was lower in He, entering the type-I ELMy regime appeared to require equally much or even more heating power than in H. Suppression of ELMs by resonant magnetic perturbations was studied on AUG but was only possible in plasmas with a He content below 19%; the reason for this unexpected behaviour remains still unclear and various theoretical approaches are being pursued to properly understand the physics behind ELM suppression. The erosion rates of tungsten (W) plasma-facing components (PFCs) were an order of magnitude larger than what has been reported in hydrogenic plasmas, which can be attributed to the prominent role of He2+ ions in the plasma. For the first time, the formation of nanoscale structures (W fuzz) was unambiguously demonstrated in H-mode He plasmas on AUG. However, no direct evidence of fuzz creation on JET was obtained despite the main conditions for its occurrence being met. The reason could be a delicate balance between W erosion by ELMs, competition between the growth and annealing of the fuzz, and coverage of the surface with co-deposits.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"14 8","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141648101","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-15DOI: 10.1088/1741-4326/ad6336
Philipp Lauber, M. Falessi, G. Meng, T. Hayward-Schneider, Virgil - Alin Popa, F. Zonca, Mireille Schneider
In this paper we report on the implementation and verification of a phase- space resolved energetic particle (EP) transport model. It is based on a first-principle theoretical framework, i.e. the system of non-linear gyrokinetic equations and the related transport equations. Its focus is primarily directed toward understanding the meso-scale character of EPs and its consequences. Compared to the conventional description of thermal radial transport via a one-dimensional radial diffusion equation, the newly developed model is three-dimensional using canonical constants-of-motion (CoM) variables. The model does not assume diffusive processes to be dominant a priori, instead the EP fluxes are self-consistently calculated and directly evolved in CoM space. We use the EP-Stability workflow and the HAGIS code to determine the phase space fluxes explicitly either in the limit of constant mode amplitudes or an energy-conserving quasi-linear model. As an application of the model the transport of neutral-beam-generated EPs due to a toroidal Alfv ́en eigenmode in an ITER plasma is investigated. As there are no sources and collisions taken into account so far (for an extension of the model see the companion paper ref. [1]), the results cannot be considered as an exhaustive study, but rather as a practical demonstration of the conceptual framework on the way to a comprehensive reduced description of burning plasmas.
{"title":"ATEP: An advanced transport model for energetic particles","authors":"Philipp Lauber, M. Falessi, G. Meng, T. Hayward-Schneider, Virgil - Alin Popa, F. Zonca, Mireille Schneider","doi":"10.1088/1741-4326/ad6336","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6336","url":null,"abstract":"\u0000 In this paper we report on the implementation and verification of a phase- space resolved energetic particle (EP) transport model. It is based on a first-principle theoretical framework, i.e. the system of non-linear gyrokinetic equations and the related transport equations. Its focus is primarily directed toward understanding the meso-scale character of EPs and its consequences. Compared to the conventional description of thermal radial transport via a one-dimensional radial diffusion equation, the newly developed model is three-dimensional using canonical constants-of-motion (CoM) variables. The model does not assume diffusive processes to be dominant a priori, instead the EP fluxes are self-consistently calculated and directly evolved in CoM space. We use the EP-Stability workflow and the HAGIS code to determine the phase space fluxes explicitly either in the limit of constant mode amplitudes or an energy-conserving quasi-linear model. As an application of the model the transport of neutral-beam-generated EPs due to a toroidal Alfv ́en eigenmode in an ITER plasma is investigated. As there are no sources and collisions taken into account so far (for an extension of the model see the companion paper ref. [1]), the results cannot be considered as an exhaustive study, but rather as a practical demonstration of the conceptual framework on the way to a comprehensive reduced description of burning plasmas.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"14 20","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141645780","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-15DOI: 10.1088/1741-4326/ad6338
H. Tanabe, Yunhan Cai, Haruaki Tanaka, T. Ahmadi, M. Inomoto, Y. Ono
Here we report ion heating/transport characteristics of merging startup scenario in the TS-6 spherical tokamak. In addition to the previously investigated impulsive heating process during magnetic reconnection, here we also focused on a longer time scale response of ion temperature profile both during and after merging including the semi-steady plasma confinement phase. During magnetic reconnection, (i) ion temperature profile forms poloidally asymmetric profile around the X-point at the initiation phase and (ii) radially asymmetric higher deposition is obtained in the high field side. After merging, (iii) the radially asymmetric double-peak structure is affected by parallel heat conduction and is aligned with field lines, but it does not simply become flux function in microsecond time scale: inboard/outboard asymmetry lasts even in the semi-steady confinement phase. (iv) Under the influence of low aspect ratio configuration, there is 2~3 times higher toroidal field in the high field side on a same closed flux surface, characteristic asymmetry of inboard/outboard ion temperature has experimentally been found for the first time.
{"title":"Ion heating/transport characteristics of merging startup plasma scenario in the TS-6 spherical tokamak","authors":"H. Tanabe, Yunhan Cai, Haruaki Tanaka, T. Ahmadi, M. Inomoto, Y. Ono","doi":"10.1088/1741-4326/ad6338","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6338","url":null,"abstract":"\u0000 Here we report ion heating/transport characteristics of merging startup scenario in the TS-6 spherical tokamak. In addition to the previously investigated impulsive heating process during magnetic reconnection, here we also focused on a longer time scale response of ion temperature profile both during and after merging including the semi-steady plasma confinement phase. During magnetic reconnection, (i) ion temperature profile forms poloidally asymmetric profile around the X-point at the initiation phase and (ii) radially asymmetric higher deposition is obtained in the high field side. After merging, (iii) the radially asymmetric double-peak structure is affected by parallel heat conduction and is aligned with field lines, but it does not simply become flux function in microsecond time scale: inboard/outboard asymmetry lasts even in the semi-steady confinement phase. (iv) Under the influence of low aspect ratio configuration, there is 2~3 times higher toroidal field in the high field side on a same closed flux surface, characteristic asymmetry of inboard/outboard ion temperature has experimentally been found for the first time.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"24 12","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141649304","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The interaction and interpenetration of two counter-propagating plasma shocks are investigated via hybrid fluid-PIC (particle-in-cell) simulations. This study seeks to probe the kinetic effects and ion collisions on the structure of colliding plasma shocks in complex multi-ion-species plasma, in particular, the presence of the expansion of high-Z plasma bubbles against the low-Z filled gas inside an ICF hohlraum. The superposition of shock wave results in a wave-like electric field in the downstream region. The electric field can further reduce the kinetic energy of the incoming particles, and modulate the ion density profile. It finally generates a new downstream platform of high temperature and high density. However, on the hundred-ps time scale, cumulative ion collisions can still significantly alter the structure of the shock wave and the reflection of ions by the shock front. This study will help to improve the predictions of hohlraum plasma states and the understanding of the shock wave interactions.
通过混合流体-PIC（粒子在单元中）模拟研究了两个反向传播等离子体冲击的相互作用和相互渗透。这项研究试图探究复杂多离子种等离子体中碰撞等离子体冲击波结构的动力学效应和离子碰撞，特别是 ICF 霍尔阱内高 Z 等离子体气泡对低 Z 填充气体的膨胀。冲击波的叠加会在下游区域产生波状电场。电场可进一步降低进入粒子的动能，并调节离子密度曲线。最终产生一个新的高温、高密度下游平台。然而，在数百 ps 的时间尺度上，累积离子碰撞仍会显著改变冲击波的结构和冲击前沿对离子的反射。这项研究将有助于改进对霍尔姆等离子体状态的预测和对冲击波相互作用的理解。
{"title":"Kinetic effects on the interaction of counter-propagating plasma shocks inside an ICF hohlraum","authors":"Xu Zhang, Qing-kang Liu, Wen-shuai Zhang, E. Zhang, Xiaochuan Ning, Fan-qi Meng, Yi-peng Wang, Hongbo Cai, Shao-ping Zhu","doi":"10.1088/1741-4326/ad61fd","DOIUrl":"https://doi.org/10.1088/1741-4326/ad61fd","url":null,"abstract":"\u0000 The interaction and interpenetration of two counter-propagating plasma shocks are investigated via hybrid fluid-PIC (particle-in-cell) simulations. This study seeks to probe the kinetic effects and ion collisions on the structure of colliding plasma shocks in complex multi-ion-species plasma, in particular, the presence of the expansion of high-Z plasma bubbles against the low-Z filled gas inside an ICF hohlraum. The superposition of shock wave results in a wave-like electric field in the downstream region. The electric field can further reduce the kinetic energy of the incoming particles, and modulate the ion density profile. It finally generates a new downstream platform of high temperature and high density. However, on the hundred-ps time scale, cumulative ion collisions can still significantly alter the structure of the shock wave and the reflection of ions by the shock front. This study will help to improve the predictions of hohlraum plasma states and the understanding of the shock wave interactions.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"13 3","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141655780","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-08DOI: 10.1088/1741-4326/ad6012
Morten Lennholm, Spyridon Aleiferis, Sam Bakes, O. Bardsley, M. van Berkel, Francis J Casson, Fazal Chaudry, Neil Conway, T. Hender, S. Henderson, A. Hudoba, B. Kool, Mark Lafferty, H. Meyer, Joshua Mitchell, Ankit Mitra, Ryoko Tatsumi Osawa, R. Otin, Adam Parrot, Terry Michael Thompson, Guoliang Xia
In 2019 the UK launched the STEP programme to design and build a prototype electricity producing nuclear fusion power plant, aiming to start operation around 2040. The plant should lay the foundation for the development of commercial nuclear fusion power plants. The design is based on the spherical tokamak principle, which opens a route to high pressure, steady state, operation. While facilitating steady state operation, the spherical design introduces some specific plasma control challenges: i) All plasma current during the burn phase should to be generated through non-inductively means, dominated by bootstrap current. This leads to operation at high normalised plasma pressure β_"N" with high plasma elongation, which in turn imposes effective active stabilisation of the vertical plasma position. ii) The tight aspect ratio means very limited space for a central solenoid, imposing that even the current ramp up must be non-inductively generated. iii) The compact design leads to extreme heat loads on plasma facing components. A double null design has been chosen to spread this load, putting strict demands on the control of the unstable vertical plasma position. iv) The heat pulses associated with unmitigated ELMs are unlikely to be unacceptable imposing ELM free operation or active ELM control. v) To reduce and spread heat loads, core and divertor radiation and momentum loss has to be controlled, aiming to operate with simultaneously detached upper and lower divertors. vi) High pressure operation is likely to require active resistive wall mode stabilisation. vii) The conductivity distribution in structures near the plasma must be carefully selected to reduce the growth rates for the vertical instability and the resistive wall mode without damping the penetration of the of magnetic fields from active control coils too much. This article describes the initial work carried out to develop a STEP plasma control system.
{"title":"Plasma control for the STEP prototype power plant","authors":"Morten Lennholm, Spyridon Aleiferis, Sam Bakes, O. Bardsley, M. van Berkel, Francis J Casson, Fazal Chaudry, Neil Conway, T. Hender, S. Henderson, A. Hudoba, B. Kool, Mark Lafferty, H. Meyer, Joshua Mitchell, Ankit Mitra, Ryoko Tatsumi Osawa, R. Otin, Adam Parrot, Terry Michael Thompson, Guoliang Xia","doi":"10.1088/1741-4326/ad6012","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6012","url":null,"abstract":"\u0000 In 2019 the UK launched the STEP programme to design and build a prototype electricity producing nuclear fusion power plant, aiming to start operation around 2040. The plant should lay the foundation for the development of commercial nuclear fusion power plants. The design is based on the spherical tokamak principle, which opens a route to high pressure, steady state, operation. While facilitating steady state operation, the spherical design introduces some specific plasma control challenges: i) All plasma current during the burn phase should to be generated through non-inductively means, dominated by bootstrap current. This leads to operation at high normalised plasma pressure β_\"N\" with high plasma elongation, which in turn imposes effective active stabilisation of the vertical plasma position. ii) The tight aspect ratio means very limited space for a central solenoid, imposing that even the current ramp up must be non-inductively generated. iii) The compact design leads to extreme heat loads on plasma facing components. A double null design has been chosen to spread this load, putting strict demands on the control of the unstable vertical plasma position. iv) The heat pulses associated with unmitigated ELMs are unlikely to be unacceptable imposing ELM free operation or active ELM control. v) To reduce and spread heat loads, core and divertor radiation and momentum loss has to be controlled, aiming to operate with simultaneously detached upper and lower divertors. vi) High pressure operation is likely to require active resistive wall mode stabilisation. vii) The conductivity distribution in structures near the plasma must be carefully selected to reduce the growth rates for the vertical instability and the resistive wall mode without damping the penetration of the of magnetic fields from active control coils too much. This article describes the initial work carried out to develop a STEP plasma control system.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"111 45","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141667917","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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