Pub Date : 2024-04-16DOI: 10.1088/1741-4326/ad3f2c
Philippe Jacquet, Pierre Dumortier, E. Lerche, I. Monakhov, Craig Noble, Jason Roberts, Haroon K Sheikh, Alex Goodyear, N. Balshaw, Dragoslav Ciric, R. Lobel, Peter Lomas, Chris Lowry, F. Rimini, S. Silburn, Lorne D Horton
The JET-ILW pure Tritium and Deuterium-Tritium (DTE2) experimental campaigns took place in 2021-2022. Tritium (T) and Deuterium-Tritium (D-T) operations present challenges not encountered in present day tokamaks [1]. This contribution focuses on Ion Cyclotron Resonance Heating (ICRH) operations in tritium and deuterium-tritium plasmas, starting with a summary of the program of improvements to the ICRH system which spanned a few years prior to these experiments. Procedures were implemented to address specific constraints from tritium and deuterium-tritium operations (tritium safety and reduced access to the RF generator area) and increase the system reliability and power availability during plasma pulses. Operation of the upgraded Real Time RF power control system that maximises the launched power while taking into account limitations from the system or antenna coupling is described. We also report on the result from dedicated pulses performed to assess the potential harmful impact of the 2nd harmonic tritium resonance in the plasma, close to the inner wall, when using the standard central hydrogen minority ICRH scheme. During DTE2, the ITER-Like Antenna (ILA) was not used because water leaked from an in-vessel capacitor into the vessel on day-2 of the experimental campaign. The lessons learnt from this incident are highlighted. Finally, the ICRH plant adjustments required to safely perform Ion Cyclotron Wall Cleaning discharges are described.
{"title":"ICRH operations during the JET Tritium and DTE2 campaigns","authors":"Philippe Jacquet, Pierre Dumortier, E. Lerche, I. Monakhov, Craig Noble, Jason Roberts, Haroon K Sheikh, Alex Goodyear, N. Balshaw, Dragoslav Ciric, R. Lobel, Peter Lomas, Chris Lowry, F. Rimini, S. Silburn, Lorne D Horton","doi":"10.1088/1741-4326/ad3f2c","DOIUrl":"https://doi.org/10.1088/1741-4326/ad3f2c","url":null,"abstract":"\u0000 The JET-ILW pure Tritium and Deuterium-Tritium (DTE2) experimental campaigns took place in 2021-2022. Tritium (T) and Deuterium-Tritium (D-T) operations present challenges not encountered in present day tokamaks [1]. This contribution focuses on Ion Cyclotron Resonance Heating (ICRH) operations in tritium and deuterium-tritium plasmas, starting with a summary of the program of improvements to the ICRH system which spanned a few years prior to these experiments. Procedures were implemented to address specific constraints from tritium and deuterium-tritium operations (tritium safety and reduced access to the RF generator area) and increase the system reliability and power availability during plasma pulses. Operation of the upgraded Real Time RF power control system that maximises the launched power while taking into account limitations from the system or antenna coupling is described. We also report on the result from dedicated pulses performed to assess the potential harmful impact of the 2nd harmonic tritium resonance in the plasma, close to the inner wall, when using the standard central hydrogen minority ICRH scheme. During DTE2, the ITER-Like Antenna (ILA) was not used because water leaked from an in-vessel capacitor into the vessel on day-2 of the experimental campaign. The lessons learnt from this incident are highlighted. Finally, the ICRH plant adjustments required to safely perform Ion Cyclotron Wall Cleaning discharges are described.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"31 4","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-04-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140696884","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 : 2024-04-15DOI: 10.1088/1741-4326/ad3e8c
T. Hemmi, B. Bellesia, Shin Hasegawa, Piergiorgio Aprili, A. Bonito-Oliva, E. Boter, Maria Paz Casas, Marc Cornelis, K. Gomikawa, Robert Harrison, Angela Hernandez-Sanchez, M. Iguchi, Takaaki Isono, M. Jiménez, Hideki Kajitani, S. Koczorowski, Norikiyo Koizumi, A. Lo Bue, K. Matsui, Masataka Nakahira, M. Nakamoto, K. Saito, Kazuyuki Sakamoto, T. Sakurai, Tatsuya Shimizu, T. Suwa, K. Takano, Keiya Takebayashi, Nobuhiko Tanaka, F. Tsutsumi, Yasuhiro Uno, Eduard Viladiu Martinez
In 2023, the manufacturing of all the ITER TF coils has been completed 15 years after the sign of procurement arrangements in 2008. This paper has been jointly submitted by F4E and QST to recollect the lessons learnt in the production of the two parties along the last 15 years.
{"title":"Lessons learned from European and Japanese production of ITER toroidal field coils","authors":"T. Hemmi, B. Bellesia, Shin Hasegawa, Piergiorgio Aprili, A. Bonito-Oliva, E. Boter, Maria Paz Casas, Marc Cornelis, K. Gomikawa, Robert Harrison, Angela Hernandez-Sanchez, M. Iguchi, Takaaki Isono, M. Jiménez, Hideki Kajitani, S. Koczorowski, Norikiyo Koizumi, A. Lo Bue, K. Matsui, Masataka Nakahira, M. Nakamoto, K. Saito, Kazuyuki Sakamoto, T. Sakurai, Tatsuya Shimizu, T. Suwa, K. Takano, Keiya Takebayashi, Nobuhiko Tanaka, F. Tsutsumi, Yasuhiro Uno, Eduard Viladiu Martinez","doi":"10.1088/1741-4326/ad3e8c","DOIUrl":"https://doi.org/10.1088/1741-4326/ad3e8c","url":null,"abstract":"\u0000 In 2023, the manufacturing of all the ITER TF coils has been completed 15 years after the sign of procurement arrangements in 2008. This paper has been jointly submitted by F4E and QST to recollect the lessons learnt in the production of the two parties along the last 15 years.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"54 27","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140701135","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 : 2024-04-15DOI: 10.1088/1741-4326/ad3e8d
Young-Hoon Lee, Jungpyo Lee
The effects of poloidally inhomogeneous heat sources are investigated through a gyrokinetic formula in collisionless toroidal plasmas. The gyrokinetic dispersion relation is newly derived assuming that equilibrium parallel heat flows are generated to remove the injected poloidally nonuniform heat source. The dispersion relation is numerically solved using both inboard and outboard heat sources. For the inboard source injection, both Stringer spin-up (SSU) and geodesic acoustic mode (GAM) can be driven. For the case of outboard source injection, a heat source-driven geodesic acoustic mode (so-called, Q-GAM) is newly found, which features a frequency around half of the standard GAM frequency. It originates from a strongly damped Landau pole when there is no source injection. However, once the heat source intensity is larger than a certain threshold, it becomes unstable while maintaining its frequency. The parametric dependencies of the Q-GAM frequency and source threshold are carried out, and an empirical equation for the source, the threshold is also derived. The Q-GAM frequency is similar to that of EGAM driven by the energetic particles because the main driving terms have the similar structures in poloidal, radial, and parallel velocity coordinates, giving the similar response function of the perturbed parallel pressure to the potential.
{"title":"Destabilization of geodesic acoustic-like mode in the presence of a poloidally inhomogeneous heat sources in tokamak plasmas","authors":"Young-Hoon Lee, Jungpyo Lee","doi":"10.1088/1741-4326/ad3e8d","DOIUrl":"https://doi.org/10.1088/1741-4326/ad3e8d","url":null,"abstract":"\u0000 The effects of poloidally inhomogeneous heat sources are investigated through a gyrokinetic formula in collisionless toroidal plasmas. The gyrokinetic dispersion relation is newly derived assuming that equilibrium parallel heat flows are generated to remove the injected poloidally nonuniform heat source. The dispersion relation is numerically solved using both inboard and outboard heat sources. For the inboard source injection, both Stringer spin-up (SSU) and geodesic acoustic mode (GAM) can be driven. For the case of outboard source injection, a heat source-driven geodesic acoustic mode (so-called, Q-GAM) is newly found, which features a frequency around half of the standard GAM frequency. It originates from a strongly damped Landau pole when there is no source injection. However, once the heat source intensity is larger than a certain threshold, it becomes unstable while maintaining its frequency. The parametric dependencies of the Q-GAM frequency and source threshold are carried out, and an empirical equation for the source, the threshold is also derived. The Q-GAM frequency is similar to that of EGAM driven by the energetic particles because the main driving terms have the similar structures in poloidal, radial, and parallel velocity coordinates, giving the similar response function of the perturbed parallel pressure to the potential.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"278 2","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-04-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140704154","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 : 2024-04-12DOI: 10.1088/1741-4326/ad3e15
Yanmin Liu, T. Long, P. H. Diamond, R. Ke, Zhipeng Chen, Xin Xu, Wenjing Tian, Rongjie Hong, M. Cao, Min Xu, Lu Wang, Zhoujun Yang, Jinbang Yuan, Yongkang Zhou, Qinghao Yan, Qinghu Yang, C. Shen, Lin Nie, Zhanhui Wang, G. Hao, Nengchao Wang, Z. Chen, Jiquan Li, Wei Chen, Wulyu Zhong
Experimental studies of the dynamics of shear flow and turbulence spreading at the edge of tokamak plasmas are reported. Scans of line-averaged density and plasma current are carried out while approaching the Greenwald density limit on the J-TEXT tokamak. In all scans, when the Greenwald fraction f_G=n ̅/n_G=n ̅/(I_p/πa^2) increases, a common feature of enhanced turbulence spreading and edge cooling is found. The result suggests that turbulence spreading is a good indicator of edge cooling, indeed better than turbulent particle transport is. The normalized turbulence spreading power increases significantly when the normalized E×B shearing rate decreases. This indicates that turbulence spreading becomes prominent when the shearing rate is weaker than the turbulence scattering rate. The asymmetry between positive/negative (blobs/holes) spreading events, turbulence spreading power and shear flow are discussed. These results elucidate the important effects of interaction between shear flow and turbulence spreading on plasma edge cooling.
{"title":"The role of shear flow collapse and enhanced turbulence spreading in edge cooling approaching the density limit","authors":"Yanmin Liu, T. Long, P. H. Diamond, R. Ke, Zhipeng Chen, Xin Xu, Wenjing Tian, Rongjie Hong, M. Cao, Min Xu, Lu Wang, Zhoujun Yang, Jinbang Yuan, Yongkang Zhou, Qinghao Yan, Qinghu Yang, C. Shen, Lin Nie, Zhanhui Wang, G. Hao, Nengchao Wang, Z. Chen, Jiquan Li, Wei Chen, Wulyu Zhong","doi":"10.1088/1741-4326/ad3e15","DOIUrl":"https://doi.org/10.1088/1741-4326/ad3e15","url":null,"abstract":"\u0000 Experimental studies of the dynamics of shear flow and turbulence spreading at the edge of tokamak plasmas are reported. Scans of line-averaged density and plasma current are carried out while approaching the Greenwald density limit on the J-TEXT tokamak. In all scans, when the Greenwald fraction f_G=n ̅/n_G=n ̅/(I_p/πa^2) increases, a common feature of enhanced turbulence spreading and edge cooling is found. The result suggests that turbulence spreading is a good indicator of edge cooling, indeed better than turbulent particle transport is. The normalized turbulence spreading power increases significantly when the normalized E×B shearing rate decreases. This indicates that turbulence spreading becomes prominent when the shearing rate is weaker than the turbulence scattering rate. The asymmetry between positive/negative (blobs/holes) spreading events, turbulence spreading power and shear flow are discussed. These results elucidate the important effects of interaction between shear flow and turbulence spreading on plasma edge cooling.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"19 12","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-04-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140710209","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 : 2024-04-12DOI: 10.1088/1741-4326/ad3e14
Yongkyoon In, Hyungho Lee, Kimin Kim, Alberto Loarte, Inhwan Choi, Jinyoung Heo, Yoon Seong Han, Wonho Choe, Junghoo Hwang, Haewon Shin
Edge-localized-modes (ELMs) suppression by non-axisymmetric resonant-magnetic-perturbation (RMP) provides the way to reach high performance fusion plasmas without a threatening level of transient heat fluxes to the walls of fusion devices. The application of RMP, however, strongly modifies the heat flux pattern onto in-vessel components in contact with the plasma (especially the divertor) leading to local “hot spots”. Radiative dissipation by partially ionized species (impurities and deuterium) lowers the heat flux peaks on the walls but has been poorly compatible with such RMP-driven, ELM-crash-suppression. Here, we show how KSTAR has radiatively tamed divertor thermal loading down to more than a factor of 7 in the off-separatrix region without losing ELM-crash-suppression using ITER-like, three-row, RMP configurations, demonstrating its sustainment even in a partially detached plasma in the outer strike point, as desired for ITER.
{"title":"Radiatively tamed divertor thermal loading in Resonant Magnetic Perturbation(RMP)-driven, ELM-crash-suppressed plasmas","authors":"Yongkyoon In, Hyungho Lee, Kimin Kim, Alberto Loarte, Inhwan Choi, Jinyoung Heo, Yoon Seong Han, Wonho Choe, Junghoo Hwang, Haewon Shin","doi":"10.1088/1741-4326/ad3e14","DOIUrl":"https://doi.org/10.1088/1741-4326/ad3e14","url":null,"abstract":"\u0000 Edge-localized-modes (ELMs) suppression by non-axisymmetric resonant-magnetic-perturbation (RMP) provides the way to reach high performance fusion plasmas without a threatening level of transient heat fluxes to the walls of fusion devices. The application of RMP, however, strongly modifies the heat flux pattern onto in-vessel components in contact with the plasma (especially the divertor) leading to local “hot spots”. Radiative dissipation by partially ionized species (impurities and deuterium) lowers the heat flux peaks on the walls but has been poorly compatible with such RMP-driven, ELM-crash-suppression. Here, we show how KSTAR has radiatively tamed divertor thermal loading down to more than a factor of 7 in the off-separatrix region without losing ELM-crash-suppression using ITER-like, three-row, RMP configurations, demonstrating its sustainment even in a partially detached plasma in the outer strike point, as desired for ITER.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"25 56","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-04-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140711553","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 : 2024-04-12DOI: 10.1088/1741-4326/ad3e12
C. Shen, W. Zheng, B. Guo, Y. Ding, Dalong Chen, X. Ai, F. Xue, Y. Zhong, Nengchao Wang, Biao Shen, Bing-biao Xiao, Z. Chen, Yuan Pan
The high acquisition cost and the significant demand for disruptive discharges for data-driven disruption prediction models in future tokamaks pose an inherent contradiction in disruption prediction research. In this paper, we demonstrated a novel approach to predict disruption in a future tokamak using only a few discharges. The approach aims to predict disruption by finding a feature space that is universal to all tokamak. The first step is to use the existing understanding of physics to extract physics-guided features from the diagnostic signals of each tokamak, called physics-guided feature extraction (PGFE). The second step is to align a few data from the future tokamak (target domain) and a large amount of data from existing tokamak (source domain) based on a domain adaptation algorithm called CORrelation ALignment (CORAL). It is the first attempt at applying domain adaptation in the task of cross-tokamak disruption prediction. PGFE has been successfully applied in J-TEXT to predict disruption with excellent performance. PGFE can also reduce the data volume requirements due to extracting the less device-specific features, thereby establishing a solid foundation for cross-tokamak disruption prediction. We have further improved CORAL (supervised CORAL, S-CORAL) to enhance its appropriateness in feature alignment for the disruption prediction task. To simulate the existing and future tokamak case, we selected J-TEXT as the existing tokamak and EAST as the future tokamak, which has a large gap in the ranges of plasma parameters. The utilization of the S-CORAL improves the disruption prediction performance on future tokamak. Through interpretable analysis, we discovered that the learned knowledge of the disruption prediction model through this approach exhibits more similarities to the model trained on large data volumes of future tokamak. This approach provides a light, interpretable and few data-required way by aligning features to predict disruption using small data volume from the future tokamak.
{"title":"Cross-tokamak Disruption Prediction based on domain adaptation","authors":"C. Shen, W. Zheng, B. Guo, Y. Ding, Dalong Chen, X. Ai, F. Xue, Y. Zhong, Nengchao Wang, Biao Shen, Bing-biao Xiao, Z. Chen, Yuan Pan","doi":"10.1088/1741-4326/ad3e12","DOIUrl":"https://doi.org/10.1088/1741-4326/ad3e12","url":null,"abstract":"\u0000 The high acquisition cost and the significant demand for disruptive discharges for data-driven disruption prediction models in future tokamaks pose an inherent contradiction in disruption prediction research. In this paper, we demonstrated a novel approach to predict disruption in a future tokamak using only a few discharges. The approach aims to predict disruption by finding a feature space that is universal to all tokamak. The first step is to use the existing understanding of physics to extract physics-guided features from the diagnostic signals of each tokamak, called physics-guided feature extraction (PGFE). The second step is to align a few data from the future tokamak (target domain) and a large amount of data from existing tokamak (source domain) based on a domain adaptation algorithm called CORrelation ALignment (CORAL). It is the first attempt at applying domain adaptation in the task of cross-tokamak disruption prediction. PGFE has been successfully applied in J-TEXT to predict disruption with excellent performance. PGFE can also reduce the data volume requirements due to extracting the less device-specific features, thereby establishing a solid foundation for cross-tokamak disruption prediction. We have further improved CORAL (supervised CORAL, S-CORAL) to enhance its appropriateness in feature alignment for the disruption prediction task. To simulate the existing and future tokamak case, we selected J-TEXT as the existing tokamak and EAST as the future tokamak, which has a large gap in the ranges of plasma parameters. The utilization of the S-CORAL improves the disruption prediction performance on future tokamak. Through interpretable analysis, we discovered that the learned knowledge of the disruption prediction model through this approach exhibits more similarities to the model trained on large data volumes of future tokamak. This approach provides a light, interpretable and few data-required way by aligning features to predict disruption using small data volume from the future tokamak.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"11 20","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-04-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140710538","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 : 2024-04-12DOI: 10.1088/1741-4326/ad3e16
C. Maggi
In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements – new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14MeV neutron yield monitors – as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in [1], while JET experience on T and D-T operations is presented in [2].
2021 年，JET 利用其独特的能力，使用 T 和 D-T 燃料以及类似于热核实验堆的 Be/W 壁（JET-ILW）进行运行。这是继 1997 年 DTE1 之后的第二次大型 JET D-T 活动（DTE2），是 JET 一系列改进措施（新的聚变诊断、新的 T 注入能力、翻新 T 设备、增加辅助加热、舱内校准 14MeV 中子产率监测器）以及聚变界在等离子体理论和建模方面取得的重大进展的结晶。作为 DTE2 的补充，还开展了一系列同位素物理活动，包括在高 T-NBI 功率下的纯氚运行。为确保氚的安全运行，新的 T 和 D-T 实验使用了 1 千克 T（与 DTE1 中的 100 克相比），产生了迄今为止与聚变反应堆最相关的 D-T 等离子体，拓展了我们对同位素和 D-T 混合物物理学的理解。此外，由于 JET T 和 DTE2 试验是在上一次大型 D-T 托卡马克实验近 25 年后进行的，因此欧洲聚变计划的战略目标也是刷新核托卡马克的运行经验，为国际热核实验堆的运行做好准备。本文报告了在欧洲聚变 JET1 工作包内进行的 JET T 和 DTE2 实验的主要物理结果。由EUROfusion（在地平线2020框架计划内启动，并在地平线欧洲FP下继续进行）开展的JET D-T运行的技术利用、核代码的开发和验证、ITER运行的中子工具和技术等方面的进展在[1]中报告，而JET在T和D-T运行方面的经验在[2]中介绍。
{"title":"Overview of T and D-T results in JET with ITER-like Wall","authors":"C. Maggi","doi":"10.1088/1741-4326/ad3e16","DOIUrl":"https://doi.org/10.1088/1741-4326/ad3e16","url":null,"abstract":"\u0000 In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements – new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14MeV neutron yield monitors – as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in [1], while JET experience on T and D-T operations is presented in [2].","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"11 9","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-04-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140712127","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 : 2024-04-11DOI: 10.1088/1741-4326/ad3d6d
Akihiro Ishizawa, Yasuaki Kishimoto, K. Imadera, Yuji Nakamura, S. Maeyama
A higher plasma $beta$ is desirable for realizing high performance fusion reactor, in fact, one of the three goals of JT-60SA project is to achieve a high-$beta$ regime. We investigate key physical processes that regulate the $beta$ dependence of turbulent transport in L-mode plasmas by means of both local and global gyrokinetic simulations. From local simulations, we found that the turbulent transport does not decrease as $beta$ increases, because the electromagnetic stabilizing effect is canceled out by the increase of the Shafranov shift. This influence of the Shafranov shift is suppressed when the magnetic shear is weak, and thus the electromagnetic stabilization is prominent in weak shear plasmas, suggesting an advantage of weak magnetic shear plasmas for achieving a high-$beta$ regime. In high $beta$ regime, local gyrokinetic simulations are suffered from the non-saturation of turbulence level. In global simulations, by contrast, the electromagnetic turbulence gets saturated by the entropy advection in the radial direction to avoid the zonal flow erosion due to magnetic fluctuations. This breakthrough enables us to explore turbulent transport at a higher $beta$ regime by gyrokinetic simulations.
事实上，JT-60SA 项目的三大目标之一就是实现高贝塔度的等离子体。我们通过局部和全局陀螺动力学模拟，研究了调节 L 模式等离子体中湍流输运的 $beta$ 依赖性的关键物理过程。通过局部模拟，我们发现湍流输运并不会随着$beta$的增加而减少，因为电磁稳定效应被Shafranov偏移的增加所抵消。当磁剪切较弱时，沙弗拉诺夫偏移的影响会被抑制，因此电磁稳定作用在弱剪切等离子体中非常突出，这表明弱磁剪切等离子体在实现高$beta$制度方面具有优势。在高$beta$系统中，局部陀螺动力学模拟由于湍流水平不饱和而受到影响。相比之下，在全局模拟中，电磁湍流会因径向的熵平流而饱和，以避免磁波动造成的带状流侵蚀。这一突破使我们能够通过陀螺动力学模拟探索更高$beta$机制下的湍流输运。
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Pub Date : 2024-04-11DOI: 10.1088/1741-4326/ad3d6e
Masakatsu Fukumoto, Qilin Yue, K. Hanada, Shinichiro Kojima, Tomohide Nakano, N. Y. Yoshida, R. Ikezoe, Yoshihiko Nagashima, Takeshi Ido, T. Onchi, H. Idei, Hiroki Iguchi, Takumi Komiyama, T. Shikama, A. Ejiri, S. Masuzaki, Mizuki Sakamoto, Yoshio Ueda, K. Kuroda, K. Kono, S. Shimabukuro, A. Higashijima
Electron Cyclotron Wall Conditioning with Neon gas (Ne-ECWC) has been performed on the normal conducting spherial tokamak QUEST with metal walls under a trapped particle configuration with mixed O and X-mode polarization EC waves with a frequency of 8.2 GHz and an injection power of 16 kW. The Ne-ECWC removes hydrogen from the wall with small neon retention. The Ne-ECWC decreases hydrogen recycling at the following tokamak discharges, contributing to an improvement of the following tokamak plasma start-up: the plasma current increases and the start-up timing of the plasma current shifts forward. However, defects such as voids and bubbles are formed on tungsten surface exposed to the Ne-ECWC plasma.
使用氖气的电子回旋加速器壁调节（Ne-ECWC）是在正常导电球形托卡马克 QUEST 上进行的，该托卡马克具有金属壁，在捕获粒子配置下具有混合 O 型和 X 型极化 EC 波，频率为 8.2 千兆赫，注入功率为 16 千瓦。氖-ECWC 从壁上去除氢气，但氖的残留量很小。Ne-ECWC 减少了后续托卡马克放电中的氢回收，有助于改善后续托卡马克等离子体的启动：等离子体电流增加，等离子体电流的启动时间提前。然而，暴露于 Ne-ECWC 等离子体的钨表面会形成空洞和气泡等缺陷。
{"title":"Hydrogen removal by electron cyclotron wall conditioning with neon gas and its impact of tokamak plasma start-up on the QUEST spherical tokamak","authors":"Masakatsu Fukumoto, Qilin Yue, K. Hanada, Shinichiro Kojima, Tomohide Nakano, N. Y. Yoshida, R. Ikezoe, Yoshihiko Nagashima, Takeshi Ido, T. Onchi, H. Idei, Hiroki Iguchi, Takumi Komiyama, T. Shikama, A. Ejiri, S. Masuzaki, Mizuki Sakamoto, Yoshio Ueda, K. Kuroda, K. Kono, S. Shimabukuro, A. Higashijima","doi":"10.1088/1741-4326/ad3d6e","DOIUrl":"https://doi.org/10.1088/1741-4326/ad3d6e","url":null,"abstract":"\u0000 Electron Cyclotron Wall Conditioning with Neon gas (Ne-ECWC) has been performed on the normal conducting spherial tokamak QUEST with metal walls under a trapped particle configuration with mixed O and X-mode polarization EC waves with a frequency of 8.2 GHz and an injection power of 16 kW. The Ne-ECWC removes hydrogen from the wall with small neon retention. The Ne-ECWC decreases hydrogen recycling at the following tokamak discharges, contributing to an improvement of the following tokamak plasma start-up: the plasma current increases and the start-up timing of the plasma current shifts forward. However, defects such as voids and bubbles are formed on tungsten surface exposed to the Ne-ECWC plasma.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"24 11","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-04-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140713270","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 : 2024-04-09DOI: 10.1088/1741-4326/ad3c52
V. Izzo, A. Battey, R. Tinguely, R. Sweeney, C. Hansen
Extended-MHD modeling of planned runaway electron mitigation coils (REMC) for SPARC and DIII-D is performed with the NIMROD code. A coil has been designed for each machine, with the two differing in shape and location, but both having n=1 symmetry (with n the toroidal mode number). Compared to previous modeling efforts, three improvements are made to the simulations boundary conditions. First a resistive wall model is used in place of an ideal wall. Second, the ThinCurr code is used to compute the time-dependent 3D fields used as magnetic boundary conditions for the simulations. Third, the simulation boundary is moved from the the first-wall location to the vacuum vessel (VV), which extends the boundary past the location of the internal REMC. To remove the 3D coil from the simulation domain, an equivalent set of 3D fields is calculated at the VV boundary that produce approximately the same field distribution at the last closed flux surface assuming vacuum between the two. Each of these three boundary condition improvements leads to an improvement in the predicted performance of the REMC for both machines. The resistive wall alone primarily effects the resonance of the coil with the plasma after the TQ, affecting the q-profile evolution in the SPARC modeling, and allowing the applied spectrum to be modified in response to the plasma in the DIII-D modeling. The movement of the simulation boundary has the most significant effect on the RE confinement overall, including in the early stages, particularly for a DIII-D inner wall limited equilibrium, where the RE loss fraction increases from 90% to >99%, with SPARC RE losses also occurring much earlier when the boundary is placed at the VV.
{"title":"Boundary condition effects on Runaway Electron Mitigation Coil modeling for the SPARC and DIII-D tokamaks","authors":"V. Izzo, A. Battey, R. Tinguely, R. Sweeney, C. Hansen","doi":"10.1088/1741-4326/ad3c52","DOIUrl":"https://doi.org/10.1088/1741-4326/ad3c52","url":null,"abstract":"\u0000 Extended-MHD modeling of planned runaway electron mitigation coils (REMC) for SPARC and DIII-D is performed with the NIMROD code. A coil has been designed for each machine, with the two differing in shape and location, but both having n=1 symmetry (with n the toroidal mode number). Compared to previous modeling efforts, three improvements are made to the simulations boundary conditions. First a resistive wall model is used in place of an ideal wall. Second, the ThinCurr code is used to compute the time-dependent 3D fields used as magnetic boundary conditions for the simulations. Third, the simulation boundary is moved from the the first-wall location to the vacuum vessel (VV), which extends the boundary past the location of the internal REMC. To remove the 3D coil from the simulation domain, an equivalent set of 3D fields is calculated at the VV boundary that produce approximately the same field distribution at the last closed flux surface assuming vacuum between the two. Each of these three boundary condition improvements leads to an improvement in the predicted performance of the REMC for both machines. The resistive wall alone primarily effects the resonance of the coil with the plasma after the TQ, affecting the q-profile evolution in the SPARC modeling, and allowing the applied spectrum to be modified in response to the plasma in the DIII-D modeling. The movement of the simulation boundary has the most significant effect on the RE confinement overall, including in the early stages, particularly for a DIII-D inner wall limited equilibrium, where the RE loss fraction increases from 90% to >99%, with SPARC RE losses also occurring much earlier when the boundary is placed at the VV.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"98 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"140725919","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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