Pub Date : 2024-08-11DOI: 10.1088/1741-4326/ad6913
Xiaoru Chen, Yuanyuan Zhang, Liuming Wei, Qirong Zheng, Chuanguo Zhang and Yonggang Li
Hydrogen (H) retention and desorption in tungsten (W)-based plasma-facing materials are still not well understood, largely due to the limitations of ex-situ observations in experimental detection methods like thermal desorption spectroscopy (TDS). In order to reveal the fundamental mechanisms behind H retention and desorption, we developed a cluster dynamics model, IRadMat-TDS, for theoretical modeling of depth distribution and TDS of deuterium (D) in polycrystalline W. The model newly includes the saturated absorption and emission of D in inherent sinks like grain boundaries (GBs), as well as the multi-trapping effect of D in various types of GBs with different trapping energies. The simulated TDS spectra are in agreement with experimental ones. For polycrystalline W under D ion irradiation within keV-energy range, two typical thermal desorption peaks in TDS at around 490 and 550 K are explicitly attributed to D emission from GBs and vacancies, respectively. And GBs play a major role in D retention. Moreover, the broad peaks in TDS come from the convolution of multi-trapping of D in sinks with different types of trapping sites rather than a single-site approximation.
钨（W）基等离子体面材料中氢（H）的保留和解吸仍未得到很好的理解，这主要是由于热解吸光谱（TDS）等实验检测方法的原位观测存在局限性。为了揭示 H 保留和解吸背后的基本机制，我们开发了一个群集动力学模型 IRadMat-TDS，用于对多晶 W 中氘（D）的深度分布和 TDS 进行理论建模。该模型新加入了 D 在晶界（GB）等固有汇中的饱和吸收和发射，以及 D 在各种类型 GB 中不同捕获能量的多重捕获效应。模拟的 TDS 光谱与实验光谱一致。对于在 keV 能量范围内接受 D 离子辐照的多晶 W，TDS 在 490 和 550 K 左右的两个典型热解吸峰分别明确归因于 GBs 和空位的 D 发射。而 GB 在 D 的保留中起着主要作用。此外，TDS 中的宽峰来自 D 在具有不同类型捕获位点的汇中的多捕获卷积，而不是单位点近似。
{"title":"Cluster dynamics modeling of hydrogen retention and desorption in tungsten with saturation and multi-trapping effect of sinks","authors":"Xiaoru Chen, Yuanyuan Zhang, Liuming Wei, Qirong Zheng, Chuanguo Zhang and Yonggang Li","doi":"10.1088/1741-4326/ad6913","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6913","url":null,"abstract":"Hydrogen (H) retention and desorption in tungsten (W)-based plasma-facing materials are still not well understood, largely due to the limitations of ex-situ observations in experimental detection methods like thermal desorption spectroscopy (TDS). In order to reveal the fundamental mechanisms behind H retention and desorption, we developed a cluster dynamics model, IRadMat-TDS, for theoretical modeling of depth distribution and TDS of deuterium (D) in polycrystalline W. The model newly includes the saturated absorption and emission of D in inherent sinks like grain boundaries (GBs), as well as the multi-trapping effect of D in various types of GBs with different trapping energies. The simulated TDS spectra are in agreement with experimental ones. For polycrystalline W under D ion irradiation within keV-energy range, two typical thermal desorption peaks in TDS at around 490 and 550 K are explicitly attributed to D emission from GBs and vacancies, respectively. And GBs play a major role in D retention. Moreover, the broad peaks in TDS come from the convolution of multi-trapping of D in sinks with different types of trapping sites rather than a single-site approximation.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"85 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141945545","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-11DOI: 10.1088/1741-4326/ad6675
Felix Wilms, Alejandro Bañón Navarro, Thomas Windisch, Sergey Bozhenkov, Felix Warmer, Golo Fuchert, Oliver Ford, Daihong Zhang, Torsten Stange, Frank Jenko and the W7-X Team
We present the first nonlinear, gyrokinetic, radially global simulation of a discharge of the Wendelstein 7-X-like stellarator, including kinetic electrons, an equilibrium radial electric field, as well as electromagnetic and collisional effects. By comparison against flux-tube and full-flux-surface simulations, we assess the impact of the equilibrium ExB-flow and flow shear on the stabilisation of turbulence. In contrast to the existing literature, we further provide substantial evidence for the turbulent electron heat flux being driven by trapped-electron-mode and electron-temperature-gradient turbulence in the core of the plasma. The former manifests as a hybrid together with ion-temperature-gradient turbulence and is primarily driven by the finite electron temperature gradient, which has largely been neglected in nonlinear stellarator simulations presented in the existing literature.
{"title":"Global gyrokinetic analysis of Wendelstein 7-X discharge: unveiling the importance of trapped-electron-mode and electron-temperature-gradient turbulence","authors":"Felix Wilms, Alejandro Bañón Navarro, Thomas Windisch, Sergey Bozhenkov, Felix Warmer, Golo Fuchert, Oliver Ford, Daihong Zhang, Torsten Stange, Frank Jenko and the W7-X Team","doi":"10.1088/1741-4326/ad6675","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6675","url":null,"abstract":"We present the first nonlinear, gyrokinetic, radially global simulation of a discharge of the Wendelstein 7-X-like stellarator, including kinetic electrons, an equilibrium radial electric field, as well as electromagnetic and collisional effects. By comparison against flux-tube and full-flux-surface simulations, we assess the impact of the equilibrium ExB-flow and flow shear on the stabilisation of turbulence. In contrast to the existing literature, we further provide substantial evidence for the turbulent electron heat flux being driven by trapped-electron-mode and electron-temperature-gradient turbulence in the core of the plasma. The former manifests as a hybrid together with ion-temperature-gradient turbulence and is primarily driven by the finite electron temperature gradient, which has largely been neglected in nonlinear stellarator simulations presented in the existing literature.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"46 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141945546","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-07DOI: 10.1088/1741-4326/ad6885
A. LeViness, S.A. Lazerson, A. Jansen van Vuuren, J. Rueda-Rueda, M. Beurskens, S. Bozhenkov, K.J. Brunner, O.P. Ford, G. Fuchert, M. Garcìa-Muñoz, M. Isobe, C. Killer, J. Knauer, K. Ogawa, N. Pablant, E. Pasch, P. Poloskei, T. Romba and the W7-X Team
We present the first validated synthetic diagnostic for fast ion loss detectors (FILDs) in the Wendelstein 7-X (W7-X) stellarator. This model has been developed on, and validated against experimental data from, a FILD provided by the National Institute for Fusion Science (NIFS-FILD), with potential future applicability to the existing Faraday Cup FILD (FC-FILD) on W7-X as well as the scintillating FILD (S-FILD) currently under development. A workflow combining Monte Carlo codes BEAMS3D and ASCOT5 is used to track fast ions produced by neutral beam injection from the moment of ionization until they are thermalized or lost from the last closed flux surface, and from there to a virtual plane which serves as a projection of the entrance aperture to the FILD. Simulations in ASCOT5 are analyzed via a geometric method to determine the probability of transmission through the FILD aperture and onto the detector as a function of normalized momentum, pitch angle, gyrophase, and position at the virtual plane. This probability is then applied to the simulated ions arriving from the plasma, producing a simulated signal from a computationally tractable number of simulated fast ions. Simulated signals are presented for two W7-X experiments with neutral beam injection and quantitatively compared with experimental measurements from the NIFS-FILD diagnostic. An estimate of the frequency of charge-exchange with neutral particles in the edge is performed, and it is found that this process may have a significant impact on the measured signals.
{"title":"Validation of a synthetic fast ion loss detector model for Wendelstein 7-X","authors":"A. LeViness, S.A. Lazerson, A. Jansen van Vuuren, J. Rueda-Rueda, M. Beurskens, S. Bozhenkov, K.J. Brunner, O.P. Ford, G. Fuchert, M. Garcìa-Muñoz, M. Isobe, C. Killer, J. Knauer, K. Ogawa, N. Pablant, E. Pasch, P. Poloskei, T. Romba and the W7-X Team","doi":"10.1088/1741-4326/ad6885","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6885","url":null,"abstract":"We present the first validated synthetic diagnostic for fast ion loss detectors (FILDs) in the Wendelstein 7-X (W7-X) stellarator. This model has been developed on, and validated against experimental data from, a FILD provided by the National Institute for Fusion Science (NIFS-FILD), with potential future applicability to the existing Faraday Cup FILD (FC-FILD) on W7-X as well as the scintillating FILD (S-FILD) currently under development. A workflow combining Monte Carlo codes BEAMS3D and ASCOT5 is used to track fast ions produced by neutral beam injection from the moment of ionization until they are thermalized or lost from the last closed flux surface, and from there to a virtual plane which serves as a projection of the entrance aperture to the FILD. Simulations in ASCOT5 are analyzed via a geometric method to determine the probability of transmission through the FILD aperture and onto the detector as a function of normalized momentum, pitch angle, gyrophase, and position at the virtual plane. This probability is then applied to the simulated ions arriving from the plasma, producing a simulated signal from a computationally tractable number of simulated fast ions. Simulated signals are presented for two W7-X experiments with neutral beam injection and quantitatively compared with experimental measurements from the NIFS-FILD diagnostic. An estimate of the frequency of charge-exchange with neutral particles in the edge is performed, and it is found that this process may have a significant impact on the measured signals.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"93 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141945548","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-08-01DOI: 10.1088/1741-4326/ad5a1e
Someswar Dutta, Deepti Sharma, R.L. Tanna, J. Ghosh and D. Raju
Runaway electron (RE) deconfinement and subsequent suppression is of prime importance for successful long-term operation of any tokamak. In this work, to deconfine and mitigate REs, the efficacy of local vertical field (LVF) perturbation has been explored numerically. LVF perturbation-assisted RE loss studies are carried out by simulating the drift orbits of the REs in magnetostatic perturbed fields and estimating the resulting orbit losses for different initial energies and magnitudes of LVF perturbation. To this end, the pre-existing PARTICLE code has been extended to the relativistic full-orbit-following code PARTICLE-3D (P3D) integrated with the magnetic field calculation code EFFI and plasma equilibrium field calculation code IPREQ to include the required fields for studying particle dynamics in general; this is then used to numerically model LVF perturbation-assisted RE deconfinement experiments conducted in the ADITYA tokamak. Simulation results show a significant (∼90%) deconfinement of REs with the application of LVF perturbation of a suitable amplitude (∼0.1% of the total magnetic field) in a preferred direction. The existence of a threshold magnitude of the applied field is also established, which is observed to be dependent on the energy of the REs. The simulation results reproduce all the experimental observations and reveal other interesting features of RE mitigation using LVF perturbation. The temporal map of orbiting time of REs shows that REs originating from the inboard side edge region ( N > 0.5) of the plasma are relatively more prone to be lost with the application of suitable LVF perturbation than those originating from the plasma core. Interestingly, the simulation results demonstrate the existence of strong correlation between the safety factor (q) profile in the plasma edge region ( N > 0.7) and the level of RE deconfinement using LVF perturbation.
失控电子（RE）的去惑及随后的抑制对于任何托卡马克的长期成功运行都是至关重要的。在这项工作中，为了消解和缓解失控电子，对局部垂直场（LVF）扰动的功效进行了数值探索。LVF 扰动辅助 RE 损耗研究是通过模拟 RE 在磁静力扰动场中的漂移轨道，并估算不同初始能量和 LVF 扰动幅度下产生的轨道损耗来进行的。为此，已存在的 PARTICLE 代码被扩展为相对论全轨道跟踪代码 PARTICLE-3D (P3D)，并与磁场计算代码 EFFI 和等离子体平衡场计算代码 IPREQ 集成，以包括研究一般粒子动力学所需的场。仿真结果表明，在首选方向上施加适当振幅（总磁场的0.1%）的LVF扰动，可显著（∼90%）去除RE。此外，还确定了外加磁场的阈值大小，据观察，该阈值大小取决于 REs 的能量。模拟结果再现了所有的实验观测结果，并揭示了利用 LVF 扰动缓解 RE 的其他有趣特征。REs轨道时间图显示，与来自等离子体核心的REs相比，来自等离子体内侧边缘区域（N > 0.5）的REs在应用适当的LVF扰动后更容易消失。有趣的是，模拟结果表明，等离子体边缘区域（N > 0.7）的安全系数（q）曲线与使用 LVF 扰动的 RE 消能水平之间存在很强的相关性。
{"title":"Deconfinement of runaway electrons by local vertical magnetic field perturbation","authors":"Someswar Dutta, Deepti Sharma, R.L. Tanna, J. Ghosh and D. Raju","doi":"10.1088/1741-4326/ad5a1e","DOIUrl":"https://doi.org/10.1088/1741-4326/ad5a1e","url":null,"abstract":"Runaway electron (RE) deconfinement and subsequent suppression is of prime importance for successful long-term operation of any tokamak. In this work, to deconfine and mitigate REs, the efficacy of local vertical field (LVF) perturbation has been explored numerically. LVF perturbation-assisted RE loss studies are carried out by simulating the drift orbits of the REs in magnetostatic perturbed fields and estimating the resulting orbit losses for different initial energies and magnitudes of LVF perturbation. To this end, the pre-existing PARTICLE code has been extended to the relativistic full-orbit-following code PARTICLE-3D (P3D) integrated with the magnetic field calculation code EFFI and plasma equilibrium field calculation code IPREQ to include the required fields for studying particle dynamics in general; this is then used to numerically model LVF perturbation-assisted RE deconfinement experiments conducted in the ADITYA tokamak. Simulation results show a significant (∼90%) deconfinement of REs with the application of LVF perturbation of a suitable amplitude (∼0.1% of the total magnetic field) in a preferred direction. The existence of a threshold magnitude of the applied field is also established, which is observed to be dependent on the energy of the REs. The simulation results reproduce all the experimental observations and reveal other interesting features of RE mitigation using LVF perturbation. The temporal map of orbiting time of REs shows that REs originating from the inboard side edge region ( N > 0.5) of the plasma are relatively more prone to be lost with the application of suitable LVF perturbation than those originating from the plasma core. Interestingly, the simulation results demonstrate the existence of strong correlation between the safety factor (q) profile in the plasma edge region ( N > 0.7) and the level of RE deconfinement using LVF perturbation.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"219 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141885827","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-25DOI: 10.1088/1741-4326/ad5b04
J.M. Chen, P.H. Wang, Y. Zhou, Q. Li, J.L. Li, J. Wu, J. Du, B. Yang, X.B. Zhu, H. Gao, D. Hu, Y.Y. Chen, Q.M. Wang, Z.H. Liu, K. Wang, S. Liu, L.M. Bao and R. Hunt
ITER’s enhanced heat flux (EHF) first wall (FW) provides thermal shielding to the other components behind it and limits its plasma boundary, consequently bearing high surface heat loads up to 4.7 MW m−2. China developed the EHF FW technologies in steady stages by making small mock-ups, semi-prototypes and a full-scale prototype (FSP), working toward series production at the Southwestern Institute of Physics. All the key technologies for bimetallic diffusion bonding, welding and assembly have been fully qualified. The Be armor tile size effects on thermal-fatigue performance have been evaluated using a pulsed high heat flux test with an EMS-400 electron beam facility. It was found that both the Be/CuCrZr bonding and its thermal-fatigue life are highly dependent on the tile size, and 12 × 12 mm2 Be tiles are acceptable for the required performance, either as individual tiles or in castellation by cutting larger ones. Small-scale mock-ups with such Be tiles survived 16 000 thermal cycles at 4.7 MW m−2, while EHF FW fingers with 16 × 16 mm2 Be tiles demonstrated unstable performance. In contrast, only BE tiles larger than 24 × 12 mm2 showed reliable diffusion bonding with the CuCrZr heat sink by hot isostatic pressing and consequently should be castellated into smaller ones. The manufacturing of the FSP finger goes through complex thermal cycles, using CuCrZr alloys in age-strengthening plus cold-rolling states, enabling it to maintain its tensile strength to the level of 285 MPa and its mean grain size less than 100 µm. The pipe welding for finger pairing and its assembly with the central beam are demonstrated, including enabling welding to be carried out twice for repair in a narrow space. The FSP showed good finger alignment, dimensional control and reliable vacuum tightness without any blockage of cooling channels.
{"title":"Development and manufacturing of beryllium-armoring ITER enhanced heat flux FW toward series production in China","authors":"J.M. Chen, P.H. Wang, Y. Zhou, Q. Li, J.L. Li, J. Wu, J. Du, B. Yang, X.B. Zhu, H. Gao, D. Hu, Y.Y. Chen, Q.M. Wang, Z.H. Liu, K. Wang, S. Liu, L.M. Bao and R. Hunt","doi":"10.1088/1741-4326/ad5b04","DOIUrl":"https://doi.org/10.1088/1741-4326/ad5b04","url":null,"abstract":"ITER’s enhanced heat flux (EHF) first wall (FW) provides thermal shielding to the other components behind it and limits its plasma boundary, consequently bearing high surface heat loads up to 4.7 MW m−2. China developed the EHF FW technologies in steady stages by making small mock-ups, semi-prototypes and a full-scale prototype (FSP), working toward series production at the Southwestern Institute of Physics. All the key technologies for bimetallic diffusion bonding, welding and assembly have been fully qualified. The Be armor tile size effects on thermal-fatigue performance have been evaluated using a pulsed high heat flux test with an EMS-400 electron beam facility. It was found that both the Be/CuCrZr bonding and its thermal-fatigue life are highly dependent on the tile size, and 12 × 12 mm2 Be tiles are acceptable for the required performance, either as individual tiles or in castellation by cutting larger ones. Small-scale mock-ups with such Be tiles survived 16 000 thermal cycles at 4.7 MW m−2, while EHF FW fingers with 16 × 16 mm2 Be tiles demonstrated unstable performance. In contrast, only BE tiles larger than 24 × 12 mm2 showed reliable diffusion bonding with the CuCrZr heat sink by hot isostatic pressing and consequently should be castellated into smaller ones. The manufacturing of the FSP finger goes through complex thermal cycles, using CuCrZr alloys in age-strengthening plus cold-rolling states, enabling it to maintain its tensile strength to the level of 285 MPa and its mean grain size less than 100 µm. The pipe welding for finger pairing and its assembly with the central beam are demonstrated, including enabling welding to be carried out twice for repair in a narrow space. The FSP showed good finger alignment, dimensional control and reliable vacuum tightness without any blockage of cooling channels.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"36 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141782321","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-25DOI: 10.1088/1741-4326/ad59b5
Y. Zhong, W. Zheng, Z.Y. Chen, W. Yan, F. Xia, L.M. Yu, F.M. Xue, C.S. Shen, X.K. Ai, Z.Y. Yang, Y.L. Yu, Z.S. Nie, Y.H. Ding, Y.F. Liang, Z.P. Chen and J-TEXT Team
Unmitigated disruptions pose a much more serious threat when large-scale tokamaks are operating in the high performance regime. Machine learning based disruption predictors can exhibit impressive performance. However, their effectiveness is based on a substantial amount of training data. In future reactors, obtaining a substantial amount of disruption data in high performance regimes without risking damage to the machine is highly improbable. Using machine learning to develop disruption predictors on data from the low performance regime and transfer them to the high performance regime is an effective solution for a large reactor-sized tokamak like ITER and beyond. In this study, a number of models are trained using different subsets of data from the HL-2A tokamak experiment. A SHapley Additive exPlanations (SHAP) analysis is executed on the models, revealing that there are different, even contradicting, patterns between different performance regimes. Thus, simply mixing data among different performance regimes will not yield optimal results. Based on this analysis, we propose an instance-based transfer learning technique which trains the model using a dataset generated with an optimized strategy. The strategy involves instance and feature selection based on the physics behind differences in high- and low-performance discharges, as revealed by SHAP model analysis. The TrAdaBoost technique significantly improved the model performance from 0.78 BA (balanced accuracy) to 0.86 BA with a few high-performance operation data.
{"title":"High-beta disruption prediction study on HL-2A with instance-based transfer learning","authors":"Y. Zhong, W. Zheng, Z.Y. Chen, W. Yan, F. Xia, L.M. Yu, F.M. Xue, C.S. Shen, X.K. Ai, Z.Y. Yang, Y.L. Yu, Z.S. Nie, Y.H. Ding, Y.F. Liang, Z.P. Chen and J-TEXT Team","doi":"10.1088/1741-4326/ad59b5","DOIUrl":"https://doi.org/10.1088/1741-4326/ad59b5","url":null,"abstract":"Unmitigated disruptions pose a much more serious threat when large-scale tokamaks are operating in the high performance regime. Machine learning based disruption predictors can exhibit impressive performance. However, their effectiveness is based on a substantial amount of training data. In future reactors, obtaining a substantial amount of disruption data in high performance regimes without risking damage to the machine is highly improbable. Using machine learning to develop disruption predictors on data from the low performance regime and transfer them to the high performance regime is an effective solution for a large reactor-sized tokamak like ITER and beyond. In this study, a number of models are trained using different subsets of data from the HL-2A tokamak experiment. A SHapley Additive exPlanations (SHAP) analysis is executed on the models, revealing that there are different, even contradicting, patterns between different performance regimes. Thus, simply mixing data among different performance regimes will not yield optimal results. Based on this analysis, we propose an instance-based transfer learning technique which trains the model using a dataset generated with an optimized strategy. The strategy involves instance and feature selection based on the physics behind differences in high- and low-performance discharges, as revealed by SHAP model analysis. The TrAdaBoost technique significantly improved the model performance from 0.78 BA (balanced accuracy) to 0.86 BA with a few high-performance operation data.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"5 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141782320","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-25DOI: 10.1088/1741-4326/ad5190
Guo Meng, Philipp Lauber, Zhixin Lu, Andreas Bergmann and Mireille Schneider
The spatio-temporal evolution of the energetic particles in the transport time scale in tokamak plasmas is a key issue of the plasmas confinement, especially in burning plasmas. In order to include sources and sinks and collisional slowing down processes, a new solver, ATEP-3D was implemented to simulate the evolution of the energetic particle (EP) distribution in the three-dimensional constants of motion (CoM) space. The Fokker–Planck collision operator represented in the CoM space is derived and numerically calculated. The collision coefficients are averaged over the unperturbed orbits to capture the fundamental properties of EPs. ATEP-3D is fully embedded in ITER IMAS framework and combined with the LIGKA/HAGIS codes. The finite volume method and the implicit Crank-Nicholson scheme are adopted due to their optimal numerical properties for transport time scale studies. ATEP-3D allows the analysis of the particle and power balance with the source and sink during the transport process to evaluate the EP confinement properties.
托卡马克等离子体中高能粒子在传输时间尺度上的时空演变是等离子体约束的一个关键问题，尤其是在燃烧等离子体中。为了包括源和汇以及碰撞减速过程，我们采用了一种新的求解器 ATEP-3D 来模拟三维运动常数（CoM）空间中高能粒子（EP）分布的演变。推导并数值计算了 CoM 空间中的福克-普朗克碰撞算子。碰撞系数取未扰动轨道的平均值，以捕捉 EP 的基本特性。ATEP-3D 完全嵌入了 ITER IMAS 框架，并与 LIGKA/HAGIS 代码相结合。由于有限体积法和隐式 Crank-Nicholson 方案在传输时间尺度研究方面具有最佳数值特性，因此采用了这两种方法。ATEP-3D 允许在传输过程中分析粒子与源和汇的功率平衡，以评估 EP 的约束特性。
{"title":"Energetic particles transport in constants of motion space due to collisions in tokamak plasmas","authors":"Guo Meng, Philipp Lauber, Zhixin Lu, Andreas Bergmann and Mireille Schneider","doi":"10.1088/1741-4326/ad5190","DOIUrl":"https://doi.org/10.1088/1741-4326/ad5190","url":null,"abstract":"The spatio-temporal evolution of the energetic particles in the transport time scale in tokamak plasmas is a key issue of the plasmas confinement, especially in burning plasmas. In order to include sources and sinks and collisional slowing down processes, a new solver, ATEP-3D was implemented to simulate the evolution of the energetic particle (EP) distribution in the three-dimensional constants of motion (CoM) space. The Fokker–Planck collision operator represented in the CoM space is derived and numerically calculated. The collision coefficients are averaged over the unperturbed orbits to capture the fundamental properties of EPs. ATEP-3D is fully embedded in ITER IMAS framework and combined with the LIGKA/HAGIS codes. The finite volume method and the implicit Crank-Nicholson scheme are adopted due to their optimal numerical properties for transport time scale studies. ATEP-3D allows the analysis of the particle and power balance with the source and sink during the transport process to evaluate the EP confinement properties.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"42 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141782319","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-25DOI: 10.1088/1741-4326/ad60dc
T. Asai, T. Takahashi, D. Kobayashi, T. Seki, Y. Takeuchi, O. Mitarai, J. Morelli, N. Mizuguchi, S. Dettrick, H. Gota, T. Roche, T. Matsumoto, M. Binderbauer, T. Tajima, M. Inomoto and T. Takahashi
This study successfully developed a refueling technique for a field-reversed configuration (FRC) via axial plasmoid injection and demonstrated it on the FAT-CM device at Nihon University. The target FRC is generated using the collisional-merging formation technique combined with conical theta-pinch formation. Plasmoids with an FRC-like configuration are coaxially injected from both ends of the FAT-CM device toward the preexisting target FRC. Postinjection, the system achieves equilibrium, resulting in increases by factors of 1.8 and 2.4 in the total inventory and plasma energy, respectively, compared to cases without injection. This method effectively accomplishes FRC refueling while preserving the intrinsic characteristics of a simply connected, axisymmetric configuration and a high beta value approaching unity. Therefore, this approach offers potential for repetitive refueling in the reactor stage having a FRC plasma core. Experimental outcomes are compared with magnetohydrodynamic simulation results. In the collisional merging process, the characteristics of the pre-collision plasmoids, such as the strong toroidal rotation and coherent FRC-like magnetic field structures of the FRC, are not preserved. Experimental environments have been constructed to investigate such unique properties of the resulting FRCs.
{"title":"Refueling of field-reversed configuration core via axial plasmoids injection","authors":"T. Asai, T. Takahashi, D. Kobayashi, T. Seki, Y. Takeuchi, O. Mitarai, J. Morelli, N. Mizuguchi, S. Dettrick, H. Gota, T. Roche, T. Matsumoto, M. Binderbauer, T. Tajima, M. Inomoto and T. Takahashi","doi":"10.1088/1741-4326/ad60dc","DOIUrl":"https://doi.org/10.1088/1741-4326/ad60dc","url":null,"abstract":"This study successfully developed a refueling technique for a field-reversed configuration (FRC) via axial plasmoid injection and demonstrated it on the FAT-CM device at Nihon University. The target FRC is generated using the collisional-merging formation technique combined with conical theta-pinch formation. Plasmoids with an FRC-like configuration are coaxially injected from both ends of the FAT-CM device toward the preexisting target FRC. Postinjection, the system achieves equilibrium, resulting in increases by factors of 1.8 and 2.4 in the total inventory and plasma energy, respectively, compared to cases without injection. This method effectively accomplishes FRC refueling while preserving the intrinsic characteristics of a simply connected, axisymmetric configuration and a high beta value approaching unity. Therefore, this approach offers potential for repetitive refueling in the reactor stage having a FRC plasma core. Experimental outcomes are compared with magnetohydrodynamic simulation results. In the collisional merging process, the characteristics of the pre-collision plasmoids, such as the strong toroidal rotation and coherent FRC-like magnetic field structures of the FRC, are not preserved. Experimental environments have been constructed to investigate such unique properties of the resulting FRCs.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"71 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141782322","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-22DOI: 10.1088/1741-4326/ad60dd
P.O. Adebayo-Ige, K.F. Gan, C.J. Lasnier, R. Maingi and B.D. Wirth
A thermography inversion algorithm has been developed in the open-source Python-based computer code, HYPERION, to calculate the heat flux incident on plasma-facing components (PFCs) in axisymmetric tokamaks. The chosen mesh size at the surface significantly affects the calculated transient heat flux results. The calculated transient heat flux will exceed the real value when the mesh size tends to zero but will underestimate the real value when the mesh size is large. A criterion for determining the appropriate mesh size for the transient heat flux calculation will be discussed. The numerical scheme for HYPERION uses a 2D fully implicit finite-difference approach, allowing temperature-dependent thermal properties of PFC materials. The inversion algorithm is benchmarked against established heat flux calculation codes, TACO and THEODOR, based on thermography data from NSTX and DIII-D respectively. The primary benefits of HYPERION compared to TACO and THEODOR are that it is open-source and it allows for the optimization of mesh thickness along the substrate. The algorithm also accounts for the thermal properties of thin surface layers that characteristically form on PFCs due to plasma-material interactions. The agreement between HYPERION and THEODOR is excellent, as the percent difference between the codes is ∼5% on average in the case of the DIII-D data for moderate to high heat flux. Verification tests with TACO show slightly higher average percent differences of 8% and 12%. In using HYPERION to study filaments in heat flux, the initial results indicate that small ELMs filaments significantly broaden the divertor heat flux, and decrease divertor peak flux. Compared to the inter-ELM, the small ELM filaments decrease the divertor peak surface temperature. With intermittent divertor filaments, the divertor heat flux width is comparable with that found in L-mode.
在基于 Python 的开源计算机代码 HYPERION 中开发了一种热成像反演算法，用于计算轴对称托卡马克中面向等离子体部件（PFC）的热通量。所选的表面网格尺寸会对瞬态热通量的计算结果产生重大影响。当网格尺寸趋近于零时，计算出的瞬态热通量将超过实际值，但当网格尺寸较大时，计算出的瞬态热通量将低于实际值。本文将讨论如何为瞬态热通量计算确定合适的网格尺寸。HYPERION 的数值方案采用二维全隐式有限差分方法，允许 PFC 材料的热特性随温度变化。反演算法分别以 NSTX 和 DIII-D 的热成像数据为基础，与 TACO 和 THEODOR 等成熟的热通量计算代码进行了比较。与 TACO 和 THEODOR 相比，HYPERION 的主要优势在于它是开源的，而且可以优化沿基体的网格厚度。该算法还考虑到了 PFC 因等离子体与材料相互作用而形成的薄表面层的热特性。HYPERION 和 THEODOR 之间的一致性非常好，在中高热通量的 DIII-D 数据情况下，代码之间的百分比差异平均为 ∼5%。使用 TACO 进行的验证测试表明，平均百分比差异略高，分别为 8%和 12%。在使用HYPERION研究热通量中的细丝时，初步结果表明，小的ELM细丝明显扩大了分流器的热通量，并降低了分流器的峰值通量。与间歇式 ELM 相比，小 ELM 丝降低了岔流器的峰值表面温度。在间歇式分流器灯丝的作用下，分流器热通量宽度与 L 模式下的相当。
{"title":"Divertor heat load estimates on NSTX and DIII-D using new and open-source 2D inversion analysis code","authors":"P.O. Adebayo-Ige, K.F. Gan, C.J. Lasnier, R. Maingi and B.D. Wirth","doi":"10.1088/1741-4326/ad60dd","DOIUrl":"https://doi.org/10.1088/1741-4326/ad60dd","url":null,"abstract":"A thermography inversion algorithm has been developed in the open-source Python-based computer code, HYPERION, to calculate the heat flux incident on plasma-facing components (PFCs) in axisymmetric tokamaks. The chosen mesh size at the surface significantly affects the calculated transient heat flux results. The calculated transient heat flux will exceed the real value when the mesh size tends to zero but will underestimate the real value when the mesh size is large. A criterion for determining the appropriate mesh size for the transient heat flux calculation will be discussed. The numerical scheme for HYPERION uses a 2D fully implicit finite-difference approach, allowing temperature-dependent thermal properties of PFC materials. The inversion algorithm is benchmarked against established heat flux calculation codes, TACO and THEODOR, based on thermography data from NSTX and DIII-D respectively. The primary benefits of HYPERION compared to TACO and THEODOR are that it is open-source and it allows for the optimization of mesh thickness along the substrate. The algorithm also accounts for the thermal properties of thin surface layers that characteristically form on PFCs due to plasma-material interactions. The agreement between HYPERION and THEODOR is excellent, as the percent difference between the codes is ∼5% on average in the case of the DIII-D data for moderate to high heat flux. Verification tests with TACO show slightly higher average percent differences of 8% and 12%. In using HYPERION to study filaments in heat flux, the initial results indicate that small ELMs filaments significantly broaden the divertor heat flux, and decrease divertor peak flux. Compared to the inter-ELM, the small ELM filaments decrease the divertor peak surface temperature. With intermittent divertor filaments, the divertor heat flux width is comparable with that found in L-mode.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"20 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141753916","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-07-22DOI: 10.1088/1741-4326/ad6014
C. Paz-Soldan, S. Gu, N. Leuthold, P. Lunia, P. Xie, M.W. Kim, S.K. Kim, N.C. Logan, J.-K. Park, W. Suttrop, Y. Sun, D.B. Weisberg, M. Willensdorfer, the ASDEX Upgrade Team, the DIII-D Team, the EAST Team and the KSTAR Team
The operational space and global performance of plasmas with edge-localized modes (ELMs) suppressed by resonant magnetic perturbations (RMPs) are surveyed by comparing AUG, DIII-D, EAST, and KSTAR stationary operating points. RMP-ELM suppression is achieved over a range of plasma currents, toroidal fields, and RMP toroidal mode numbers. Consistent operational windows in edge safety factor are found across devices, while windows in plasma shaping parameters are distinct. Accessed pedestal parameters reveal a quantitatively similar pedestal-top density limit for RMP-ELM suppression in all devices of just over m−3. This is surprising given the wide variance of many engineering parameters and edge collisionalities, and poses a challenge to extrapolation of the regime. Wide ranges in input power, confinement time, and stored energy are observed, with the achieved triple product found to scale like the product of current, field, and radius. Observed energy confinement scaling with engineering parameters for RMP-ELM suppressed plasmas are presented and compared with expectations from established H and L-mode scalings, including treatment of uncertainty analysis. Different scaling exponents for individual engineering parameters are found as compared to the established scalings. However, extrapolation to next-step tokamaks ITER and SPARC find overall consistency within uncertainties with the established scalings, finding no obvious performance penalty when extrapolating from the assembled multi-device RMP-ELM suppressed database. Overall this work identifies common physics for RMP-ELM suppression and highlights the need to pursue this no-ELM regime at higher magnetic field and different plasma physical size.
通过比较 AUG、DIII-D、EAST 和 KSTAR 静止运行点，研究了边缘定位模式（ELM）被共振磁扰动（RMP）抑制的等离子体的运行空间和全局性能。在一定范围内的等离子体电流、环形磁场和共振磁扰动环形模式数都能实现共振磁扰动-ELM抑制。在各装置中发现了一致的边缘安全系数运行窗口，而等离子体整形参数窗口则各不相同。访问基座参数后发现，所有设备的 RMP-ELM 抑制基座顶密度极限都非常相似，都略高于 m-3。考虑到许多工程参数和边缘碰撞性的巨大差异，这一结果令人惊讶，并对该机制的推断提出了挑战。我们观察到输入功率、束缚时间和存储能量的范围很广，所实现的三乘积与电流、磁场和半径的乘积相似。本文介绍了 RMP-ELM 抑制等离子体工程参数的能量禁锢缩放观察结果，并与已建立的 H 模式和 L 模式缩放预期结果进行了比较，包括对不确定性分析的处理。与既定标度相比，各工程参数的标度指数有所不同。然而，外推到下一步的热核聚变实验堆 ITER 和 SPARC 发现，在不确定性范围内，总体上与既定标度一致，从组装的多设备 RMP-ELM 抑制数据库外推时，没有发现明显的性能损失。总之，这项工作确定了 RMP-ELM 抑制的共同物理原理，并强调了在更高磁场和不同等离子体物理尺寸下追求这种无ELM 机制的必要性。
{"title":"Plasma performance and operational space with an RMP-ELM suppressed edge","authors":"C. Paz-Soldan, S. Gu, N. Leuthold, P. Lunia, P. Xie, M.W. Kim, S.K. Kim, N.C. Logan, J.-K. Park, W. Suttrop, Y. Sun, D.B. Weisberg, M. Willensdorfer, the ASDEX Upgrade Team, the DIII-D Team, the EAST Team and the KSTAR Team","doi":"10.1088/1741-4326/ad6014","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6014","url":null,"abstract":"The operational space and global performance of plasmas with edge-localized modes (ELMs) suppressed by resonant magnetic perturbations (RMPs) are surveyed by comparing AUG, DIII-D, EAST, and KSTAR stationary operating points. RMP-ELM suppression is achieved over a range of plasma currents, toroidal fields, and RMP toroidal mode numbers. Consistent operational windows in edge safety factor are found across devices, while windows in plasma shaping parameters are distinct. Accessed pedestal parameters reveal a quantitatively similar pedestal-top density limit for RMP-ELM suppression in all devices of just over m−3. This is surprising given the wide variance of many engineering parameters and edge collisionalities, and poses a challenge to extrapolation of the regime. Wide ranges in input power, confinement time, and stored energy are observed, with the achieved triple product found to scale like the product of current, field, and radius. Observed energy confinement scaling with engineering parameters for RMP-ELM suppressed plasmas are presented and compared with expectations from established H and L-mode scalings, including treatment of uncertainty analysis. Different scaling exponents for individual engineering parameters are found as compared to the established scalings. However, extrapolation to next-step tokamaks ITER and SPARC find overall consistency within uncertainties with the established scalings, finding no obvious performance penalty when extrapolating from the assembled multi-device RMP-ELM suppressed database. Overall this work identifies common physics for RMP-ELM suppression and highlights the need to pursue this no-ELM regime at higher magnetic field and different plasma physical size.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"98 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141753993","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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