Pub Date : 2024-07-01DOI: 10.1088/1741-4326/ad5d7c
Tatsuhiro Nasu, T. Tokuzawa, Motoki Nakata, K. Ida, S. Inagaki, Masaki Nishiura, Yasuo Yoshimura, Ryoma Yanai, Kenji Tanaka, M. Yoshinuma, Tatsuya Kobayashi, Akira Ejiri, K. Y. Watanabe, Ichihiro Yamada
The electron-scale turbulence, whose wavelength is about the electron Larmor radius, is thought to have the potential to cause the stiffness of the electron temperature gradient and degrade the confinement of future burning plasma in which the electron heating by alpha particles is dominant. The dependence of electron-scale turbulence and electron heat flux on the electron temperature inverse gradient length Rax/LTe were investigated. The electron temperature gradient is successfully varied in the range of −3 < Rax/LTe < 12 by controlling the injection power of the on/off-axis electron cyclotron heating. The results show a significant increase in electron-scale turbulence with increasing Rax/LTe, especially in conditions where the Electron Temperature Gradient (ETG) instability is linearly unstable, suggesting the presence of ETG turbulence at high Rax/LTe. The electron heat flux also increases steeply with increasing Rax/LTe. In addition, electron-scale turbulence is observed even at Rax/LTe ∼ 0, which is stable in linear GKV calculations. Finding the cause of this phenomenon is an interesting task for the future.
{"title":"Electron-scale turbulence characteristics with varying electron temperature gradient in LHD","authors":"Tatsuhiro Nasu, T. Tokuzawa, Motoki Nakata, K. Ida, S. Inagaki, Masaki Nishiura, Yasuo Yoshimura, Ryoma Yanai, Kenji Tanaka, M. Yoshinuma, Tatsuya Kobayashi, Akira Ejiri, K. Y. Watanabe, Ichihiro Yamada","doi":"10.1088/1741-4326/ad5d7c","DOIUrl":"https://doi.org/10.1088/1741-4326/ad5d7c","url":null,"abstract":"\u0000 The electron-scale turbulence, whose wavelength is about the electron Larmor radius, is thought to have the potential to cause the stiffness of the electron temperature gradient and degrade the confinement of future burning plasma in which the electron heating by alpha particles is dominant. The dependence of electron-scale turbulence and electron heat flux on the electron temperature inverse gradient length Rax/LTe were investigated. The electron temperature gradient is successfully varied in the range of −3 < Rax/LTe < 12 by controlling the injection power of the on/off-axis electron cyclotron heating. The results show a significant increase in electron-scale turbulence with increasing Rax/LTe, especially in conditions where the Electron Temperature Gradient (ETG) instability is linearly unstable, suggesting the presence of ETG turbulence at high Rax/LTe. The electron heat flux also increases steeply with increasing Rax/LTe. In addition, electron-scale turbulence is observed even at Rax/LTe ∼ 0, which is stable in linear GKV calculations. Finding the cause of this phenomenon is an interesting task for the future.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"43 5","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141713091","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 : 2024-05-23DOI: 10.1088/1741-4326/ad4f9f
Tommi Tapani Lyytinen, Antti Snicker, Joonas Virtanen, Iole Palermo, Javier Alguacil, Timo Jos Bogaarts, Felix Warmer
This contribution presents neutron transport studies for the 5-period HELIAS stellarator using the Serpent2 code. These studies utilize a parametric geometry model, enabling scans in neutronics modeling by varying the thickness of the reactor layers. For example, the tritium breeding ratio (TBR) can be determined by exploring various blanket material options and thicknesses to identify the threshold configuration that meets the TBR design criterion of 1.15. We found out that with the helium-cooled pebble ped (HCPB) candidate option, the TBR criterion is met with a breeding zone thickness of 26 cm, while with the dual-coolant lithium lead (DCLL) the threshold is exceeded at a thickness of 46 cm. Furthermore, the geometry includes non-planar field coils, allowing to study the fast neutron flux in these superconducting coils with a technological limit of 1e9 1/cm2s. It is shown that the neutron fast flux is not constant at the coils, necessitating a neutron transport simulation to determine the distribution of the fast-flux at the coils. We show that the peak flux can be more than a factor of 2 higher than the average flux, and that the peak flux location rotates helically.
{"title":"Proof-of-principle of parametric stellarator neutronics modeling using Serpent2","authors":"Tommi Tapani Lyytinen, Antti Snicker, Joonas Virtanen, Iole Palermo, Javier Alguacil, Timo Jos Bogaarts, Felix Warmer","doi":"10.1088/1741-4326/ad4f9f","DOIUrl":"https://doi.org/10.1088/1741-4326/ad4f9f","url":null,"abstract":"\u0000 This contribution presents neutron transport studies for the 5-period HELIAS stellarator using the Serpent2 code. These studies utilize a parametric geometry model, enabling scans in neutronics modeling by varying the thickness of the reactor layers. For example, the tritium breeding ratio (TBR) can be determined by exploring various blanket material options and thicknesses to identify the threshold configuration that meets the TBR design criterion of 1.15. We found out that with the helium-cooled pebble ped (HCPB) candidate option, the TBR criterion is met with a breeding zone thickness of 26 cm, while with the dual-coolant lithium lead (DCLL) the threshold is exceeded at a thickness of 46 cm. Furthermore, the geometry includes non-planar field coils, allowing to study the fast neutron flux in these superconducting coils with a technological limit of 1e9 1/cm2s. It is shown that the neutron fast flux is not constant at the coils, necessitating a neutron transport simulation to determine the distribution of the fast-flux at the coils. We show that the peak flux can be more than a factor of 2 higher than the average flux, and that the peak flux location rotates helically.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"21 8","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141105507","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 : 2024-05-23DOI: 10.1088/1741-4326/ad4f9d
Jae-Sun Park, Jeremy D Lore, Matthew L Reinke, Adam Q. Kuang, Sebastian De Pascuale, Alexander J Creely
This paper presents the application of full time-dependent SOLPS-ITER simulations for actuator design in the SPARC tokamak. This study employs both the EIRENE module, a neutral solver, and the B2.5 plasma module in a time-dependent mode. This is in contrast to most SOLPS simulations, which focus on steady-state solutions, where the neutral distribution is evolved without any time limit or for a time step of 1·10-3 second, which is several orders of magnitude larger than the fluid plasma time step. The time-dependent EIRENE was tested with a fixed B2.5 background and compared with a simple conductance based model in a simplified pump chamber geometry. This comparison aimed to verify the reliability of the neutral relaxation timescale derived from the time-dependent EIRENE. Subsequently, a full time-dependent simulation was performed in a realistic geometry, with the Monte-Carlo neutral time step synchronized with the plasma fluid time step. The numerical setup of the code, including relative time steps and the size of the census data used to store Monte-Carlo particle information is considered. The full-time dependent simulations are then applied to inform the design of the SPARC louver structure, which affects divertor plasma parameters by regulating the neutral conductance from the divertor to the pump. The response of the plasma and neutral parameters was captured on a timescale that enables the design of the actuator to consider time-dependent control capability. It was found that changing the louver opacity has an equivalent effect as varying the gas throughput via puff actuation. Therefore, equivalent divertor plasma conditions can be obtained from both actuators, while the neutral pressure distribution in the pump and divertor differs for each actuator.
{"title":"Full time-dependent SOLPS-ITER simulation of the SPARC tokamak: Actuator design for particle and divertor condition control","authors":"Jae-Sun Park, Jeremy D Lore, Matthew L Reinke, Adam Q. Kuang, Sebastian De Pascuale, Alexander J Creely","doi":"10.1088/1741-4326/ad4f9d","DOIUrl":"https://doi.org/10.1088/1741-4326/ad4f9d","url":null,"abstract":"\u0000 This paper presents the application of full time-dependent SOLPS-ITER simulations for actuator design in the SPARC tokamak. This study employs both the EIRENE module, a neutral solver, and the B2.5 plasma module in a time-dependent mode. This is in contrast to most SOLPS simulations, which focus on steady-state solutions, where the neutral distribution is evolved without any time limit or for a time step of 1·10-3 second, which is several orders of magnitude larger than the fluid plasma time step. The time-dependent EIRENE was tested with a fixed B2.5 background and compared with a simple conductance based model in a simplified pump chamber geometry. This comparison aimed to verify the reliability of the neutral relaxation timescale derived from the time-dependent EIRENE. Subsequently, a full time-dependent simulation was performed in a realistic geometry, with the Monte-Carlo neutral time step synchronized with the plasma fluid time step. The numerical setup of the code, including relative time steps and the size of the census data used to store Monte-Carlo particle information is considered. The full-time dependent simulations are then applied to inform the design of the SPARC louver structure, which affects divertor plasma parameters by regulating the neutral conductance from the divertor to the pump. The response of the plasma and neutral parameters was captured on a timescale that enables the design of the actuator to consider time-dependent control capability. It was found that changing the louver opacity has an equivalent effect as varying the gas throughput via puff actuation. Therefore, equivalent divertor plasma conditions can be obtained from both actuators, while the neutral pressure distribution in the pump and divertor differs for each actuator.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"17 6","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141107139","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 : 2024-05-23DOI: 10.1088/1741-4326/ad4f9e
G. Holt, A. Keats, Stanislas Pamela, Mike Kryjak, A. Agnello, N. Amorisco, Benjamin Dudson, M. Smyrnakis
Future tokamak devices that aim to create conditions relevant to power plant operations must consider strategies for mitigating damage to plasma facing components in the divertor. One of the goals of MAST-U tokamak operations is to inform these considerations by researching advanced divertor configurations that aid stable plasma detachment. Machine design, scenario planning and detachment control would all greatly benefit from tools that enable rapid calculation of scenario-relevant quantities given some input parameters. This paper presents a method for generating large, simulated scrape-off layer data sets, which was applied to generate a data set of steady-state Hermes-3 simulations of the MAST-U tokamak. A machine learning model was constructed using a Bayesian approach to hyperparameter optimisation to predict diagnosable output quantities given control-relevant input features. The resulting best-performing model, which is based on a feedfoward neural network, achieves high accuracy when predicting electron temperature at the divertor target and carbon impurity radiation front position and runs in around 1 ms in inference mode. Techniques for interpreting the predictions made by the model were applied, and a high-resolution parameter scan of upstream conditions was performed to demonstrate the utility of rapidly generating accurate predictions using the emulator. This work represents a step forward in the design of machine learning-driven emulators of tokamak exhaust simulation codes in operational modes relevant to divertor detachment control and plasma scenario design.
{"title":"Tokamak divertor plasma emulation with machine learning","authors":"G. Holt, A. Keats, Stanislas Pamela, Mike Kryjak, A. Agnello, N. Amorisco, Benjamin Dudson, M. Smyrnakis","doi":"10.1088/1741-4326/ad4f9e","DOIUrl":"https://doi.org/10.1088/1741-4326/ad4f9e","url":null,"abstract":"\u0000 Future tokamak devices that aim to create conditions relevant to power plant operations must consider strategies for mitigating damage to plasma facing components in the divertor. One of the goals of MAST-U tokamak operations is to inform these considerations by researching advanced divertor configurations that aid stable plasma detachment. Machine design, scenario planning and detachment control would all greatly benefit from tools that enable rapid calculation of scenario-relevant quantities given some input parameters. This paper presents a method for generating large, simulated scrape-off layer data sets, which was applied to generate a data set of steady-state Hermes-3 simulations of the MAST-U tokamak. A machine learning model was constructed using a Bayesian approach to hyperparameter optimisation to predict diagnosable output quantities given control-relevant input features. The resulting best-performing model, which is based on a feedfoward neural network, achieves high accuracy when predicting electron temperature at the divertor target and carbon impurity radiation front position and runs in around 1 ms in inference mode. Techniques for interpreting the predictions made by the model were applied, and a high-resolution parameter scan of upstream conditions was performed to demonstrate the utility of rapidly generating accurate predictions using the emulator. This work represents a step forward in the design of machine learning-driven emulators of tokamak exhaust simulation codes in operational modes relevant to divertor detachment control and plasma scenario design.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"36 46","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141104070","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 : 2024-05-23DOI: 10.1088/1741-4326/ad4f9c
David Moulton, J. R. Harrison, L. Xiang, Peter Ryan, Andrew Kirk, Kevin Verhaegh, T. Wijkamp, F. Federici, James G Clark, B. Lipschultz
Measurements are presented, alongside corresponding interpretative SOLPS-ITER simulations, of the first MAST-U experiments comparing ohmically heated L-mode fuelling scans in Conventional divertor (CD) and Super-X divertor (SXD) configurations. In experiment, at comparable outer mid-plane separatrix electron density, $n_{e,rm{sep,OMP}}$, the maximum lower outer target heat load was found to be a factor 16$pm7$ lower in SXD compared to CD. In simulation, a factor 26.8 reduction was found (slightly higher than the experimental range), suggesting an additional reduction in SXD compared to the factor 9.3 expected from geometric considerations alone. According to the simulations, this additional reduction in the SXD is due to a net radial transport of the energy remaining downstream of the $T_e=5$ eV location. This energy is carried out of the critical (highest heat load) flux tube by deuterium atoms, demonstrating the importance of a longer legged divertor which provides space for this to occur. Importantly, in both simulation and experiment, the SXD has minimal impact on the upstream $n_e$ and $T_e$ profiles. Spectral inferences of detachment front movement in SXD compare well between simulation and experiment. In regions of high magnetic field gradient, the parallel movement of the front towards the X-point becomes less sensitive to increasing $n_{e,rm{sep,OMP}}$, in qualitative agreement with simplified models and previous predictive simulations. Additional aspects, regarding the target ion flux rollover, upstream separatrix temperature and drift effects, are also presented and discussed.
本文介绍了首次 MAST-U 实验的测量结果以及相应的 SOLPS-ITER 仿真解释，这些实验比较了在常规分流器（CD）和超 X 分流器（SXD）配置下的欧姆加热 L 模式燃料扫描。在实验中，在外部中平面分离矩阵电子密度$n_{e,rm{sep,OMP}}$相当的情况下，发现SXD与CD相比，最大下部外靶热负荷低了16$pm7$。在模拟中，发现降低了 26.8 倍（略高于实验范围），这表明 SXD 比仅从几何角度考虑预计的 9.3 倍还要低。根据模拟结果，SXD 的额外降低是由于 $T_e=5 eV 位置下游剩余能量的净径向传输。这些能量被氘原子带出临界（热负荷最高）通量管，这表明了为这种情况提供空间的长脚分流器的重要性。重要的是，在模拟和实验中，SXD 对上游 $n_e$ 和 $T_e$ 曲线的影响都很小。在 SXD 中，分离前沿移动的光谱推断在模拟和实验中都有很好的对比。在高磁场梯度区域，前沿向X点的平行运动对增加$n_{e,rm{sep,OMP}}$不那么敏感，这与简化模型和以前的预测模拟在质量上是一致的。此外，还介绍并讨论了有关目标离子通量翻转、上游分离矩阵温度和漂移效应的其他方面。
{"title":"Super-X and Conventional divertor configurations in MAST-U ohmic L-mode; a comparison facilitated by interpretative modelling","authors":"David Moulton, J. R. Harrison, L. Xiang, Peter Ryan, Andrew Kirk, Kevin Verhaegh, T. Wijkamp, F. Federici, James G Clark, B. Lipschultz","doi":"10.1088/1741-4326/ad4f9c","DOIUrl":"https://doi.org/10.1088/1741-4326/ad4f9c","url":null,"abstract":"\u0000 Measurements are presented, alongside corresponding interpretative SOLPS-ITER simulations, of the first MAST-U experiments comparing ohmically heated L-mode fuelling scans in Conventional divertor (CD) and Super-X divertor (SXD) configurations. In experiment, at comparable outer mid-plane separatrix electron density, $n_{e,rm{sep,OMP}}$, the maximum lower outer target heat load was found to be a factor 16$pm7$ lower in SXD compared to CD. In simulation, a factor 26.8 reduction was found (slightly higher than the experimental range), suggesting an additional reduction in SXD compared to the factor 9.3 expected from geometric considerations alone. According to the simulations, this additional reduction in the SXD is due to a net radial transport of the energy remaining downstream of the $T_e=5$ eV location. This energy is carried out of the critical (highest heat load) flux tube by deuterium atoms, demonstrating the importance of a longer legged divertor which provides space for this to occur. Importantly, in both simulation and experiment, the SXD has minimal impact on the upstream $n_e$ and $T_e$ profiles. Spectral inferences of detachment front movement in SXD compare well between simulation and experiment. In regions of high magnetic field gradient, the parallel movement of the front towards the X-point becomes less sensitive to increasing $n_{e,rm{sep,OMP}}$, in qualitative agreement with simplified models and previous predictive simulations. Additional aspects, regarding the target ion flux rollover, upstream separatrix temperature and drift effects, are also presented and discussed.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"47 3","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141103466","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 : 2024-05-23DOI: 10.1088/1741-4326/ad4fa0
Guanghou Yan, Bojiang Ding, Chenbin Wu, Miaohui Li, Seung Gyou Baek, Francesco Napoli, Taotao Zhou, K. Geng, N. Yan, Mao Wang, Xinjun Zhang, Handong Xu, J. H. Yang, W. Ma
Effect of parametric decay instability (PDI) on current drive efficiency of 4.6 GHz lower hybrid wave in EAST is investigated experimentally, showing the PDI channel bifurcation of lower hybrid (LH) wave for the first time in EAST. Firstly, experiments with 3 platforms of LH power were performed, achieving the LH power required for the PDI occurrence. Secondly, PDI bifurcation experiments were further carried out by ramping up plasma electron density. Loop voltage increases with density increase, implying the decrease of current driven by LH wave. PDI bifurcation during electron density ramp-up was studied by analyzing the parallel refractive index (n_∥) and the frequency spectrum broadening, which is measured with a radio frequency (RF) magnetic probe array recently installed close to the LH antenna. It is observed for the first time that both of them firstly increase with density, then there is not much variation and a clear sideband in the frequency spectrum is also observed when the density is up to 4×10^19 m^(-3), suggesting the change of the PDI channel. Calculation of the mode growth rate driven by PDI shows that when the edge electron density is up to 1.9×10^18 m^(-3), the growth rate of ion cyclotron quasi-mode (ICQM) will exceed that of ion sound quasi-mode (ISQM), quantitively explaining that with a density increase, PDI channel partly transits from ISQM to ICQM channel. Studies provide a possible way to reduce the power deposition in the edge region and improve LHCD capability by means of mitigating PDI behavior.
{"title":"Experimental investigation on PDI bifurcation of lower hybrid wave during electron density ramp-up in EAST","authors":"Guanghou Yan, Bojiang Ding, Chenbin Wu, Miaohui Li, Seung Gyou Baek, Francesco Napoli, Taotao Zhou, K. Geng, N. Yan, Mao Wang, Xinjun Zhang, Handong Xu, J. H. Yang, W. Ma","doi":"10.1088/1741-4326/ad4fa0","DOIUrl":"https://doi.org/10.1088/1741-4326/ad4fa0","url":null,"abstract":"\u0000 Effect of parametric decay instability (PDI) on current drive efficiency of 4.6 GHz lower hybrid wave in EAST is investigated experimentally, showing the PDI channel bifurcation of lower hybrid (LH) wave for the first time in EAST. Firstly, experiments with 3 platforms of LH power were performed, achieving the LH power required for the PDI occurrence. Secondly, PDI bifurcation experiments were further carried out by ramping up plasma electron density. Loop voltage increases with density increase, implying the decrease of current driven by LH wave. PDI bifurcation during electron density ramp-up was studied by analyzing the parallel refractive index (n_∥) and the frequency spectrum broadening, which is measured with a radio frequency (RF) magnetic probe array recently installed close to the LH antenna. It is observed for the first time that both of them firstly increase with density, then there is not much variation and a clear sideband in the frequency spectrum is also observed when the density is up to 4×10^19 m^(-3), suggesting the change of the PDI channel. Calculation of the mode growth rate driven by PDI shows that when the edge electron density is up to 1.9×10^18 m^(-3), the growth rate of ion cyclotron quasi-mode (ICQM) will exceed that of ion sound quasi-mode (ISQM), quantitively explaining that with a density increase, PDI channel partly transits from ISQM to ICQM channel. Studies provide a possible way to reduce the power deposition in the edge region and improve LHCD capability by means of mitigating PDI behavior.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"39 13","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141103594","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 : 2024-05-22DOI: 10.1088/1741-4326/ad4ef2
Qingzhou Yu, Hao Xu, Zhaoxi Chen, Qingxi Yang
An efficient optimization design for large and complex components of fusion reactor is crucial to address the engineering design requirements and further promote technical standardization. Based on research status, current engineering designs for fusion reactors have some deficiencies such as time and energy wastes, inefficiency and the difficulties in covering the typical “multi-variable, multi-objective” design requirements. They are pressing and common problems that urgently need to be overcome. To deal with the aforementioned technical challenges, it is vitally important to design an efficient, precise, and normalized approach tailored for the development of future fusion reactors. Therefore, this paper proposes a process-oriented optimization design method, which involves Coupled external parameterized modeling, Experimental points design, Response surface optimization and Structural integrity validation (CERS), to improve currently inefficient design methods. And the vacuum cryostat, the largest and complex component of tokamak, is taken as an example to present the basic procedures of CERS. Firstly, the functions, basic structures, load types, analysis methods and verification criteria of the cryostat are presented in detail. And then real-time data interaction between external global parametric variables and ANSYS via coupling is established by CERS, which achieves parametric modeling of cryostat and the efficient experiment point design and optimization analysis with multi-variables and multi-objectives in an automatic way. Subsequently, this study demonstrates the significance and sensitivity of various structural parameters of cryostat from such objectives as maximum deformation, maximum equivalent stress, and total mass. And the optimal set of its structural parameters is obtained by establishing mathematical optimization model. Finally, the structural integrity is verified. The result indicates that the optimized cryostat maintains a minimum safety margin of 23% and will not suffer fatigue damage under various load events during its service. Besides, the nonlinear buckling load multiplier ∅ is 5.4, obtained by analyzing the load-displacement curve of cryostat according to the zero-curvature criterion. This shows that the designed cryostat is stable enough. The proposed method is simple, efficient and reliable which can be applied to both cryostat and other complex components of fusion reactors in the engineering design fields. It has great value of practically technical reference and can further promote the standardization of engineering design technology for future fusion reactor.
{"title":"Research on the efficient process-oriented structural optimization method of the large-scale vacuum cryostat for fusion reactor","authors":"Qingzhou Yu, Hao Xu, Zhaoxi Chen, Qingxi Yang","doi":"10.1088/1741-4326/ad4ef2","DOIUrl":"https://doi.org/10.1088/1741-4326/ad4ef2","url":null,"abstract":"\u0000 An efficient optimization design for large and complex components of fusion reactor is crucial to address the engineering design requirements and further promote technical standardization. Based on research status, current engineering designs for fusion reactors have some deficiencies such as time and energy wastes, inefficiency and the difficulties in covering the typical “multi-variable, multi-objective” design requirements. They are pressing and common problems that urgently need to be overcome. To deal with the aforementioned technical challenges, it is vitally important to design an efficient, precise, and normalized approach tailored for the development of future fusion reactors. Therefore, this paper proposes a process-oriented optimization design method, which involves Coupled external parameterized modeling, Experimental points design, Response surface optimization and Structural integrity validation (CERS), to improve currently inefficient design methods. And the vacuum cryostat, the largest and complex component of tokamak, is taken as an example to present the basic procedures of CERS. Firstly, the functions, basic structures, load types, analysis methods and verification criteria of the cryostat are presented in detail. And then real-time data interaction between external global parametric variables and ANSYS via coupling is established by CERS, which achieves parametric modeling of cryostat and the efficient experiment point design and optimization analysis with multi-variables and multi-objectives in an automatic way. Subsequently, this study demonstrates the significance and sensitivity of various structural parameters of cryostat from such objectives as maximum deformation, maximum equivalent stress, and total mass. And the optimal set of its structural parameters is obtained by establishing mathematical optimization model. Finally, the structural integrity is verified. The result indicates that the optimized cryostat maintains a minimum safety margin of 23% and will not suffer fatigue damage under various load events during its service. Besides, the nonlinear buckling load multiplier ∅ is 5.4, obtained by analyzing the load-displacement curve of cryostat according to the zero-curvature criterion. This shows that the designed cryostat is stable enough. The proposed method is simple, efficient and reliable which can be applied to both cryostat and other complex components of fusion reactors in the engineering design fields. It has great value of practically technical reference and can further promote the standardization of engineering design technology for future fusion reactor.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"57 8","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141110480","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 : 2024-05-22DOI: 10.1088/1741-4326/ad4ef3
A. C. C. Sips, Francesca Turco, Charles Mark Greenfield, Lothar Schmitz, Timothy C Luce, Tomas Odstrcil, Adam G McLean, Igor Bykov, A W Hyatt, Tom H Osborne
FEC 2023 contribution: Experiments in DIII-D document the ITER Baseline Scenario (IBS) at q95 ~ 3 and PIN/PLH ~ 1-2, in both deuterium and hydrogen utilizing Kr and Xe as Tungsten-equivalent radiators. The power threshold for H-mode operation (PLH) was found to be about a factor of two higher than the scaling law. In recent IBS experiments in deuterium, intrinsic levels of metals such as Tungsten (W) or molybdenum and inconel are present that reduce the pedestal pressure by 20-25%. A complete radiative collapse of deuterium IBS plasmas occurs at W core concentrations CW = 10-5. Simulations show that for core temperatures expected for ITER, the plasmas would not have a radiative collapse at CW = 1x10-5, moreover Q = 8-10 would still be achieved for CW up to 3x10-5. In contrast to deuterium, the IBS in hydrogen is not affected by intrinsic high-Z impurities. Krypton was injected in a matrix scan of input power and impurity flow in IBS hydrogen discharges. Krypton impurity density profiles in hydrogen are similar to deuterium plasmas, but at Kr flows that are 2-3 times higher for the same input power. Krypton is transported into the core and affects the whole radius; at the highest injection rates a radiative collapse occurs at core radiation fractions of 0.3-0.35, consistent with the expected maximum W radiation fraction for ITER core plasmas. Comparing the results with previous ITPA database studies of the IBS confirms that at higher radiation fraction due to high-Z impurities, a drop in H98 of >10% is observed. On the other hand, the results using Kr as a W-equivalent radiator indicate that metal (W) devices at lower core temperatures than ITER may provide overly pessimistic performance extrapolations to ITER for deuterium-tritium operation.
FEC 2023 的贡献：DIII-D 中的实验记录了热核实验堆基线方案（IBS）在 q95 ~ 3 和 PIN/PLH ~ 1-2 条件下，利用 Kr 和 Xe 作为钨等效辐射器在氘和氢中进行的实验。研究发现，H 模式工作（PLH）的功率阈值比比例定律高出约 2 倍。在最近的氘核 IBS 实验中，钨（W）或钼和镍镉等金属的内在含量降低了 20-25%的基座压力。在钨核心浓度 CW = 10-5 时，氘 IBS 等离子体会发生完全辐射坍缩。模拟显示，对于热核实验堆的预期堆芯温度，等离子体在 CW = 1x10-5 时不会发生辐射塌缩，而且在 CW 高达 3x10-5 时仍可达到 Q = 8-10。与氘相比，氢中的 IBS 不受内在高 Z 杂质的影响。在对 IBS 氢放电中的输入功率和杂质流进行矩阵扫描时注入了氪。氢中的氪杂质密度曲线与氘等离子体相似，但在相同输入功率下，Kr流量要高出2-3倍。氪被输送到内核并影响整个半径；在最高注入率下，内核辐射分数为 0.3-0.35 时会发生辐射塌缩，这与预期的 ITER 内核等离子体最大 W 辐射分数一致。将这一结果与之前国际热核聚变实验堆数据库的研究结果进行比较后发现，在高 Z 杂质导致的较高辐射分率下，H98 的下降幅度大于 10%。另一方面，使用 Kr 作为等效 W 辐射体的结果表明，在比 ITER 核心温度更低的情况下，金属（W）设备可能会为 ITER 的氘氚运行提供过于悲观的性能推断。
{"title":"Power and isotope effects in the ITER Baseline Scenario with tungsten and tungsten-equivalent radiators in DIII-D","authors":"A. C. C. Sips, Francesca Turco, Charles Mark Greenfield, Lothar Schmitz, Timothy C Luce, Tomas Odstrcil, Adam G McLean, Igor Bykov, A W Hyatt, Tom H Osborne","doi":"10.1088/1741-4326/ad4ef3","DOIUrl":"https://doi.org/10.1088/1741-4326/ad4ef3","url":null,"abstract":"\u0000 FEC 2023 contribution: Experiments in DIII-D document the ITER Baseline Scenario (IBS) at q95 ~ 3 and PIN/PLH ~ 1-2, in both deuterium and hydrogen utilizing Kr and Xe as Tungsten-equivalent radiators. The power threshold for H-mode operation (PLH) was found to be about a factor of two higher than the scaling law. In recent IBS experiments in deuterium, intrinsic levels of metals such as Tungsten (W) or molybdenum and inconel are present that reduce the pedestal pressure by 20-25%. A complete radiative collapse of deuterium IBS plasmas occurs at W core concentrations CW = 10-5. Simulations show that for core temperatures expected for ITER, the plasmas would not have a radiative collapse at CW = 1x10-5, moreover Q = 8-10 would still be achieved for CW up to 3x10-5. In contrast to deuterium, the IBS in hydrogen is not affected by intrinsic high-Z impurities. Krypton was injected in a matrix scan of input power and impurity flow in IBS hydrogen discharges. Krypton impurity density profiles in hydrogen are similar to deuterium plasmas, but at Kr flows that are 2-3 times higher for the same input power. Krypton is transported into the core and affects the whole radius; at the highest injection rates a radiative collapse occurs at core radiation fractions of 0.3-0.35, consistent with the expected maximum W radiation fraction for ITER core plasmas. Comparing the results with previous ITPA database studies of the IBS confirms that at higher radiation fraction due to high-Z impurities, a drop in H98 of >10% is observed. On the other hand, the results using Kr as a W-equivalent radiator indicate that metal (W) devices at lower core temperatures than ITER may provide overly pessimistic performance extrapolations to ITER for deuterium-tritium operation.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"8 6","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141110170","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 : 2024-05-22DOI: 10.1088/1741-4326/ad4ef6
Carmine Castaldo, Lorenzo Della Volpe, Renato Fedele, William Bin, Paolo Buratti, Alessandro Cardinali, Francesco Napoli, Massimo Marinucci, Gerarda Apruzzese, C. Cianfarani, Edmondo Giovannozzi, O. Tudisco
The local stability analysis of plasma waves driven by runaway electrons (RE) has been performed considering hot plasma Maxwellian background, with electron and ion temperatures of the order of 1 keV. It is shown that hot plasma waves, namely electron plasma waves (EPW) and ion Bernstein waves (IBW) can be driven unstable by RE at their coalescence frequency via Cherenkov resonance by RE with energy distribution peaked at about 8 MeV. A skew-normal distribution is used as a model of the RE energy distribution. The EPW and IBW couples of waves occur between any successive ion-cyclotron harmonics frequencies nfci, above the lower hybrid resonance. At their confluence, the perpendicular group velocity vanishes and significant RF emissions are expected. The frequency gap between two successive confluences is ~fci. Groups of RF line emissions, separated by almost constant frequency gap ~fci/2 are detected during various quiescent runaway plasma discharges in the FTU tokamak. The analysis of a specific discharge suggests that the frequencies of the line emissions observed and the frequencies occurring at the EPW-IBW confluences are in reasonable agreement. A possible explanation of the line emissions with ~fci/2 gap in terms of nonlinear mode coupling is proposed.
{"title":"Stability analysis of plasma waves driven by runaway electrons in tokamak hot plasmas","authors":"Carmine Castaldo, Lorenzo Della Volpe, Renato Fedele, William Bin, Paolo Buratti, Alessandro Cardinali, Francesco Napoli, Massimo Marinucci, Gerarda Apruzzese, C. Cianfarani, Edmondo Giovannozzi, O. Tudisco","doi":"10.1088/1741-4326/ad4ef6","DOIUrl":"https://doi.org/10.1088/1741-4326/ad4ef6","url":null,"abstract":"\u0000 The local stability analysis of plasma waves driven by runaway electrons (RE) has been performed considering hot plasma Maxwellian background, with electron and ion temperatures of the order of 1 keV. It is shown that hot plasma waves, namely electron plasma waves (EPW) and ion Bernstein waves (IBW) can be driven unstable by RE at their coalescence frequency via Cherenkov resonance by RE with energy distribution peaked at about 8 MeV. A skew-normal distribution is used as a model of the RE energy distribution. The EPW and IBW couples of waves occur between any successive ion-cyclotron harmonics frequencies nfci, above the lower hybrid resonance. At their confluence, the perpendicular group velocity vanishes and significant RF emissions are expected. The frequency gap between two successive confluences is ~fci. Groups of RF line emissions, separated by almost constant frequency gap ~fci/2 are detected during various quiescent runaway plasma discharges in the FTU tokamak. The analysis of a specific discharge suggests that the frequencies of the line emissions observed and the frequencies occurring at the EPW-IBW confluences are in reasonable agreement. A possible explanation of the line emissions with ~fci/2 gap in terms of nonlinear mode coupling is proposed.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"3 12","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141111232","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 : 2024-05-22DOI: 10.1088/1741-4326/ad4ef7
Xuele Zhao, C. Sang, Yilin Wang, Chen Zhang, Dezhen Wang
The effect of drifts combined with pumping on particle exhaust is assessed using the SOLPS-ITER code package, considering full drifts. Both drifts and pumping speed S can affect particle exhaust. Drifts change the neutral density by influencing plasma flow and the resulting particle recycling. This leads to the accumulation of neutral particles either far away or close to the pump opening location. While the particle exhaust is enhanced as S raises. When the pump opening is positioned at the common flux region (CFR) of the outer divertor (referred to as Pump CFR/OD), particle exhaust is suppressed by drifts in forward Bt, while it is enhanced by drifts in reversed Bt, with fixed S. On the other hand, when the pump is situated in the private flux region (PFR) of the outer divertor (referred to as Pump PFR/OD), particle exhaust is enhanced by drifts in both reversed and forward Bt compared to the case without drifts. Moreover, the effective pumping in reversed Bt is stronger than in forward Bt. In the same Bt direction, Pump PFR/OD has a higher effective pumping than Pump CFR/OD. Increased S results in higher particle exhaust in all Bt direction and pump location cases. The plasma detachment is affected by drifts, S and pump opening location, respectively. For the specified Bt direction and pump opening location case, higher S suppresses plasma detachment. For an identical particle exhaust rate, stronger pumping capacity can promote plasma detachment. Therefore, Pump PFR/OD can more easily achieve outer divertor detachment than Pump CFR/OD in the same Bt direction. Overall, placing the pump at the PFR side of the outer divertor while running in reversed Bt is the best option from the divertor particle exhaust and plasma detachment point of view.
使用 SOLPS-ITER 代码包评估了漂移与泵送相结合对颗粒排气的影响，并考虑了全漂移。漂移和泵送速度 S 都会影响粒子排气。漂移通过影响等离子体流和由此产生的粒子循环来改变中性密度。这导致中性粒子在远离或靠近泵开口位置的地方聚集。而粒子排气会随着 S 的升高而增强。当泵开口位于外分流器的公共通量区（CFR）（称为泵 CFR/OD）时，在 S 固定的情况下，正向 Bt 的漂移会抑制粒子排气，而反向 Bt 的漂移则会增强粒子排气；另一方面，当泵位于外分流器的私人通量区（PFR）（称为泵 PFR/OD）时，与没有漂移的情况相比，反向和正向 Bt 的漂移都会增强粒子排气。此外，反向 Bt 的有效泵送比正向 Bt 更强。在相同的 Bt 方向上，泵 PFR/OD 的有效泵送量高于泵 CFR/OD。在所有 Bt 方向和泵位置情况下，增加 S 会导致更高的粒子排气量。等离子体脱离分别受到漂移、S 和泵开口位置的影响。在指定的 Bt 方向和泵开启位置情况下，较高的 S 会抑制等离子体脱离。在粒子排气速率相同的情况下，更强的泵送能力会促进等离子体脱离。因此，在相同的 Bt 方向上，泵 PFR/OD 比泵 CFR/OD 更容易实现外分流器脱离。总之，从分流器粒子排气和等离子体脱离的角度来看，将泵置于外分流器的 PFR 侧并反向 Bt 运行是最佳选择。
{"title":"The role of divertor pumping combined with full drifts in particle exhaust and divertor plasma","authors":"Xuele Zhao, C. Sang, Yilin Wang, Chen Zhang, Dezhen Wang","doi":"10.1088/1741-4326/ad4ef7","DOIUrl":"https://doi.org/10.1088/1741-4326/ad4ef7","url":null,"abstract":"\u0000 The effect of drifts combined with pumping on particle exhaust is assessed using the SOLPS-ITER code package, considering full drifts. Both drifts and pumping speed S can affect particle exhaust. Drifts change the neutral density by influencing plasma flow and the resulting particle recycling. This leads to the accumulation of neutral particles either far away or close to the pump opening location. While the particle exhaust is enhanced as S raises. When the pump opening is positioned at the common flux region (CFR) of the outer divertor (referred to as Pump CFR/OD), particle exhaust is suppressed by drifts in forward Bt, while it is enhanced by drifts in reversed Bt, with fixed S. On the other hand, when the pump is situated in the private flux region (PFR) of the outer divertor (referred to as Pump PFR/OD), particle exhaust is enhanced by drifts in both reversed and forward Bt compared to the case without drifts. Moreover, the effective pumping in reversed Bt is stronger than in forward Bt. In the same Bt direction, Pump PFR/OD has a higher effective pumping than Pump CFR/OD. Increased S results in higher particle exhaust in all Bt direction and pump location cases. The plasma detachment is affected by drifts, S and pump opening location, respectively. For the specified Bt direction and pump opening location case, higher S suppresses plasma detachment. For an identical particle exhaust rate, stronger pumping capacity can promote plasma detachment. Therefore, Pump PFR/OD can more easily achieve outer divertor detachment than Pump CFR/OD in the same Bt direction. Overall, placing the pump at the PFR side of the outer divertor while running in reversed Bt is the best option from the divertor particle exhaust and plasma detachment point of view.","PeriodicalId":503481,"journal":{"name":"Nuclear Fusion","volume":"62 36","pages":""},"PeriodicalIF":0.0,"publicationDate":"2024-05-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141110740","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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