Pub Date : 2024-09-02DOI: 10.1088/1741-4326/ad6f27
P.T. Lang, G.D. Conway, O.J.W.F. Kardaun, M. Maraschek, B. Pégourié, B. Ploeckl, R. Samulyak, the ASDEX Upgrade Teama
In a future fusion reactor, the main fuelling method will likely rely on the injection of solid hydrogen pellets. Current predictions assume that this goal can be achieved, since being based on a technology which is already largely developed. However, this belief is founded on modelling tools that are usually aligned to the observation made in existing devices and then extrapolated to reactor conditions. This approach needs a sound consideration of its intrinsic restrictions and any observed feature not reproduced by the utilised codes should be applied to check their validation and possibly contribute to their refinement. One specific feature still lacking an explanation of a reasonable and self-consistent mechanism in the current models is the appearance of a phenomenon called striations, which are high frequency variations in the radiation emitted during the pellet ablation process. In order to provide a sound and reliable database for further considerations, a dedicated analysis of this effect has been performed on the mid-size tokamak ASDEX Upgrade. Therefore, such cases have been selected with the relevant signal recorded with sufficient temporal resolution during experiments covering a wide variation of plasma and pellet parameters which are regarded to be potentially influential on the striation pattern. In addition, it was ensured that for any specific case the observed behaviour was reproducible for several individual ablation events under identical conditions. In all cases considered, the observed radiation-intensity variations appear with a typical pattern showing a broad peak of frequencies in the range 50–150 kHz. This characteristic unveils a notable resilience against any parameter variation. This new collection of data can now act as firm basis to corroborate future modelling code-validation efforts. In addition, the analysis method can provide a relatively simple way of reviewing future modelling predictions.
{"title":"Analysis of the periodic variation of pellet ablation radiation intensity in ASDEX Upgrade","authors":"P.T. Lang, G.D. Conway, O.J.W.F. Kardaun, M. Maraschek, B. Pégourié, B. Ploeckl, R. Samulyak, the ASDEX Upgrade Teama","doi":"10.1088/1741-4326/ad6f27","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6f27","url":null,"abstract":"In a future fusion reactor, the main fuelling method will likely rely on the injection of solid hydrogen pellets. Current predictions assume that this goal can be achieved, since being based on a technology which is already largely developed. However, this belief is founded on modelling tools that are usually aligned to the observation made in existing devices and then extrapolated to reactor conditions. This approach needs a sound consideration of its intrinsic restrictions and any observed feature not reproduced by the utilised codes should be applied to check their validation and possibly contribute to their refinement. One specific feature still lacking an explanation of a reasonable and self-consistent mechanism in the current models is the appearance of a phenomenon called striations, which are high frequency variations in the radiation emitted during the pellet ablation process. In order to provide a sound and reliable database for further considerations, a dedicated analysis of this effect has been performed on the mid-size tokamak ASDEX Upgrade. Therefore, such cases have been selected with the relevant signal recorded with sufficient temporal resolution during experiments covering a wide variation of plasma and pellet parameters which are regarded to be potentially influential on the striation pattern. In addition, it was ensured that for any specific case the observed behaviour was reproducible for several individual ablation events under identical conditions. In all cases considered, the observed radiation-intensity variations appear with a typical pattern showing a broad peak of frequencies in the range 50<bold>–</bold>150 kHz. This characteristic unveils a notable resilience against any parameter variation. This new collection of data can now act as firm basis to corroborate future modelling code-validation efforts. In addition, the analysis method can provide a relatively simple way of reviewing future modelling predictions.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"21 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212124","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 : 2024-09-02DOI: 10.1088/1741-4326/ad6e9e
X.R. Duan, M. Xu, W.L. Zhong, X.Q. Ji, W. Chen, Z.B. Shi, X.L. Liu, B. Lu, B. Li, Y.Q. Wang, J.Q. Li, G.Y. Zheng, Yong Liu, Q.W. Yang, L.W. Yan, L.J. Cai, Q. Li, Y. Liu, X.Y. Bai, Z. Cao, X. Chen, H.T. Chen, Y.H. Chen, G.Q. Dong, H.L. Du, D.M. Fan, J.M. Gao, S.F. Geng, G.Z. Hao, H.M. He, M. Huang, M. Jiang, R. Ke, A.S. Liang, J.X. Li, Qing Li, Yongge Li, L.C. Li, H.J. Li, W.B. Li, D.Q. Liu, T. Long, L.F. Lu, L. Nie, P.W. Shi, J.F. Peng, A.P. Sun, T.F. Sun, R.H. Tong, H.L. Wei, S. Wang, G.L. Xiao, X.P. Xiao, L. Xue, H.B. Xu, Z.Y. Yang, D.L. Yu, L.M. Yu, Y.P. Zhang, X. Zheng, L. Zhang, Y. Zhang, F. Zhang, X.L. Zhang, HL-3 Team & Collaborators2345678910111213141516171819
Since the first plasma realized in 2020, a series of key systems on HL-3 (known as HL-2M before) tokamak have been equipped/upgraded, including in-vessel components (the first wall, lower divertor, and toroidal cryogenic/water-cooling/baking/glow discharge systems, etc.), auxiliary heating system of 11 MW, and 28 diagnostic systems (to measure the plasma density, electron temperature, radiation, magnetic field, etc.). Magnet field systems were commissioned firstly for divertor plasma discharges. During the 2nd experimental campaign of HL-3 tokamak, several great progresses have been achieved. Firstly, the successful operation with plasma current larger than 1 MA was achieved under a divertor configuration. Secondly, the advanced divertor concept with two distinct snowflake configurations was realized. It is found that the distribution of ion saturation current and heat flux on bottom plate becomes wide due to magnetic surface expansion, demonstrating the advantage of such configuration in the heat flux mitigation. In addition, using the combination of NBI, ECRH and LHCD, the standard sawtoothing high confinement mode of megampere plasma was firstly accessed on the HL-3. The successful commissioning of HL-3 is beneficial for the initial operation of ITER.
{"title":"Recent advance progress of HL-3 experiments","authors":"X.R. Duan, M. Xu, W.L. Zhong, X.Q. Ji, W. Chen, Z.B. Shi, X.L. Liu, B. Lu, B. Li, Y.Q. Wang, J.Q. Li, G.Y. Zheng, Yong Liu, Q.W. Yang, L.W. Yan, L.J. Cai, Q. Li, Y. Liu, X.Y. Bai, Z. Cao, X. Chen, H.T. Chen, Y.H. Chen, G.Q. Dong, H.L. Du, D.M. Fan, J.M. Gao, S.F. Geng, G.Z. Hao, H.M. He, M. Huang, M. Jiang, R. Ke, A.S. Liang, J.X. Li, Qing Li, Yongge Li, L.C. Li, H.J. Li, W.B. Li, D.Q. Liu, T. Long, L.F. Lu, L. Nie, P.W. Shi, J.F. Peng, A.P. Sun, T.F. Sun, R.H. Tong, H.L. Wei, S. Wang, G.L. Xiao, X.P. Xiao, L. Xue, H.B. Xu, Z.Y. Yang, D.L. Yu, L.M. Yu, Y.P. Zhang, X. Zheng, L. Zhang, Y. Zhang, F. Zhang, X.L. Zhang, HL-3 Team & Collaborators2345678910111213141516171819","doi":"10.1088/1741-4326/ad6e9e","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6e9e","url":null,"abstract":"Since the first plasma realized in 2020, a series of key systems on HL-3 (known as HL-2M before) tokamak have been equipped/upgraded, including in-vessel components (the first wall, lower divertor, and toroidal cryogenic/water-cooling/baking/glow discharge systems, etc.), auxiliary heating system of 11 MW, and 28 diagnostic systems (to measure the plasma density, electron temperature, radiation, magnetic field, etc.). Magnet field systems were commissioned firstly for divertor plasma discharges. During the 2nd experimental campaign of HL-3 tokamak, several great progresses have been achieved. Firstly, the successful operation with plasma current larger than 1 MA was achieved under a divertor configuration. Secondly, the advanced divertor concept with two distinct snowflake configurations was realized. It is found that the distribution of ion saturation current and heat flux on bottom plate becomes wide due to magnetic surface expansion, demonstrating the advantage of such configuration in the heat flux mitigation. In addition, using the combination of NBI, ECRH and LHCD, the standard sawtoothing high confinement mode of megampere plasma was firstly accessed on the HL-3. The successful commissioning of HL-3 is beneficial for the initial operation of ITER.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"31 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212123","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 : 2024-08-30DOI: 10.1088/1741-4326/ad7159
Xiujie Zhang, Yao Zhao, Zhenchao Sun, Lei Wang, Xinting Lv
In most designs of liquid metal blankets, the reduced activation ferritic martensitic steel with high relative magnetic permeability is proposed as the structural wall material, which will have an obvious influence on the magnetic field distribution inside the duct and consequently modify the liquid metal magnetohydrodynamics (MHD) flow state. However, the MHD flow state considering the influence of the ferromagnetic wall is lack of systematic investigations especially under the relevant conditions of magnetic confinement fusion reactors. In this work, systematic investigations on the ferromagnetic MHD effect are conducted by experiments and numerical simulations considering the relevant condition of fusion reactors such as high magnetic fields up to 10 T and the actual magnetic permeability of ferromagnetic walls. It is found that magnetic field lines are mainly gathered through the side wall for ferromagnetic rectangular ducts, which will result in the overall magnetic shielding effect. As applied magnetic fields increase, the magnetic shielding effect weakens, increasing the aspect ratio and wall thickness of the duct is benefit to enhance the overall magnetic shielding effect. A slightly magnetic strengthening effect is firstly observed in our experimental and numerical investigations, which is characterized that the average magnetic flux intensity in the fluid region is slightly greater than applied magnetic fields and the pressure drop in ferromagnetic ducts is also higher than that in non-ferromagnetic ducts when the applied magnetic field is bigger than the threshold of transition. The dimensionless pressure gradient in ferromagnetic rectangular ducts usually increases firstly and then decreases with the increase of applied magnetic fields, the pressure drop estimated from the coefficient of the square of the average magnetic flux intensity in the fluid region is generally accurate with exceptions in some extreme conditions. These findings will provide a theoretical guidance for future liquid metal blanket designs.
在大多数液态金属毯的设计中，具有高相对磁导率的还原活化铁素体马氏体钢被建议作为结构壁材料，这将对管道内的磁场分布产生明显影响，并进而改变液态金属磁流体力学（MHD）流态。然而，考虑到铁磁壁影响的 MHD 流动状态还缺乏系统研究，尤其是在磁约束聚变反应堆的相关条件下。在这项工作中，考虑到核聚变反应堆的相关条件，如高达 10 T 的磁场和铁磁壁的实际磁导率，通过实验和数值模拟对铁磁 MHD 效应进行了系统研究。研究发现，磁场线主要通过铁磁性矩形管道的侧壁聚集，从而产生整体磁屏蔽效应。随着外加磁场的增加，磁屏蔽效应减弱，增加风道的长宽比和壁厚有利于增强整体磁屏蔽效应。在我们的实验和数值研究中，首先观察到了轻微的磁强化效应，其特征是流体区域的平均磁通强度略大于外加磁场，当外加磁场大于过渡阈值时，铁磁性风道中的压降也高于非铁磁性风道中的压降。铁磁性矩形风道中的无量纲压力梯度通常随着外加磁场的增加而先增大后减小，根据流体区域平均磁通强度平方系数估算的压降基本准确，但在某些极端条件下例外。这些发现将为未来的液态金属毯设计提供理论指导。
{"title":"MHD flows through ferromagnetic rectangular ducts in liquid metal blankets","authors":"Xiujie Zhang, Yao Zhao, Zhenchao Sun, Lei Wang, Xinting Lv","doi":"10.1088/1741-4326/ad7159","DOIUrl":"https://doi.org/10.1088/1741-4326/ad7159","url":null,"abstract":"In most designs of liquid metal blankets, the reduced activation ferritic martensitic steel with high relative magnetic permeability is proposed as the structural wall material, which will have an obvious influence on the magnetic field distribution inside the duct and consequently modify the liquid metal magnetohydrodynamics (MHD) flow state. However, the MHD flow state considering the influence of the ferromagnetic wall is lack of systematic investigations especially under the relevant conditions of magnetic confinement fusion reactors. In this work, systematic investigations on the ferromagnetic MHD effect are conducted by experiments and numerical simulations considering the relevant condition of fusion reactors such as high magnetic fields up to 10 T and the actual magnetic permeability of ferromagnetic walls. It is found that magnetic field lines are mainly gathered through the side wall for ferromagnetic rectangular ducts, which will result in the overall magnetic shielding effect. As applied magnetic fields increase, the magnetic shielding effect weakens, increasing the aspect ratio and wall thickness of the duct is benefit to enhance the overall magnetic shielding effect. A slightly magnetic strengthening effect is firstly observed in our experimental and numerical investigations, which is characterized that the average magnetic flux intensity in the fluid region is slightly greater than applied magnetic fields and the pressure drop in ferromagnetic ducts is also higher than that in non-ferromagnetic ducts when the applied magnetic field is bigger than the threshold of transition. The dimensionless pressure gradient in ferromagnetic rectangular ducts usually increases firstly and then decreases with the increase of applied magnetic fields, the pressure drop estimated from the coefficient of the square of the average magnetic flux intensity in the fluid region is generally accurate with exceptions in some extreme conditions. These findings will provide a theoretical guidance for future liquid metal blanket designs.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"64 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212128","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 : 2024-08-30DOI: 10.1088/1741-4326/ad6ea2
E. Tholerus, F.J. Casson, S.P. Marsden, T. Wilson, D. Brunetti, P. Fox, S.J. Freethy, T.C. Hender, S.S. Henderson, A. Hudoba, K.K. Kirov, F. Koechl, H. Meyer, S.I. Muldrew, C. Olde, B.S. Patel, C.M. Roach, S. Saarelma, G. Xia, the STEP team1
STEP is a spherical tokamak prototype power plant that is being designed to demonstrate net electric power. The design phase involves the exploitation of plasma models to optimise fusion performance subject to satisfying various physics and engineering constraints. A modelling workflow, including integrated core plasma modelling, MHD stability analysis, SOL and pedestal modelling, coil set and free boundary equilibrium solvers, and whole plant design, has been developed to specify the design parameters and to develop viable scenarios. The integrated core plasma model JETTO is used to develop individual flat-top operating points that satisfy imposed criteria for fusion power performance within operational constraints. Key plasma parameters such as normalised beta, Greenwald density fraction, auxiliary power and radiated power have been scanned to scope the operational space and to derive a collection of candidate non-inductive flat-top points. The assumed auxiliary heating and current drive is either from electron cyclotron (EC) systems only or a combination of EC and electron Bernstein waves. At present stages of transport modelling, there is a large uncertainty in overall confinement for relevant parameter regimes. For each of the two auxiliary heating and current drive systems scenarios, two candidate flat-top points have been developed based on different confinement assumptions, totalling to four operating points. A lower confinement assumption generally suggests operating points in high-density, high auxiliary power regimes, whereas higher confinement would allow access to a broader parameter regime in density and power while maintaining target fusion power performance.
{"title":"Flat-top plasma operational space of the STEP power plant","authors":"E. Tholerus, F.J. Casson, S.P. Marsden, T. Wilson, D. Brunetti, P. Fox, S.J. Freethy, T.C. Hender, S.S. Henderson, A. Hudoba, K.K. Kirov, F. Koechl, H. Meyer, S.I. Muldrew, C. Olde, B.S. Patel, C.M. Roach, S. Saarelma, G. Xia, the STEP team1","doi":"10.1088/1741-4326/ad6ea2","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6ea2","url":null,"abstract":"STEP is a spherical tokamak prototype power plant that is being designed to demonstrate net electric power. The design phase involves the exploitation of plasma models to optimise fusion performance subject to satisfying various physics and engineering constraints. A modelling workflow, including integrated core plasma modelling, MHD stability analysis, SOL and pedestal modelling, coil set and free boundary equilibrium solvers, and whole plant design, has been developed to specify the design parameters and to develop viable scenarios. The integrated core plasma model JETTO is used to develop individual flat-top operating points that satisfy imposed criteria for fusion power performance within operational constraints. Key plasma parameters such as normalised beta, Greenwald density fraction, auxiliary power and radiated power have been scanned to scope the operational space and to derive a collection of candidate non-inductive flat-top points. The assumed auxiliary heating and current drive is either from electron cyclotron (EC) systems only or a combination of EC and electron Bernstein waves. At present stages of transport modelling, there is a large uncertainty in overall confinement for relevant parameter regimes. For each of the two auxiliary heating and current drive systems scenarios, two candidate flat-top points have been developed based on different confinement assumptions, totalling to four operating points. A lower confinement assumption generally suggests operating points in high-density, high auxiliary power regimes, whereas higher confinement would allow access to a broader parameter regime in density and power while maintaining target fusion power performance.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"21 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212126","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 : 2024-08-30DOI: 10.1088/1741-4326/ad70ca
Xiaohui Ji, Zhibin Guo, Yi Zhang
We study a reduced electrostatic fluid model for the tokamak scrape-off layer, which incorporates temperature gradient and vorticity gradient as two free energy fields. Two scenarios of field coupling are addressed: (1) sheath condition; (2) vortex wave coupling. For the sheath condition induced field coupling, the poloidal E×B flow shear is coupled with the temperature gradient. Combining an eigenmode analysis and the nonlinear phase dynamics approach, our findings indicate that in the absence of a vorticity gradient, the overall effect of the sheath condition induced flow shear can either stabilize or destabilize the interchange mode, depending on the competition between the flow shear suppression and the temperature gradient driving. This is different from the case where the gradient drive and shear damping are decoupled. When the field coupling is mediated by wave interactions, by setting an idealized step-like temperature and vorticity profiles, a joint mode forms through resonant interaction between the interfacial waves driven by the temperature and vorticity gradients, respectively. Near the phase locking condition, the joint mode can be more unstable than pure temperature gradient driven mode.
{"title":"Multi-field coupling in the scrape-off layer of tokamak plasma","authors":"Xiaohui Ji, Zhibin Guo, Yi Zhang","doi":"10.1088/1741-4326/ad70ca","DOIUrl":"https://doi.org/10.1088/1741-4326/ad70ca","url":null,"abstract":"We study a reduced electrostatic fluid model for the tokamak scrape-off layer, which incorporates temperature gradient and vorticity gradient as two free energy fields. Two scenarios of field coupling are addressed: (1) sheath condition; (2) vortex wave coupling. For the sheath condition induced field coupling, the poloidal <inline-formula>\u0000<tex-math><?CDATA $mathbf{E} times mathbf{B}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:mrow><mml:mi mathvariant=\"bold\">E</mml:mi></mml:mrow><mml:mo>×</mml:mo><mml:mrow><mml:mi mathvariant=\"bold\">B</mml:mi></mml:mrow></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad70caieqn1.gif\"></inline-graphic></inline-formula> flow shear is coupled with the temperature gradient. Combining an eigenmode analysis and the nonlinear phase dynamics approach, our findings indicate that in the absence of a vorticity gradient, the overall effect of the sheath condition induced flow shear can either stabilize or destabilize the interchange mode, depending on the competition between the flow shear suppression and the temperature gradient driving. This is different from the case where the gradient drive and shear damping are decoupled. When the field coupling is mediated by wave interactions, by setting an idealized step-like temperature and vorticity profiles, a joint mode forms through resonant interaction between the interfacial waves driven by the temperature and vorticity gradients, respectively. Near the phase locking condition, the joint mode can be more unstable than pure temperature gradient driven mode.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"78 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212129","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 : 2024-08-30DOI: 10.1088/1741-4326/ad6c5e
L. Cappelli, N. Fedorczak, E. Serre
A steady-state, 1D semi-analytical model for prompt redeposition based on the separation between redeposition caused by the electric field in the sheath and redeposition related to gyromotion is here described. The model allows for the estimation of not only the fraction of promptly redeposited flux but also the energy and angular distribution of the non-promptly redeposited population, along with their average charge state. Thus, the temperature and mean parallel-to-B velocity of the non-promptly redeposited flux are also available. The semi-analytical model was validated against equivalent Monte Carlo simulations across a broad range of input parameters. In this paper the eroded material under exam was tungsten (W) for which the code demonstrated consistent agreement with respect to numerical results, within its defined validity limits. The model can theoretically provide a solution for any material, temperature and electron density profile in the sheath, monotonic potential drop profile, and sputtered particles energy and angular distribution at the wall. As such, this code emerges as a potential tool for addressing the boundary redeposition phenomenon in fluid impurity transport simulations.
{"title":"Semi-analytical modeling of prompt redeposition in a steady-state plasma","authors":"L. Cappelli, N. Fedorczak, E. Serre","doi":"10.1088/1741-4326/ad6c5e","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6c5e","url":null,"abstract":"A steady-state, 1D semi-analytical model for prompt redeposition based on the separation between redeposition caused by the electric field in the sheath and redeposition related to gyromotion is here described. The model allows for the estimation of not only the fraction of promptly redeposited flux but also the energy and angular distribution of the non-promptly redeposited population, along with their average charge state. Thus, the temperature and mean parallel-to-B velocity of the non-promptly redeposited flux are also available. The semi-analytical model was validated against equivalent Monte Carlo simulations across a broad range of input parameters. In this paper the eroded material under exam was tungsten (W) for which the code demonstrated consistent agreement with respect to numerical results, within its defined validity limits. The model can theoretically provide a solution for any material, temperature and electron density profile in the sheath, monotonic potential drop profile, and sputtered particles energy and angular distribution at the wall. As such, this code emerges as a potential tool for addressing the boundary redeposition phenomenon in fluid impurity transport simulations.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"64 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212125","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 : 2024-08-30DOI: 10.1088/1741-4326/ad6c61
J. Decker, M. Hoppe, U. Sheikh, B.P. Duval, G. Papp, L. Simons, T. Wijkamp, J. Cazabonne, S. Coda, E. Devlaminck, O. Ficker, R. Hellinga, U. Kumar, Y. Savoye-Peysson, L. Porte, C. Reux, C. Sommariva, A. Tema Biwolé, B. Vincent, L. Votta, the TCV Team, the EUROfusion Tokamak Exploitation Teamb
Runaway electrons (REs) are a concern for tokamak fusion reactors from discharge startup to termination. A sudden localized loss of a multi-megaampere RE beam can inflict severe damage to the first wall. Should a disruption occur, the existence of a RE seed may play a significant role in the formation of a RE beam and the magnitude of its current. The application of central electron cyclotron resonance heating (ECRH) in the Tokamak à Configuration Variable (TCV) reduces an existing RE seed population by up to three orders of magnitude within only a few hundred milliseconds. Applying ECRH before a disruption can also prevent the formation of a post-disruption RE beam in TCV where it would otherwise be expected. The RE expulsion rate and consequent RE current reduction are found to increase with applied ECRH power. Whereas central ECRH is effective in expelling REs, off-axis ECRH has a comparatively limited effect. A simple 0-D model for the evolution of the RE population is presented that explains how the effective ECRH-induced RE expulsion results from the combined effects of increased electron temperature and enhanced RE transport.
失控电子（REs）是托卡马克聚变反应堆从放电启动到终止的一个问题。数百万安培的失控电子束的突然局部消失会对第一壁造成严重破坏。一旦发生中断，RE 种子的存在可能会对 RE 束的形成及其电流大小产生重要影响。在托卡马克 à 配置变量（TCV）中应用中央电子回旋共振加热（ECRH）可在短短几百毫秒内将现有的 RE 种子数量减少多达三个数量级。在中断前应用 ECRH 还能防止在 TCV 中形成预期的中断后 RE 束。研究发现，随着 ECRH 功率的增加，RE 驱逐率和随之而来的 RE 电流降低率也会增加。中心 ECRH 能有效驱逐 RE，而偏轴 ECRH 的效果相对有限。本文提出了一个简单的 RE 群体演化 0-D 模型，解释了电子温度升高和 RE 传输增强的共同作用如何导致有效的 ECRH 诱导 RE 驱逐。
{"title":"Expulsion of runaway electrons using ECRH in the TCV tokamak","authors":"J. Decker, M. Hoppe, U. Sheikh, B.P. Duval, G. Papp, L. Simons, T. Wijkamp, J. Cazabonne, S. Coda, E. Devlaminck, O. Ficker, R. Hellinga, U. Kumar, Y. Savoye-Peysson, L. Porte, C. Reux, C. Sommariva, A. Tema Biwolé, B. Vincent, L. Votta, the TCV Team, the EUROfusion Tokamak Exploitation Teamb","doi":"10.1088/1741-4326/ad6c61","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6c61","url":null,"abstract":"Runaway electrons (REs) are a concern for tokamak fusion reactors from discharge startup to termination. A sudden localized loss of a multi-megaampere RE beam can inflict severe damage to the first wall. Should a disruption occur, the existence of a RE seed may play a significant role in the formation of a RE beam and the magnitude of its current. The application of central electron cyclotron resonance heating (ECRH) in the Tokamak à Configuration Variable (TCV) reduces an existing RE seed population by up to three orders of magnitude within only a few hundred milliseconds. Applying ECRH before a disruption can also prevent the formation of a post-disruption RE beam in TCV where it would otherwise be expected. The RE expulsion rate and consequent RE current reduction are found to increase with applied ECRH power. Whereas central ECRH is effective in expelling REs, off-axis ECRH has a comparatively limited effect. A simple 0-D model for the evolution of the RE population is presented that explains how the effective ECRH-induced RE expulsion results from the combined effects of increased electron temperature and enhanced RE transport.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"178 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212127","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 : 2024-08-29DOI: 10.1088/1741-4326/ad6ba7
S.A.M. McNamara, A. Alieva, M.S. Anastopoulos Tzanis, O. Asunta, J. Bland, H. Bohlin, P.F. Buxton, C. Colgan, A. Dnestrovskii, E. du Toit, M. Fontana, M. Gemmell, M.P. Gryaznevich, J. Hakosalo, M.R. Hardman, D. Harryman, D. Hoffman, M. Iliasova, S. Janhunen, F. Janky, J.B. Lister, H.F. Lowe, E. Maartensson, C. Marsden, S.Y. Medvedev, S.R. Mirfayzi, M. Moscheni, G. Naylor, V. Nemytov, J. Njau, T. O’Gorman, D. Osin, T. Pyragius, A. Rengle, M. Romanelli, C. Romero, M. Sertoli, V. Shevchenko, J. Sinha, A. Sladkomedova, S. Sridhar, J. Stirling, Y. Takase, P.R. Thomas, J. Varje, E. Vekshina, B. Vincent, H.V. Willett, J. Wood, E. Wooldridge, D. Zakhar, X. Zhang, D. Battaglia, N. Bertelli, P.J. Bonofiglo, L.F. Delgado-Aparicio, V.N. Duarte, N.N. Gorelenkov, M. de Haas, S.M. Kaye, R. Maingi, D. Mueller, M. Ono, M. Podesta, Y. Ren, S. Trieu, E. Delabie, T.K. Gray, B. Lomanowski, E.A. Unterberg, O. Marchuk, the ST40 Team1
ST40 is a compact, high-field (BT0⩽2.1T) spherical tokamak (ST) with a mission to expand the physics and technology basis for the ST route to commercial fusion. The ST40 research programme covers confinement and stability; solenoid-free start-up; high-performance operating scenarios; and plasma exhaust. In 2022, ST40 obtained central deuterium ion temperatures of 9.6±0.4keV, demonstrating for the first time that pilot plant relevant ion temperatures can be reached in a compact, high-field ST. Analysis of these high-ion temperature plasmas is presented, including a summary of confinement, transport and microstability characteristics, and energetic particle instabilities. Recent scenario development activities have focused on establishing diverted H-mode plasmas across a range of toroidal fields and plasma currents, along with scenarios with high non-inductive current fractions. In future operations, beginning in 2025, a 1 MW dual frequency (104/137 GHz) electron cyclotron (EC) system will be installed to enable the study of EC and electron Bernstein wave plasma start-up and current drive. Predictive modelling of the potential performance of these systems is presented.
ST40 是一个结构紧凑的高磁场（BT0⩽2.1T）球形托卡马克（ST），其任务是为商业核聚变的 ST 路线扩大物理和技术基础。ST40 研究计划涵盖约束和稳定性、无螺线管启动、高性能运行方案和等离子体排气。2022 年，ST40 获得了 9.6±0.4 千伏的中心氘离子温度，首次证明了在紧凑型高场 ST 中可以达到与试验工厂相关的离子温度。本文介绍了对这些高离子温度等离子体的分析，包括对约束、传输和微稳定性特征以及高能粒子不稳定性的总结。最近的方案开发活动侧重于在一系列环形场和等离子体电流范围内建立分流 H 模式等离子体，以及具有高非感应电流分数的方案。在未来的运行中，从 2025 年开始，将安装一个 1 兆瓦的双频（104/137 千兆赫）电子回旋加速器（EC）系统，以研究 EC 和电子伯恩斯坦波等离子体的启动和电流驱动。介绍了这些系统潜在性能的预测模型。
{"title":"Overview of recent results from the ST40 compact high-field spherical tokamak","authors":"S.A.M. McNamara, A. Alieva, M.S. Anastopoulos Tzanis, O. Asunta, J. Bland, H. Bohlin, P.F. Buxton, C. Colgan, A. Dnestrovskii, E. du Toit, M. Fontana, M. Gemmell, M.P. Gryaznevich, J. Hakosalo, M.R. Hardman, D. Harryman, D. Hoffman, M. Iliasova, S. Janhunen, F. Janky, J.B. Lister, H.F. Lowe, E. Maartensson, C. Marsden, S.Y. Medvedev, S.R. Mirfayzi, M. Moscheni, G. Naylor, V. Nemytov, J. Njau, T. O’Gorman, D. Osin, T. Pyragius, A. Rengle, M. Romanelli, C. Romero, M. Sertoli, V. Shevchenko, J. Sinha, A. Sladkomedova, S. Sridhar, J. Stirling, Y. Takase, P.R. Thomas, J. Varje, E. Vekshina, B. Vincent, H.V. Willett, J. Wood, E. Wooldridge, D. Zakhar, X. Zhang, D. Battaglia, N. Bertelli, P.J. Bonofiglo, L.F. Delgado-Aparicio, V.N. Duarte, N.N. Gorelenkov, M. de Haas, S.M. Kaye, R. Maingi, D. Mueller, M. Ono, M. Podesta, Y. Ren, S. Trieu, E. Delabie, T.K. Gray, B. Lomanowski, E.A. Unterberg, O. Marchuk, the ST40 Team1","doi":"10.1088/1741-4326/ad6ba7","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6ba7","url":null,"abstract":"ST40 is a compact, high-field (<inline-formula>\u0000<tex-math><?CDATA $B_{mathrm{T0}}unicode{x2A7D} 2.1,,,textrm{T}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:msub><mml:mi>B</mml:mi><mml:mrow><mml:mrow><mml:mi mathvariant=\"normal\">T</mml:mi><mml:mn>0</mml:mn></mml:mrow></mml:mrow></mml:msub><mml:mtext>⩽</mml:mtext><mml:mn>2.1</mml:mn><mml:mstyle scriptlevel=\"0\"></mml:mstyle><mml:mstyle scriptlevel=\"0\"></mml:mstyle><mml:mstyle scriptlevel=\"0\"></mml:mstyle><mml:mtext>T</mml:mtext></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad6ba7ieqn1.gif\"></inline-graphic></inline-formula>) spherical tokamak (ST) with a mission to expand the physics and technology basis for the ST route to commercial fusion. The ST40 research programme covers confinement and stability; solenoid-free start-up; high-performance operating scenarios; and plasma exhaust. In 2022, ST40 obtained central deuterium ion temperatures of <inline-formula>\u0000<tex-math><?CDATA $9.6 pm 0.4 textrm{keV}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:mn>9.6</mml:mn><mml:mo>±</mml:mo><mml:mn>0.4</mml:mn><mml:mtext> </mml:mtext><mml:mtext>keV</mml:mtext></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad6ba7ieqn2.gif\"></inline-graphic></inline-formula>, demonstrating for the first time that pilot plant relevant ion temperatures can be reached in a compact, high-field ST. Analysis of these high-ion temperature plasmas is presented, including a summary of confinement, transport and microstability characteristics, and energetic particle instabilities. Recent scenario development activities have focused on establishing diverted H-mode plasmas across a range of toroidal fields and plasma currents, along with scenarios with high non-inductive current fractions. In future operations, beginning in 2025, a 1 MW dual frequency (104/137 GHz) electron cyclotron (EC) system will be installed to enable the study of EC and electron Bernstein wave plasma start-up and current drive. Predictive modelling of the potential performance of these systems is presented.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"46 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142211997","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 : 2024-08-29DOI: 10.1088/1741-4326/ad6011
J.R. Harrison, A. Aboutaleb, S. Ahmed, M. Aljunid, S.Y. Allan, H. Anand, Y. Andrew, L.C. Appel, A. Ash, J. Ashton, O. Bachmann, M. Barnes, B. Barrett, D. Baver, D. Beckett, J. Bennett, J. Berkery, M. Bernert, W. Boeglin, C. Bowman, J. Bradley, D. Brida, P.K. Browning, D. Brunetti, P. Bryant, J. Bryant, J. Buchanan, N. Bulmer, A. Carruthers, M. Cecconello, Z.P. Chen, J. Clark, C. Cowley, M. Coy, N. Crocker, G. Cunningham, I. Cziegler, T. Da Assuncao, Y. Damizia, P. Davies, I.E. Day, G.L. Derks, S. Dixon, R. Doyle, M. Dreval, M. Dunne, B.P. Duval, T. Eagles, J. Edmond, H. El-Haroun, S.D. Elmore, Y. Enters, M. Faitsch, F. Federici, N. Fedorczak, F. Felici, A.R. Field, M. Fitzgerald, I. Fitzgerald, R. Fitzpatrick, L. Frassinetti, W. Fuller, D. Gahle, J. Galdon-Quiroga, L. Garzotti, S. Gee, T. Gheorghiu, S. Gibson, K.J. Gibson, C. Giroud, D. Greenhouse, V.H. Hall-Chen, C.J. Ham, R. Harrison, S.S. Henderson, C. Hickling, B. Hnat, L. Howlett, J. Hughes, R. Hussain, K. Imada, P. Jacquet, P. Jepson, B. Kandan, I. Katramados, Y.O. Kazakov, D. King, R. King, A. Kirk, M. Knolker, M. Kochan, L. Kogan, B. Kool, M. Kotschenreuther, M. Lees, A.W. Leonard, G. Liddiard, B. Lipschultz, Y.Q. Liu, B.A. Lomanowski, N. Lonigro, J. Lore, J. Lovell, S. Mahajan, F. Maiden, C. Man-Friel, F. Mansfield, S. Marsden, R. Martin, S. Mazzi, R. McAdams, G. McArdle, K.G. McClements, J. McClenaghan, D. McConville, K. McKay, C. McKnight, P. McKnight, A. McLean, B.F. McMillan, A. McShee, J. Measures, N. Mehay, C.A. Michael, F. Militello, D. Morbey, S. Mordijck, D. Moulton, O. Myatra, A.O. Nelson, M. Nicassio, M.G. O’Mullane, H.J.C. Oliver, P. Ollus, T. Osborne, N. Osborne, E. Parr, B. Parry, B.S. Patel, D. Payne, C. Paz-Soldan, A. Phelps, L. Piron, C. Piron, G. Prechel, M. Price, B. Pritchard, R. Proudfoot, H. Reimerdes, T. Rhodes, P. Richardson, J. Riquezes, J.F. Rivero-Rodriguez, C.M. Roach, M. Robson, K. Ronald, E. Rose, P. Ryan, D. Ryan, S. Saarelma, S. Sabbagh, R. Sarwar, P. Saunders, O. Sauter, R. Scannell, T. Schuett, R. Seath, R. Sharma, P. Shi, B. Sieglin, M. Simmonds, J. Smith, A. Smith, V. A. Soukhanovskii, D. Speirs, G. Staebler, R. Stephen, P. Stevenson, J. Stobbs, M. Stott, C. Stroud, C. Tame, C. Theiler, N. Thomas-Davies, A.J. Thornton, M. Tobin, M. Vallar, R.G.L. Vann, L. Velarde, K. Verhaegh, E. Viezzer, C. Vincent, G. Voss, M. Warr, W. Wehner, S. Wiesen, T.A. Wijkamp, D. Wilkins, T. Williams, T. Wilson, H.R. Wilson, H. Wong, M. Wood, V. Zamkovska
Recent results from MAST Upgrade are presented, emphasising understanding the capabilities of this new device and deepening understanding of key physics issues for the operation of ITER and the design of future fusion power plants. The impact of MHD instabilities on fast ion confinement have been studied, including the first observation of fast ion losses correlated with Compressional and Global Alfvén Eigenmodes. High-performance plasma scenarios have been developed by tailoring the early plasma current ramp phase to avoid internal reconnection events, resulting in a more monotonic q profile with low central shear. The impact of m/n = 3/2, 2/1 and 1/1 modes on thermal plasma confinement and rotation profiles has been quantified, and scenarios optimised to avoid them have transiently reached values of normalised beta approaching 4.2. In pedestal and ELM physics, a maximum pedestal top temperature of ∼350 eV has been achieved, exceeding the value achieved on MAST at similar heating power. Mitigation of type-I ELMs with n = 1 RMPs has been observed. Studies of plasma exhaust have concentrated on comparing conventional and Super-X divertor configurations, while X-point target, X-divertor and snowflake configurations have been developed and studied in parallel. In L-mode discharges, the separatrix density required to detach the outer divertors is approximately a factor 2 lower in the Super-X than the conventional configuration, in agreement with simulations. Detailed analysis of spectroscopy data from studies of the Super-X configuration reveal the importance of including plasma-molecule interactions and D2 Fulcher band emission to properly quantify the rates of ionisation, plasma-molecule interactions and volumetric recombination processes governing divertor detachment. In H-mode with conventional and Super-X configurations, the outer divertors are attached in the former and detached in the latter with no impact on core or pedestal confinement.
介绍了 MAST 升级的最新成果，强调了解这一新装置的能力，加深对热核实验堆运行和未来聚变发电厂设计的关键物理问题的理解。研究了 MHD 不稳定性对快速离子约束的影响，包括首次观测到与压缩和全局阿尔费文特征模相关的快速离子损失。通过调整早期等离子体电流斜坡阶段以避免内部再连接事件，从而产生了具有低中心剪切力的更单调的q曲线，开发出了高性能等离子体方案。m/n = 3/2、2/1 和 1/1 模式对热等离子体约束和旋转剖面的影响已被量化，为避免它们而优化的方案已瞬时达到接近 4.2 的归一化贝塔值。在基座和ELM物理学方面，基座最高温度达到了350 eV，超过了MAST在类似加热功率下达到的数值。还观察到 n = 1 RMP 对 I 型 ELM 的缓解作用。对等离子体排气的研究主要集中在传统和超 X 分流器配置的比较上，同时还开发和研究了 X 点目标、X 分流器和雪花配置。在 L 模式放电中，分离外部分流器所需的分离矩阵密度在 Super-X 配置中比传统配置低约 2 倍，这与模拟结果一致。对超级-X 配置研究的光谱数据进行的详细分析表明，要正确量化电离率、等离子体-分子相互作用和分流器分离的体积重组过程，就必须包括等离子体-分子相互作用和 D2 富彻带发射。在具有传统配置和超级-X 配置的 H 模式中，前者的外部分流器附着，后者的外部分流器脱离，对核心或基座约束没有影响。
{"title":"Overview of physics results from MAST upgrade towards core-pedestal-exhaust integration","authors":"J.R. Harrison, A. Aboutaleb, S. Ahmed, M. Aljunid, S.Y. Allan, H. Anand, Y. Andrew, L.C. Appel, A. Ash, J. Ashton, O. Bachmann, M. Barnes, B. Barrett, D. Baver, D. Beckett, J. Bennett, J. Berkery, M. Bernert, W. Boeglin, C. Bowman, J. Bradley, D. Brida, P.K. Browning, D. Brunetti, P. Bryant, J. Bryant, J. Buchanan, N. Bulmer, A. Carruthers, M. Cecconello, Z.P. Chen, J. Clark, C. Cowley, M. Coy, N. Crocker, G. Cunningham, I. Cziegler, T. Da Assuncao, Y. Damizia, P. Davies, I.E. Day, G.L. Derks, S. Dixon, R. Doyle, M. Dreval, M. Dunne, B.P. Duval, T. Eagles, J. Edmond, H. El-Haroun, S.D. Elmore, Y. Enters, M. Faitsch, F. Federici, N. Fedorczak, F. Felici, A.R. Field, M. Fitzgerald, I. Fitzgerald, R. Fitzpatrick, L. Frassinetti, W. Fuller, D. Gahle, J. Galdon-Quiroga, L. Garzotti, S. Gee, T. Gheorghiu, S. Gibson, K.J. Gibson, C. Giroud, D. Greenhouse, V.H. Hall-Chen, C.J. Ham, R. Harrison, S.S. Henderson, C. Hickling, B. Hnat, L. Howlett, J. Hughes, R. Hussain, K. Imada, P. Jacquet, P. Jepson, B. Kandan, I. Katramados, Y.O. Kazakov, D. King, R. King, A. Kirk, M. Knolker, M. Kochan, L. Kogan, B. Kool, M. Kotschenreuther, M. Lees, A.W. Leonard, G. Liddiard, B. Lipschultz, Y.Q. Liu, B.A. Lomanowski, N. Lonigro, J. Lore, J. Lovell, S. Mahajan, F. Maiden, C. Man-Friel, F. Mansfield, S. Marsden, R. Martin, S. Mazzi, R. McAdams, G. McArdle, K.G. McClements, J. McClenaghan, D. McConville, K. McKay, C. McKnight, P. McKnight, A. McLean, B.F. McMillan, A. McShee, J. Measures, N. Mehay, C.A. Michael, F. Militello, D. Morbey, S. Mordijck, D. Moulton, O. Myatra, A.O. Nelson, M. Nicassio, M.G. O’Mullane, H.J.C. Oliver, P. Ollus, T. Osborne, N. Osborne, E. Parr, B. Parry, B.S. Patel, D. Payne, C. Paz-Soldan, A. Phelps, L. Piron, C. Piron, G. Prechel, M. Price, B. Pritchard, R. Proudfoot, H. Reimerdes, T. Rhodes, P. Richardson, J. Riquezes, J.F. Rivero-Rodriguez, C.M. Roach, M. Robson, K. Ronald, E. Rose, P. Ryan, D. Ryan, S. Saarelma, S. Sabbagh, R. Sarwar, P. Saunders, O. Sauter, R. Scannell, T. Schuett, R. Seath, R. Sharma, P. Shi, B. Sieglin, M. Simmonds, J. Smith, A. Smith, V. A. Soukhanovskii, D. Speirs, G. Staebler, R. Stephen, P. Stevenson, J. Stobbs, M. Stott, C. Stroud, C. Tame, C. Theiler, N. Thomas-Davies, A.J. Thornton, M. Tobin, M. Vallar, R.G.L. Vann, L. Velarde, K. Verhaegh, E. Viezzer, C. Vincent, G. Voss, M. Warr, W. Wehner, S. Wiesen, T.A. Wijkamp, D. Wilkins, T. Williams, T. Wilson, H.R. Wilson, H. Wong, M. Wood, V. Zamkovska","doi":"10.1088/1741-4326/ad6011","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6011","url":null,"abstract":"Recent results from MAST Upgrade are presented, emphasising understanding the capabilities of this new device and deepening understanding of key physics issues for the operation of ITER and the design of future fusion power plants. The impact of MHD instabilities on fast ion confinement have been studied, including the first observation of fast ion losses correlated with Compressional and Global Alfvén Eigenmodes. High-performance plasma scenarios have been developed by tailoring the early plasma current ramp phase to avoid internal reconnection events, resulting in a more monotonic q profile with low central shear. The impact of <italic toggle=\"yes\">m</italic>/<italic toggle=\"yes\">n</italic> = 3/2, 2/1 and 1/1 modes on thermal plasma confinement and rotation profiles has been quantified, and scenarios optimised to avoid them have transiently reached values of normalised beta approaching 4.2. In pedestal and ELM physics, a maximum pedestal top temperature of ∼350 eV has been achieved, exceeding the value achieved on MAST at similar heating power. Mitigation of type-I ELMs with <italic toggle=\"yes\">n</italic> = 1 RMPs has been observed. Studies of plasma exhaust have concentrated on comparing conventional and Super-X divertor configurations, while X-point target, X-divertor and snowflake configurations have been developed and studied in parallel. In L-mode discharges, the separatrix density required to detach the outer divertors is approximately a factor 2 lower in the Super-X than the conventional configuration, in agreement with simulations. Detailed analysis of spectroscopy data from studies of the Super-X configuration reveal the importance of including plasma-molecule interactions and D<sub>2</sub> Fulcher band emission to properly quantify the rates of ionisation, plasma-molecule interactions and volumetric recombination processes governing divertor detachment. In H-mode with conventional and Super-X configurations, the outer divertors are attached in the former and detached in the latter with no impact on core or pedestal confinement.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"28 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212149","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 : 2024-08-28DOI: 10.1088/1741-4326/ad70cb
Yueqiang Liu, Xue Bai, Y. Gribov, F. Koechl, A. Loarte, S.D. Pinches, L. Schmitz
A systematic calculation is performed on the ripple-induced neoclassical toroidal viscous (NTV) torque for new ITER scenarios designed for the Augmented-First Plasma (A-FP) operation phase with the full tungsten wall, where the plasma-wall gap is varied in view of mitigating the impact of tungsten wall-plasma interactions. The torque calculation includes drift kinetic response of the plasma thermal and energetic particles to the n = 18 (n is the toroidal harmonic number) ripple field. For the plasma scenario with ~45 cm plasma-wall gap at the outboard mid-plane and considering the corrected ripple level of 0.17% by the ferritic steel inserts, the computed net NTV torque acting on the plasma column is in the sub-Nm level. However, with decreasing the plasma-wall gap, the computed net NTV torque can reach a level comparable to that produced by the neutral-beam momentum injection in ITER. Ripple correction by ferritic inserts reduces the net torque by a factor of 3.3 for all the three A-FP scenarios considered. The nωd=lωb (with ωd and ωb being the toroidal precession and bounce frequencies of trapped particles, respectively, and l an integer number) type of resonance-enhancement of the NTV torque, due to thermal particles, is found to be weak in ITER despite high-n of 18. The same also holds for the ITER 10 MA steady state scenario from the D-T operation phase, where the aforementioned resonance associated with fusion-born alphas is also included. The ripple-induced NTV torque is well below that produced by the resonant magnetic perturbation applied for controlling the type-I edge-localized mode in ITER.
针对为全钨壁增强第一等离子体（A-FP）运行阶段设计的新热核实验堆方案，对波纹引起的新古典环形粘性（NTV）扭矩进行了系统计算。扭矩计算包括等离子体热粒子和高能粒子对 n = 18（n 为环形谐波数）波纹场的漂移动力学响应。在等离子体外侧中平面的等离子体壁间隙约为 45 厘米的情况下，考虑到铁素体钢插入件的校正波纹水平为 0.17%，计算出的作用于等离子体柱的净 NTV 扭矩处于亚 Nm 水平。然而，随着等离子体壁间隙的减小，计算出的净 NTV 扭矩可以达到与热核实验堆中中性束动量注入所产生的扭矩相当的水平。在考虑的所有三种 A-FP 方案中，通过铁素体插件进行波纹校正可将净扭矩降低 3.3 倍。nωd=lωb（ωd 和 ωb 分别为被困粒子的环向前驱和反弹频率，l 为整数）是热粒子导致的 NTV 扭矩共振增强类型，尽管 ITER 中的 n 高达 18，但这种共振增强很弱。同样的情况也适用于从 D-T 运行阶段开始的热核实验堆 10 MA 稳态方案，其中也包括上述与核聚变产生的字母相关的共振。纹波引起的 NTV 扭矩远低于用于控制热核实验堆中 I 型边缘定位模式的共振磁扰动所产生的 NTV 扭矩。
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