Pub Date : 2024-09-04DOI: 10.1088/1741-4326/ad703b
A.S. Moore, L. Divol, B. Bachmann, R. Bionta, D. Bradley, D.T. Casey, P. Celliers, H. Chen, A. Do, E. Dewald, M. Eckart, D. Fittinghoff, J. Frenje, M. Gatu-Johnson, H. Geppert-Kleinrath, V. Geppert-Kleinrath, G. Grim, K. Hahn, M. Hohenberger, J. Holder, O. Hurricane, N. Izumi, S. Kerr, S.F. Khan, J.D. Kilkenny, Y. Kim, B. Kozioziemski, N. Lemos, A.G. MacPhee, P. Michel, M. Millot, K.D. Meaney, S. Nagel, A. Pak, J.E. Ralph, J.S. Ross, M.S. Rubery, D.J. Schlossberg, V. Smalyuk, G. Swadling, R. Tommasini, C. Trosseille, A.B. Zylstra, A. Mackinnon, J.D. Moody, O.L. Landen, R. Town
Fusion ignition by inertial confinement requires compression and heating of the fusion fuel to temperatures in excess of 5 keV and densities exceeding hundreds of g/cc. In August 2021, this scientific milestone was surpassed at the National Ignition Facility (NIF), when the Lawson criterion for ignition was exceeded generating 1.37MJ of fusion energy (Abu-Shawareb et al 2022 Phys. Rev. Lett.129 075001), and then in December 2022 target gain >1 was realized with the production of 3.1MJ of fusion energy from a target driven by 2.0MJ of laser energy (Abu-Shawareb et al 2024 Phys. Rev. Lett.132 065102). At the NIF, inertial confinement fusion research primarily uses a laser indirect drive in which the fusion capsule is surrounded by a high-Z enclosure (‘hohlraum’) used to convert the directed laser energy into a symmetric x-ray drive on the capsule. Precise measurements of the plasma conditions, x-rays, γ-rays and neutrons produced are key to understanding the pathway to higher performance. This paper discusses the diagnostics and measurement techniques developed to understand these experiments, focusing on three main topics: (1) key diagnostic developments for achieving igniting plasmas, (2) novel signatures related to thermonuclear burn and (3) advances to diagnostic capabilities in the igniting regime with a perspective toward developments for intertial fusion energy.
{"title":"Diagnosing inertial confinement fusion ignition","authors":"A.S. Moore, L. Divol, B. Bachmann, R. Bionta, D. Bradley, D.T. Casey, P. Celliers, H. Chen, A. Do, E. Dewald, M. Eckart, D. Fittinghoff, J. Frenje, M. Gatu-Johnson, H. Geppert-Kleinrath, V. Geppert-Kleinrath, G. Grim, K. Hahn, M. Hohenberger, J. Holder, O. Hurricane, N. Izumi, S. Kerr, S.F. Khan, J.D. Kilkenny, Y. Kim, B. Kozioziemski, N. Lemos, A.G. MacPhee, P. Michel, M. Millot, K.D. Meaney, S. Nagel, A. Pak, J.E. Ralph, J.S. Ross, M.S. Rubery, D.J. Schlossberg, V. Smalyuk, G. Swadling, R. Tommasini, C. Trosseille, A.B. Zylstra, A. Mackinnon, J.D. Moody, O.L. Landen, R. Town","doi":"10.1088/1741-4326/ad703b","DOIUrl":"https://doi.org/10.1088/1741-4326/ad703b","url":null,"abstract":"Fusion ignition by inertial confinement requires compression and heating of the fusion fuel to temperatures in excess of 5 keV and densities exceeding hundreds of g/cc. In August 2021, this scientific milestone was surpassed at the National Ignition Facility (NIF), when the Lawson criterion for ignition was exceeded generating 1.37MJ of fusion energy (Abu-Shawareb <italic toggle=\"yes\">et al</italic> 2022 <italic toggle=\"yes\">Phys. Rev. Lett.</italic> <bold>129</bold> 075001), and then in December 2022 target gain >1 was realized with the production of 3.1MJ of fusion energy from a target driven by 2.0MJ of laser energy (Abu-Shawareb <italic toggle=\"yes\">et al</italic> 2024 <italic toggle=\"yes\">Phys. Rev. Lett.</italic> <bold>132</bold> 065102). At the NIF, inertial confinement fusion research primarily uses a laser indirect drive in which the fusion capsule is surrounded by a high-Z enclosure (‘hohlraum’) used to convert the directed laser energy into a symmetric x-ray drive on the capsule. Precise measurements of the plasma conditions, x-rays, <italic toggle=\"yes\">γ</italic>-rays and neutrons produced are key to understanding the pathway to higher performance. This paper discusses the diagnostics and measurement techniques developed to understand these experiments, focusing on three main topics: (1) key diagnostic developments for achieving igniting plasmas, (2) novel signatures related to thermonuclear burn and (3) advances to diagnostic capabilities in the igniting regime with a perspective toward developments for intertial fusion energy.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"20 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212114","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 : 2024-09-03DOI: 10.1088/1741-4326/ad6ce7
M. Peterka, J. Seidl, T. Markovic, A. Loarte, N.C. Logan, J.-K. Park, P. Cahyna, J. Havlicek, M. Imrisek, L. Kripner, R. Panek, M. Sos, P. Bilkova, K. Bogar, P. Bohm, A. Casolari, Y. Gribov, O. Grover, P. Hacek, M. Hron, K. Kovarik, M. Tomes, D. Tskhakaya, J. Varju, P. Vondracek, V. Weinzettl, the COMPASS Teama
This work presents the first analysis of the disruptive locked mode (LM) triggered by the dynamics of a confinement change. It shows that, under certain conditions, the LM threshold during the transient is significantly lower than expected from steady states. We investigate the sensitivity to a controlled n = 1 error field (EF) activated prior to the L-H transition in the COMPASS tokamak, at q95 ∼ 3, βN ∼ 1, and using EF coils on the high-field side of the vessel. A threshold for EF penetration subsequent to the L-H transition is identified, which shows no significant trend with density or applied torque, and is an apparent consequence of the reduced intrinsic rotation of the 2/1 mode during this transient phase. This finding challenges the assumption made in theoretical and empirical works that natural mode rotation can be predicted by global plasma parameters and urges against using any parametric EF penetration scaling derived from steady-state experiments to define the EF correction strategy in the entire discharge. Furthermore, even at EFs below the identified penetration threshold, disruptive locking of sawtooth-seeded 2/1 tearing modes is observed after about 30% of L-H transitions without external torque.
这项研究首次分析了由约束变化动态引发的破坏性锁定模式（LM）。它表明，在某些条件下，瞬态期间的锁定模式阈值明显低于稳定状态下的预期阈值。我们研究了在 COMPASS 托卡马克的 L-H 转换之前，在 q95 ∼ 3，βN ∼ 1 的条件下，使用容器高场侧的 EF 线圈激活的受控 n = 1 误差场（EF）的敏感性。确定了 L-H 转换后 EF 穿透的阈值，该阈值与密度或应用扭矩无明显趋势，是 2/1 模式在此瞬态阶段本征旋转减少的明显结果。这一发现挑战了理论和经验研究中的假设，即自然模式旋转可由全局等离子体参数预测，并敦促不要使用从稳态实验中得出的任何参数 EF 穿透比例来定义整个放电过程中的 EF 校正策略。此外，即使在低于所确定的穿透阈值的 EF 下，在没有外部扭矩的情况下，在约 30% 的 L-H 转变之后，也能观察到锯齿种子 2/1 撕裂模式的破坏性锁定。
{"title":"Quantification of locked mode instability triggered by a change in confinement","authors":"M. Peterka, J. Seidl, T. Markovic, A. Loarte, N.C. Logan, J.-K. Park, P. Cahyna, J. Havlicek, M. Imrisek, L. Kripner, R. Panek, M. Sos, P. Bilkova, K. Bogar, P. Bohm, A. Casolari, Y. Gribov, O. Grover, P. Hacek, M. Hron, K. Kovarik, M. Tomes, D. Tskhakaya, J. Varju, P. Vondracek, V. Weinzettl, the COMPASS Teama","doi":"10.1088/1741-4326/ad6ce7","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6ce7","url":null,"abstract":"This work presents the first analysis of the disruptive locked mode (LM) triggered by the dynamics of a confinement change. It shows that, under certain conditions, the LM threshold during the transient is significantly lower than expected from steady states. We investigate the sensitivity to a controlled <italic toggle=\"yes\">n</italic> = 1 error field (EF) activated prior to the L-H transition in the COMPASS tokamak, at <italic toggle=\"yes\">q</italic><sub>95</sub> ∼ 3, <italic toggle=\"yes\">β</italic><sub>N</sub> ∼ 1, and using EF coils on the high-field side of the vessel. A threshold for EF penetration subsequent to the L-H transition is identified, which shows no significant trend with density or applied torque, and is an apparent consequence of the reduced intrinsic rotation of the 2/1 mode during this transient phase. This finding challenges the assumption made in theoretical and empirical works that natural mode rotation can be predicted by global plasma parameters and urges against using any parametric EF penetration scaling derived from steady-state experiments to define the EF correction strategy in the entire discharge. Furthermore, even at EFs below the identified penetration threshold, disruptive locking of sawtooth-seeded 2/1 tearing modes is observed after about 30% of L-H transitions without external torque.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"76 2 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212116","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 : 2024-09-03DOI: 10.1088/1741-4326/ad6ba5
G. Staebler, C. Bourdelle, J. Citrin, R. Waltz
The theory, development, and validation of reduced quasilinear models of gyrokinetic turbulent transport in the closed flux surface core of tokamaks is reviewed. In combination with neoclassical collisional transport, these models are successful in accurately predicting core tokamak plasma temperature, density, rotation, and impurity profiles in a variety of confinement regimes. Refined experimental tests have been performed to validate the predictions of the quasilinear models, probing changes in the dominant gyrokinetic instabilities, as reflected in fluctuation measurements, cross-phases, and transport properties. These tests continue to produce a deeper understanding of the complex mix of instabilities at both electron and ion gyroradius scales.
{"title":"Quasilinear theory and modelling of gyrokinetic turbulent transport in tokamaks","authors":"G. Staebler, C. Bourdelle, J. Citrin, R. Waltz","doi":"10.1088/1741-4326/ad6ba5","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6ba5","url":null,"abstract":"The theory, development, and validation of reduced quasilinear models of gyrokinetic turbulent transport in the closed flux surface core of tokamaks is reviewed. In combination with neoclassical collisional transport, these models are successful in accurately predicting core tokamak plasma temperature, density, rotation, and impurity profiles in a variety of confinement regimes. Refined experimental tests have been performed to validate the predictions of the quasilinear models, probing changes in the dominant gyrokinetic instabilities, as reflected in fluctuation measurements, cross-phases, and transport properties. These tests continue to produce a deeper understanding of the complex mix of instabilities at both electron and ion gyroradius scales.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"24 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212115","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 : 2024-09-03DOI: 10.1088/1741-4326/ad7272
H.R. Strauss, B.E. Chapman, B.C. Lyons
This paper deals with resistive wall tearing mode (RWTM) disruptions. RWTMs are closely related to resistive wall modes. RWTMs are tearing modes whose linear and nonlinear behavior is strongly dependent on the resistive wall outside the plasma. The consequence for ITER, is that the thermal quench timescale could be much longer than previously conjectured. Active feedback stabilization is another possible way to mitigate or prevent RWTM disruptions. Simulations of RWTM disruptions are reviewed for DIII-D and MST. MST has a longer resistive wall time than ITER, and disruptions are not observed experimentally when MST is operated as a standard tokamak. Simulations indicate that the RWTM disruption time scale is longer than the experimental shot time. Edge cooling causes contraction of the current profile, which can destabilize RWTMs. The equilibria studied here have the q = 2 rational surface close to the edge of the plasma, and low current density between the q = 2 surface and the wall. A sequence of low edge current model equilibria has major disruptions only for a resistive, not ideal, wall, and edge q⩽3.4. This is consistent with regimes of tokamak disruptivity, suggesting that tokamak disruptions caused by edge cooling at low edge q could be RWTMs.
{"title":"Resistive wall tearing mode disruptions","authors":"H.R. Strauss, B.E. Chapman, B.C. Lyons","doi":"10.1088/1741-4326/ad7272","DOIUrl":"https://doi.org/10.1088/1741-4326/ad7272","url":null,"abstract":"This paper deals with resistive wall tearing mode (RWTM) disruptions. RWTMs are closely related to resistive wall modes. RWTMs are tearing modes whose linear and nonlinear behavior is strongly dependent on the resistive wall outside the plasma. The consequence for ITER, is that the thermal quench timescale could be much longer than previously conjectured. Active feedback stabilization is another possible way to mitigate or prevent RWTM disruptions. Simulations of RWTM disruptions are reviewed for DIII-D and MST. MST has a longer resistive wall time than ITER, and disruptions are not observed experimentally when MST is operated as a standard tokamak. Simulations indicate that the RWTM disruption time scale is longer than the experimental shot time. Edge cooling causes contraction of the current profile, which can destabilize RWTMs. The equilibria studied here have the <italic toggle=\"yes\">q</italic> = 2 rational surface close to the edge of the plasma, and low current density between the <italic toggle=\"yes\">q</italic> = 2 surface and the wall. A sequence of low edge current model equilibria has major disruptions only for a resistive, not ideal, wall, and edge <inline-formula>\u0000<tex-math><?CDATA $q unicode{x2A7D} 3.4.$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:mi>q</mml:mi><mml:mtext>⩽</mml:mtext><mml:mn>3.4</mml:mn><mml:mo>.</mml:mo></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad7272ieqn1.gif\"></inline-graphic></inline-formula> This is consistent with regimes of tokamak disruptivity, suggesting that tokamak disruptions caused by edge cooling at low edge <italic toggle=\"yes\">q</italic> could be RWTMs.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"7 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212148","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 : 2024-09-03DOI: 10.1088/1741-4326/ad7273
N.J. Richner, L. Bardóczi, J.D. Callen, R.J. La Haye, N.C. Logan, E.J. Strait
Plasma differential rotation is found to be capable of preventing disruptive neoclassical tearing modes (NTMs) seeded by nonlinear three-wave coupling. As tearing modes degrade confinement and can lead to disruptions, stabilization strategies are crucial to the successful operation of future devices. In ITER-relevant scenarios on DIII-D, rotationally coupled m/n = 1/1 and 3/2 modes have been observed to drive 2/1 islands through three-wave coupling. The frequency of the driven 2/1 mode is set by matching conditions and the frequencies of the driving modes. When the driven mode frequency matches the local plasma rotation frequency, e.g. at low differential rotation, the driven 2/1 island can grow into a disruptive NTM. Using neutral beam torque as an actuator to scan the differential rotation, these experiments demonstrate that a sufficiently large frequency mismatch prevents destabilization of disruptive 2/1 NTMs by three-wave coupling. This work indicates that differential rotation can be used as an actuator to prevent NTMs seeded by three-wave coupling.
{"title":"Use of differential plasma rotation to prevent disruptive tearing mode onset from 3-wave coupling","authors":"N.J. Richner, L. Bardóczi, J.D. Callen, R.J. La Haye, N.C. Logan, E.J. Strait","doi":"10.1088/1741-4326/ad7273","DOIUrl":"https://doi.org/10.1088/1741-4326/ad7273","url":null,"abstract":"Plasma differential rotation is found to be capable of preventing disruptive neoclassical tearing modes (NTMs) seeded by nonlinear three-wave coupling. As tearing modes degrade confinement and can lead to disruptions, stabilization strategies are crucial to the successful operation of future devices. In ITER-relevant scenarios on DIII-D, rotationally coupled <inline-formula>\u0000<tex-math><?CDATA $m/n$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:mi>m</mml:mi><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:mi>n</mml:mi></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad7273ieqn1.gif\"></inline-graphic></inline-formula> = 1/1 and 3/2 modes have been observed to drive 2/1 islands through three-wave coupling. The frequency of the driven 2/1 mode is set by matching conditions and the frequencies of the driving modes. When the driven mode frequency matches the local plasma rotation frequency, e.g. at low differential rotation, the driven 2/1 island can grow into a disruptive NTM. Using neutral beam torque as an actuator to scan the differential rotation, these experiments demonstrate that a sufficiently large frequency mismatch prevents destabilization of disruptive 2/1 NTMs by three-wave coupling. This work indicates that differential rotation can be used as an actuator to prevent NTMs seeded by three-wave coupling.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"7 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212120","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 : 2024-09-03DOI: 10.1088/1741-4326/ad7303
H. Frerichs
The FLARE code is a magnetic mesh generator that is integrated within a suite of tools for the analysis of the magnetic geometry in toroidal fusion devices. A magnetic mesh is constructed from field line segments and permits fast reconstruction of field lines in 3D boundary plasma codes such as EMC3-EIRENE. Both intrinsically non-axisymmetric configurations (stellarators) and those with symmetry breaking perturbations of an axisymmetric equilibrium (tokamaks) are supported. The code itself is written in Modern Fortran with MPI support for parallel computing, and it incorporates object-oriented programming for the definition of the magnetic field and the material surface geometry. Extended derived types for a number of different magnetohydrodynamic equilibrium and plasma response models are implemented. The core element of FLARE is a field line tracer with adaptive step-size control, and this is integrated into tools for the construction of Poincaré maps and invariant manifolds of X-points. A collection of high-level procedures that generate output files for visualization is build on top of that. The analysis modules are build with Python frontends that facilitate customization of tasks and/or scripting of parameter scans.
FLARE 代码是一种磁场网格生成器，集成在一套工具中，用于分析环形聚变装置中的磁场几何形状。磁网格由磁场线段构建，允许在三维边界等离子体代码（如 EMC3-EIRENE）中快速重建磁场线。既支持非轴对称结构（恒星器），也支持轴对称平衡的对称破缺扰动结构（托卡马克）。代码本身采用现代Fortran语言编写，支持MPI并行计算，并采用面向对象编程来定义磁场和材料表面几何形状。它还为许多不同的磁流体动力学平衡和等离子体响应模型实现了扩展的派生类型。FLARE 的核心要素是一个具有自适应步长控制的场线追踪器，并将其集成到用于构建 Poincaré 地图和 X 点不变流形的工具中。在此基础上，还构建了一系列高级程序，用于生成可视化输出文件。分析模块采用 Python 前端，便于定制任务和/或编写参数扫描脚本。
{"title":"FLARE: field line analysis and reconstruction for 3D boundary plasma modeling","authors":"H. Frerichs","doi":"10.1088/1741-4326/ad7303","DOIUrl":"https://doi.org/10.1088/1741-4326/ad7303","url":null,"abstract":"The FLARE code is a magnetic mesh generator that is integrated within a suite of tools for the analysis of the magnetic geometry in toroidal fusion devices. A magnetic mesh is constructed from field line segments and permits fast reconstruction of field lines in 3D boundary plasma codes such as EMC3-EIRENE. Both intrinsically non-axisymmetric configurations (stellarators) and those with symmetry breaking perturbations of an axisymmetric equilibrium (tokamaks) are supported. The code itself is written in Modern Fortran with MPI support for parallel computing, and it incorporates object-oriented programming for the definition of the magnetic field and the material surface geometry. Extended derived types for a number of different magnetohydrodynamic equilibrium and plasma response models are implemented. The core element of FLARE is a field line tracer with adaptive step-size control, and this is integrated into tools for the construction of Poincaré maps and invariant manifolds of X-points. A collection of high-level procedures that generate output files for visualization is build on top of that. The analysis modules are build with Python frontends that facilitate customization of tasks and/or scripting of parameter scans.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"64 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212122","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 : 2024-09-03DOI: 10.1088/1741-4326/ad6e07
Hao Yang, Guido Ciraolo, Olivier Février, Nicolas Fedorczak, Nicolas Rivals, Andreas Bierwage, Hugo Bufferand, Gloria L Falchetto, Tomohide Nakano, Patrick Tamain, Jérôme Bucalossi, the WEST teama
The parameter <inline-formula> <tex-math><?CDATA $R_mathrm{D} = P_mathrm{rad}/P_mathrm{cond}$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:msub><mml:mi>R</mml:mi><mml:mrow><mml:mi mathvariant="normal">D</mml:mi></mml:mrow></mml:msub><mml:mo>=</mml:mo><mml:msub><mml:mi>P</mml:mi><mml:mrow><mml:mi>rad</mml:mi></mml:mrow></mml:msub><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:msub><mml:mi>P</mml:mi><mml:mrow><mml:mi>cond</mml:mi></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href="nfad6e07ieqn1.gif"></inline-graphic></inline-formula>, which measures the ratio of radiated power to conductive heat flux at downstream <italic toggle="yes">Scrape-Off-Layer</italic> (SOL), is proposed as a robust and practically useful figure of merit for divertor detachment control. The simulations performed using the SOLEDGE3X-EIRENE code predict that the instant where <inline-formula> <tex-math><?CDATA $R_mathrm{D}$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:msub><mml:mi>R</mml:mi><mml:mrow><mml:mi mathvariant="normal">D</mml:mi></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href="nfad6e07ieqn2.gif"></inline-graphic></inline-formula> passes through unity (that is, when <inline-formula> <tex-math><?CDATA $P_mathrm{rad} approx P_mathrm{cond}$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:msub><mml:mi>P</mml:mi><mml:mrow><mml:mi>rad</mml:mi></mml:mrow></mml:msub><mml:mo>≈</mml:mo><mml:msub><mml:mi>P</mml:mi><mml:mrow><mml:mi>cond</mml:mi></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href="nfad6e07ieqn3.gif"></inline-graphic></inline-formula>) coincides with the detachment of the radiation front from the divertor target. Furthermore, as a function of <inline-formula> <tex-math><?CDATA $R_mathrm{D}$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:msub><mml:mi>R</mml:mi><mml:mrow><mml:mi mathvariant="normal">D</mml:mi></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href="nfad6e07ieqn4.gif"></inline-graphic></inline-formula>, there is a decrease in target temperature and an increase in the distance at which the radiation front detaches from the target. These simulations cover scenarios in WEST and TCV with different levels of confinement, divertor closure, impurity concentration, and input power. The physical rationale underlying the above definition of <inline-formula> <tex-math><?CDATA $R_mathrm{D}$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:msub><mml:mi>R</mml:mi><mml:mrow><mml:mi mathvariant="normal">D</mml:mi></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href="nfad6e07ieqn5.gif"></inline-graphic></inline-formula> is that when the divertor radiated power is comparable to the conductive heat flux, there will be a lack of energy reaching the target. Consequently, the radiation front detaches some distance from the divertor target. <inline-formula> <tex-math><?CDATA $R_mathrm{D}$?></tex-math><mml:math overflow="scroll"><mml:mrow><mm
{"title":"Numerical study of a general criterion for divertor detachment control","authors":"Hao Yang, Guido Ciraolo, Olivier Février, Nicolas Fedorczak, Nicolas Rivals, Andreas Bierwage, Hugo Bufferand, Gloria L Falchetto, Tomohide Nakano, Patrick Tamain, Jérôme Bucalossi, the WEST teama","doi":"10.1088/1741-4326/ad6e07","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6e07","url":null,"abstract":"The parameter <inline-formula>\u0000<tex-math><?CDATA $R_mathrm{D} = P_mathrm{rad}/P_mathrm{cond}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:msub><mml:mi>R</mml:mi><mml:mrow><mml:mi mathvariant=\"normal\">D</mml:mi></mml:mrow></mml:msub><mml:mo>=</mml:mo><mml:msub><mml:mi>P</mml:mi><mml:mrow><mml:mi>rad</mml:mi></mml:mrow></mml:msub><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:msub><mml:mi>P</mml:mi><mml:mrow><mml:mi>cond</mml:mi></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad6e07ieqn1.gif\"></inline-graphic></inline-formula>, which measures the ratio of radiated power to conductive heat flux at downstream <italic toggle=\"yes\">Scrape-Off-Layer</italic> (SOL), is proposed as a robust and practically useful figure of merit for divertor detachment control. The simulations performed using the SOLEDGE3X-EIRENE code predict that the instant where <inline-formula>\u0000<tex-math><?CDATA $R_mathrm{D}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:msub><mml:mi>R</mml:mi><mml:mrow><mml:mi mathvariant=\"normal\">D</mml:mi></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad6e07ieqn2.gif\"></inline-graphic></inline-formula> passes through unity (that is, when <inline-formula>\u0000<tex-math><?CDATA $P_mathrm{rad} approx P_mathrm{cond}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:msub><mml:mi>P</mml:mi><mml:mrow><mml:mi>rad</mml:mi></mml:mrow></mml:msub><mml:mo>≈</mml:mo><mml:msub><mml:mi>P</mml:mi><mml:mrow><mml:mi>cond</mml:mi></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad6e07ieqn3.gif\"></inline-graphic></inline-formula>) coincides with the detachment of the radiation front from the divertor target. Furthermore, as a function of <inline-formula>\u0000<tex-math><?CDATA $R_mathrm{D}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:msub><mml:mi>R</mml:mi><mml:mrow><mml:mi mathvariant=\"normal\">D</mml:mi></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad6e07ieqn4.gif\"></inline-graphic></inline-formula>, there is a decrease in target temperature and an increase in the distance at which the radiation front detaches from the target. These simulations cover scenarios in WEST and TCV with different levels of confinement, divertor closure, impurity concentration, and input power. The physical rationale underlying the above definition of <inline-formula>\u0000<tex-math><?CDATA $R_mathrm{D}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:msub><mml:mi>R</mml:mi><mml:mrow><mml:mi mathvariant=\"normal\">D</mml:mi></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad6e07ieqn5.gif\"></inline-graphic></inline-formula> is that when the divertor radiated power is comparable to the conductive heat flux, there will be a lack of energy reaching the target. Consequently, the radiation front detaches some distance from the divertor target. <inline-formula>\u0000<tex-math><?CDATA $R_mathrm{D}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mm","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"63 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212118","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 : 2024-09-03DOI: 10.1088/1741-4326/ad7275
G.J. Kramer, A. Bortolon, A. Diallo, R. Maingi
Full orbit-following simulations of thermal ions show that finite ion-orbit width effects create charge separation near the last closed flux surface (LCFS) which generates a localized radial electric field. Experimentally, edge electric fields are observed in H-mode plasmas and they are necessary for the edge turbulence suppression via the E×B flow shear mechanism. Confined trapped (and to a lesser extent co-passing) ions near the plasma edge form a positive charge distribution outside the LCFS, while thermal electrons are tied more tightly to field lines owing to their small mass and are poorly confined outside the LCFS, hence charge neutrality is violated outside the LCFS. A large number of reported observations from spherical and conventional tokamaks support the results from the simulations although the simulations were not performed fully self consistently. The results suggest ways to lower the H-mode power threshold and optimize the H-mode plasma edge.
热离子的全轨道跟踪模拟表明，有限离子轨道宽度效应会在最后一个封闭通量面（LCFS）附近产生电荷分离，从而产生局部径向电场。实验在 H 模式等离子体中观测到了边缘电场，它们是通过 E×B 流剪切机制抑制边缘湍流的必要条件。等离子体边缘附近受限的被困离子（其次是共通离子）在 LCFS 外形成正电荷分布，而热电子由于质量较小，与场线的束缚较紧，在 LCFS 外受限程度较低，因此在 LCFS 外违反了电荷中性原则。尽管模拟并不是完全自洽地进行的，但大量来自球形和常规托卡马克的观测报告支持了模拟结果。结果提出了降低 H 模式功率阈值和优化 H 模式等离子体边缘的方法。
{"title":"The formation of an radial edge electric field due to finite ion orbit width effects is the possible root cause of the H-mode edge","authors":"G.J. Kramer, A. Bortolon, A. Diallo, R. Maingi","doi":"10.1088/1741-4326/ad7275","DOIUrl":"https://doi.org/10.1088/1741-4326/ad7275","url":null,"abstract":"Full orbit-following simulations of thermal ions show that finite ion-orbit width effects create charge separation near the last closed flux surface (LCFS) which generates a localized radial electric field. Experimentally, edge electric fields are observed in H-mode plasmas and they are necessary for the edge turbulence suppression via the <inline-formula>\u0000<tex-math><?CDATA $boldsymbol{ E times B }$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:mi mathvariant=\"bold-italic\">E</mml:mi><mml:mo mathvariant=\"bold\">×</mml:mo><mml:mi mathvariant=\"bold-italic\">B</mml:mi></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad7275ieqn1.gif\"></inline-graphic></inline-formula> flow shear mechanism. Confined trapped (and to a lesser extent co-passing) ions near the plasma edge form a positive charge distribution outside the LCFS, while thermal electrons are tied more tightly to field lines owing to their small mass and are poorly confined outside the LCFS, hence charge neutrality is violated outside the LCFS. A large number of reported observations from spherical and conventional tokamaks support the results from the simulations although the simulations were not performed fully self consistently. The results suggest ways to lower the H-mode power threshold and optimize the H-mode plasma edge.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"21 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212121","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 : 2024-09-03DOI: 10.1088/1741-4326/ad6e06
F. Crisanti, R. Ambrosino, M.V. Falessi, L. Gabellieri, G. Giruzzi, G. Granucci, P. Innocente, P. Mantica, G. Ramogida, G. Vlad, R. Albanese, E. Alessi, C. Angioni, P. Agostinetti, L. Aucone, F. Auriemma, B. Baiocchi, L. Balbinot, A. Balestri, T. Barberis, M. Baruzzo, T. Bolzonella, N. Bonanomi, D. Bonfiglio, S. Brezinsek, G. Calabrò, F. Cani, I. Casiraghi, A. Castaldo, C. Castaldo, M. Cavedon, S. Ceccuzzi, F. Cichocki, M. Ciotti, C. Day, C. De Piccoli, G. Dose, E. Emanueli, L. Frassinetti, L. Figini, V. Fusco, E. Giovannozzi, M. Gobbin, F. Koechi, A. Kryzhanovskyy, Y. Li, R. Lombroni, T. Luda, A. Mariani, P. Martin, C. Meineri, A. Murari, P. Muscente, F. Napoli, E. Nardon, R. Neu, M. Nocente, M. Notazio, S. Nowak, L. Pigatto, C. Piron, F. Porcelli, S. Roccella, G. Rubino, M. Scarpari, C. Sozzi, G. Spizzo, F. Subba, F. Taccogna, C. Tantos, D. Terranova, E. Tsitrone, A. Uccello, D. Van Eester, N. Vianello, P. Vincenzi, M. Wischmeier, F. Zonca
This paper is dealing with the physics basis used for the design of the Divertor Tokamak Test facility (DTT), under construction in Frascati (DTT 2019 DTT interim design report (2019)) Italy, and with the description of the main target plasma scenarios of the device. The main goal of the facility will be the study of the power exhaust, intended as a fully integrated core-edge problem, and eventually to propose an optimized divertor for the European DEMO plant. The approach used to design the facility is described and their main features are reported, by using simulations performed by state-of-the-art codes both for the bulk and edge studies. A detailed analysis of MHD, including also the possibility to study disruption events and Energetic Particles physics is also reported. Eventually, a description of the ongoing work to build-up a Research Plan written and shared by the full EUROfusion community is presented.
{"title":"Physics basis for the divertor tokamak test facility","authors":"F. Crisanti, R. Ambrosino, M.V. Falessi, L. Gabellieri, G. Giruzzi, G. Granucci, P. Innocente, P. Mantica, G. Ramogida, G. Vlad, R. Albanese, E. Alessi, C. Angioni, P. Agostinetti, L. Aucone, F. Auriemma, B. Baiocchi, L. Balbinot, A. Balestri, T. Barberis, M. Baruzzo, T. Bolzonella, N. Bonanomi, D. Bonfiglio, S. Brezinsek, G. Calabrò, F. Cani, I. Casiraghi, A. Castaldo, C. Castaldo, M. Cavedon, S. Ceccuzzi, F. Cichocki, M. Ciotti, C. Day, C. De Piccoli, G. Dose, E. Emanueli, L. Frassinetti, L. Figini, V. Fusco, E. Giovannozzi, M. Gobbin, F. Koechi, A. Kryzhanovskyy, Y. Li, R. Lombroni, T. Luda, A. Mariani, P. Martin, C. Meineri, A. Murari, P. Muscente, F. Napoli, E. Nardon, R. Neu, M. Nocente, M. Notazio, S. Nowak, L. Pigatto, C. Piron, F. Porcelli, S. Roccella, G. Rubino, M. Scarpari, C. Sozzi, G. Spizzo, F. Subba, F. Taccogna, C. Tantos, D. Terranova, E. Tsitrone, A. Uccello, D. Van Eester, N. Vianello, P. Vincenzi, M. Wischmeier, F. Zonca","doi":"10.1088/1741-4326/ad6e06","DOIUrl":"https://doi.org/10.1088/1741-4326/ad6e06","url":null,"abstract":"This paper is dealing with the physics basis used for the design of the Divertor Tokamak Test facility (DTT), under construction in Frascati (DTT 2019 DTT interim design report (2019)) Italy, and with the description of the main target plasma scenarios of the device. The main goal of the facility will be the study of the power exhaust, intended as a fully integrated core-edge problem, and eventually to propose an optimized divertor for the European DEMO plant. The approach used to design the facility is described and their main features are reported, by using simulations performed by state-of-the-art codes both for the bulk and edge studies. A detailed analysis of MHD, including also the possibility to study disruption events and Energetic Particles physics is also reported. Eventually, a description of the ongoing work to build-up a Research Plan written and shared by the full EUROfusion community is presented.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"91 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212117","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 : 2024-09-03DOI: 10.1088/1741-4326/ad703e
N. Chaudhary, M. Hirsch, T. Andreeva, J. Geiger, R.C. Wolf, G.A. Wurden, the W7-X Teama
The low magnetic shear in the Wendelstein 7-X (W7-X) stellarator makes it feasible to shape the separatrix by the large islands constituting an island-divertor, and this can be exploited to access various magnetic configurations, including samples of different internal island sizes and locations. To investigate the configuration effects on the plasma confinement, a configuration scan was performed by changing the coil currents to vary the rotational transform between values <inline-formula> <tex-math><?CDATA $5/4$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:mn>5</mml:mn><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:mn>4</mml:mn></mml:mrow></mml:math><inline-graphic xlink:href="nfad703eieqn1.gif"></inline-graphic></inline-formula> and <inline-formula> <tex-math><?CDATA $5/6$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:mn>5</mml:mn><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:mn>6</mml:mn></mml:mrow></mml:math><inline-graphic xlink:href="nfad703eieqn2.gif"></inline-graphic></inline-formula> at the plasma boundary with different power levels (2, 4, 6 MW) of electron cyclotron resonance heating (ECRH) at a maximum plasma density of <inline-formula> <tex-math><?CDATA $8 times 10^{19},textrm{m}^{-2}$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:mn>8</mml:mn><mml:mo>×</mml:mo><mml:msup><mml:mn>10</mml:mn><mml:mrow><mml:mn>19</mml:mn></mml:mrow></mml:msup><mml:mstyle scriptlevel="0"></mml:mstyle><mml:msup><mml:mtext>m</mml:mtext><mml:mrow><mml:mo>−</mml:mo><mml:mn>2</mml:mn></mml:mrow></mml:msup></mml:mrow></mml:math><inline-graphic xlink:href="nfad703eieqn3.gif"></inline-graphic></inline-formula>. neutral beam injection (NBI) heating was also applied during some configurations of the scan to create a density ramp and access high densities beyond the X2 ECRH cutoff. For the magnetic configurations, where the <inline-formula> <tex-math><?CDATA $5/5$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:mn>5</mml:mn><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:mn>5</mml:mn></mml:mrow></mml:math><inline-graphic xlink:href="nfad703eieqn4.gif"></inline-graphic></inline-formula> and <inline-formula> <tex-math><?CDATA $5/6$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:mn>5</mml:mn><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:mn>6</mml:mn></mml:mrow></mml:math><inline-graphic xlink:href="nfad703eieqn5.gif"></inline-graphic></inline-formula> island chains were moved closer to separatrix but remaining inside the last closed flux surface, the electron cyclotron emission shows that an electron temperature, <inline-formula> <tex-math><?CDATA $T_{mathrm e}$?></tex-math><mml:math overflow="scroll"><mml:mrow><mml:msub><mml:mi>T</mml:mi><mml:mrow><mml:mrow><mml:mi mathvariant="normal">e</mml:mi></mml:mrow></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href="nfad703eieqn6.gif"></inline-graphic></inline-formula>, pedestal develops already during ECRH heated plasma buildup phase indicating a transport barrier, and the
{"title":"Electron transport barrier and high confinement in configurations with internal islands close to the plasma edge of W7-X","authors":"N. Chaudhary, M. Hirsch, T. Andreeva, J. Geiger, R.C. Wolf, G.A. Wurden, the W7-X Teama","doi":"10.1088/1741-4326/ad703e","DOIUrl":"https://doi.org/10.1088/1741-4326/ad703e","url":null,"abstract":"The low magnetic shear in the Wendelstein 7-X (W7-X) stellarator makes it feasible to shape the separatrix by the large islands constituting an island-divertor, and this can be exploited to access various magnetic configurations, including samples of different internal island sizes and locations. To investigate the configuration effects on the plasma confinement, a configuration scan was performed by changing the coil currents to vary the rotational transform between values <inline-formula>\u0000<tex-math><?CDATA $5/4$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:mn>5</mml:mn><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:mn>4</mml:mn></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad703eieqn1.gif\"></inline-graphic></inline-formula> and <inline-formula>\u0000<tex-math><?CDATA $5/6$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:mn>5</mml:mn><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:mn>6</mml:mn></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad703eieqn2.gif\"></inline-graphic></inline-formula> at the plasma boundary with different power levels (2, 4, 6 MW) of electron cyclotron resonance heating (ECRH) at a maximum plasma density of <inline-formula>\u0000<tex-math><?CDATA $8 times 10^{19},textrm{m}^{-2}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:mn>8</mml:mn><mml:mo>×</mml:mo><mml:msup><mml:mn>10</mml:mn><mml:mrow><mml:mn>19</mml:mn></mml:mrow></mml:msup><mml:mstyle scriptlevel=\"0\"></mml:mstyle><mml:msup><mml:mtext>m</mml:mtext><mml:mrow><mml:mo>−</mml:mo><mml:mn>2</mml:mn></mml:mrow></mml:msup></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad703eieqn3.gif\"></inline-graphic></inline-formula>. neutral beam injection (NBI) heating was also applied during some configurations of the scan to create a density ramp and access high densities beyond the X2 ECRH cutoff. For the magnetic configurations, where the <inline-formula>\u0000<tex-math><?CDATA $5/5$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:mn>5</mml:mn><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:mn>5</mml:mn></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad703eieqn4.gif\"></inline-graphic></inline-formula> and <inline-formula>\u0000<tex-math><?CDATA $5/6$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:mn>5</mml:mn><mml:mrow><mml:mo>/</mml:mo></mml:mrow><mml:mn>6</mml:mn></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad703eieqn5.gif\"></inline-graphic></inline-formula> island chains were moved closer to separatrix but remaining inside the last closed flux surface, the electron cyclotron emission shows that an electron temperature, <inline-formula>\u0000<tex-math><?CDATA $T_{mathrm e}$?></tex-math><mml:math overflow=\"scroll\"><mml:mrow><mml:msub><mml:mi>T</mml:mi><mml:mrow><mml:mrow><mml:mi mathvariant=\"normal\">e</mml:mi></mml:mrow></mml:mrow></mml:msub></mml:mrow></mml:math><inline-graphic xlink:href=\"nfad703eieqn6.gif\"></inline-graphic></inline-formula>, pedestal develops already during ECRH heated plasma buildup phase indicating a transport barrier, and the","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"31 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-09-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142212119","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":1,"RegionCategory":"物理与天体物理","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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