Pub Date : 2024-06-26DOI: 10.1088/1741-4326/ad58f5
T. Sizyuk, J.N. Brooks, T. Abrams and A. Hassanein
The performance of silicon carbide as an alternative plasma facing material (PFM) was studied at various irradiation conditions relevant to ion energies and fluxes of a fusion reactor. This analysis involves detailed modeling of subsurface plasma/material interactions, sputtered particle transport above the surface and redeposition, and related changes in material composition and microstructure induced by steady-state and Edge Localized Mode ion fluxes. Transition of a crystalline SiC surface to semi-crystalline and amorphous phases was analyzed based on advanced modeling of DIII-D tokamak experiments where SiC was irradiated in single- and multiple- L-mode and H-mode discharges. This analysis shows that displacement damage, particle deposition/redeposition, and D accumulation on the SiC divertor surface can lead to significant microstructural changes that result in enhanced sputtering erosion in comparison with the original crystalline material. However, the resulting total net erosion rate for a full-coverage, advanced tokamak, SiC coated divertor may well be acceptably low. Moreover, the C sputtering yield from the evolved SiC surface can be seven times lower than from a pure graphite surface; this would imply significantly reduced tritium co-deposition rates in a D-T tokamak reactor, compared with a pure carbon surface. It was also determined that chemical sputtering of both C and Si should not result in any noticeable effect on the net erosion, for attached plasma regimes. Our results thus show encouraging results overall for use of SiC as a PFM in tokamaks.
在与聚变反应堆离子能量和通量相关的各种辐照条件下，研究了碳化硅作为替代等离子体表面材料（PFM）的性能。该分析包括对表面下等离子体/材料相互作用、表面上方溅射粒子传输和再沉积以及稳态和边缘局部模式离子通量诱导的材料成分和微观结构的相关变化进行详细建模。基于 DIII-D 托卡马克实验的高级建模，分析了结晶碳化硅表面向半结晶和无定形相的转变，在这些实验中，碳化硅受到单次和多次 L 模式和 H 模式放电的辐照。该分析表明，SiC 分流器表面的位移损伤、粒子沉积/再沉积和 D 积累会导致显著的微观结构变化，从而使溅射侵蚀与原始晶体材料相比有所增强。不过，对于全覆盖的先进托卡马克 SiC 涂层岔流器来说，由此产生的总净侵蚀率可能很低，可以接受。此外，与纯碳表面相比，SiC 表面演化产生的碳溅射量可能比纯石墨表面低七倍；这意味着 D-T 托卡马克反应堆中的氚共沉积率将大大降低。另外还确定，在附加等离子体状态下，C 和 Si 的化学溅射不会对净侵蚀产生任何明显影响。因此，我们的研究结果表明，在托卡马克中使用碳化硅作为 PFM 的总体效果令人鼓舞。
{"title":"Comprehensive new insights on the potential use of SiC as plasma-facing materials in future fusion reactors","authors":"T. Sizyuk, J.N. Brooks, T. Abrams and A. Hassanein","doi":"10.1088/1741-4326/ad58f5","DOIUrl":"https://doi.org/10.1088/1741-4326/ad58f5","url":null,"abstract":"The performance of silicon carbide as an alternative plasma facing material (PFM) was studied at various irradiation conditions relevant to ion energies and fluxes of a fusion reactor. This analysis involves detailed modeling of subsurface plasma/material interactions, sputtered particle transport above the surface and redeposition, and related changes in material composition and microstructure induced by steady-state and Edge Localized Mode ion fluxes. Transition of a crystalline SiC surface to semi-crystalline and amorphous phases was analyzed based on advanced modeling of DIII-D tokamak experiments where SiC was irradiated in single- and multiple- L-mode and H-mode discharges. This analysis shows that displacement damage, particle deposition/redeposition, and D accumulation on the SiC divertor surface can lead to significant microstructural changes that result in enhanced sputtering erosion in comparison with the original crystalline material. However, the resulting total net erosion rate for a full-coverage, advanced tokamak, SiC coated divertor may well be acceptably low. Moreover, the C sputtering yield from the evolved SiC surface can be seven times lower than from a pure graphite surface; this would imply significantly reduced tritium co-deposition rates in a D-T tokamak reactor, compared with a pure carbon surface. It was also determined that chemical sputtering of both C and Si should not result in any noticeable effect on the net erosion, for attached plasma regimes. Our results thus show encouraging results overall for use of SiC as a PFM in tokamaks.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"48 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141530675","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 : 2024-06-26DOI: 10.1088/1741-4326/ad5a20
Tajinder Singh, Kajal Shah, Deepti Sharma, Joydeep Ghosh, Kumarpalsinh A. Jadeja, Rakesh L. Tanna, M.B. Chowdhuri, Zhihong Lin, Abhijit Sen, Sarveshwar Sharma and Animesh Kuley
The effect of impurity on the electrostatic microturbulence in ADITYA-U tokamak is assessed using global gyrokinetic simulations. The realistic geometry and experimental profiles of the ADITYA-U are used, before and after argon gas seeding, to perform the simulations. Before the impurity seeding, the simulations show the existence of the trapped electron mode (TEM) instability in three distinct regions on the radial-poloidal plane. The mode is identified by its linear eigenmode structure and its characteristic propagation in the electron diamagnetic direction. The simulations with Ar1+ impurity ions in the outer-core region show a significant reduction in the turbulence and transport due to a reduction in the linear instability drive, with respect to the case without impurity. A decrease in particle and heat transport in the outer-core region modifies the plasma density profile measured after the impurity seeding. It, thus, results in the stabilization of the TEM instability in the core region. Due to the reduced turbulence activity, the electron and ion temperatures in the central region increase by about 10%.
利用全局陀螺动力学模拟评估了杂质对ADITYA-U托卡马克中静电微扰的影响。模拟采用了 ADITYA-U 在氩气注入前后的真实几何形状和实验剖面。在添加杂质之前，模拟显示在径向-波状平面上的三个不同区域存在困电子模式（TEM）不稳定性。该模式可通过其线性特征模式结构及其在电子二磁方向上的传播特征识别出来。在外核区域加入 Ar1+ 杂质离子的模拟结果表明，由于线性不稳定性驱动力的减弱，湍流和传输较无杂质的情况显著减弱。外核区粒子和热传输的减少改变了杂质添加后测量到的等离子体密度曲线。因此，这导致了核心区域 TEM 不稳定性的稳定。由于湍流活动减少，中心区域的电子和离子温度上升了约 10%。
{"title":"Gyrokinetic simulations of electrostatic microturbulence in ADITYA-U tokamak with argon impurity","authors":"Tajinder Singh, Kajal Shah, Deepti Sharma, Joydeep Ghosh, Kumarpalsinh A. Jadeja, Rakesh L. Tanna, M.B. Chowdhuri, Zhihong Lin, Abhijit Sen, Sarveshwar Sharma and Animesh Kuley","doi":"10.1088/1741-4326/ad5a20","DOIUrl":"https://doi.org/10.1088/1741-4326/ad5a20","url":null,"abstract":"The effect of impurity on the electrostatic microturbulence in ADITYA-U tokamak is assessed using global gyrokinetic simulations. The realistic geometry and experimental profiles of the ADITYA-U are used, before and after argon gas seeding, to perform the simulations. Before the impurity seeding, the simulations show the existence of the trapped electron mode (TEM) instability in three distinct regions on the radial-poloidal plane. The mode is identified by its linear eigenmode structure and its characteristic propagation in the electron diamagnetic direction. The simulations with Ar1+ impurity ions in the outer-core region show a significant reduction in the turbulence and transport due to a reduction in the linear instability drive, with respect to the case without impurity. A decrease in particle and heat transport in the outer-core region modifies the plasma density profile measured after the impurity seeding. It, thus, results in the stabilization of the TEM instability in the core region. Due to the reduced turbulence activity, the electron and ion temperatures in the central region increase by about 10%.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"108 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-06-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141530676","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 : 2024-06-25DOI: 10.1088/1741-4326/ad4d02
J.F. Parisi, A.O. Nelson, W. Guttenfelder, R. Gaur, J.W. Berkery, S.M. Kaye, K. Barada, C. Clauser, A. Diallo, D.R. Hatch, A. Kleiner, M. Lampert, T. Macwan and J.E. Menard
A gyrokinetic threshold model for pedestal width–height scaling prediction is applied to multiple devices. A shaping and aspect ratio scan is performed on National Spherical Torus Experiment (NSTX) equilibria, finding for the wide-pedestal branch with pedestal width , aspect ratio A, elongation κ, triangularity δ, and normalized pedestal height . The width–transport scaling is found to vary significantly if the pedestal height is varied either with a fixed density or fixed temperature, showing how fueling and heating sources affect the pedestal density and temperature profiles for the kinetic-ballooning-mode (KBM) limited profiles. For an NSTX equilibrium, at fixed density, the wide branch is and at fixed temperature , where and are turbulent electron heat and particle fluxes and for an electron temperature and density . Pedestals close to the KBM limit are shown to have modified turbulent transport coefficients compared to the strongly driven KBMs. The role of flow shear is studied as a width–height scaling constraint and pedestal saturation mechanism for a standard and lithiated wide pedestal discharge. Finally, the stability, transport, and flow shear constraints are combined and examined for an NSTX experiment.
{"title":"Stability and transport of gyrokinetic critical pedestals","authors":"J.F. Parisi, A.O. Nelson, W. Guttenfelder, R. Gaur, J.W. Berkery, S.M. Kaye, K. Barada, C. Clauser, A. Diallo, D.R. Hatch, A. Kleiner, M. Lampert, T. Macwan and J.E. Menard","doi":"10.1088/1741-4326/ad4d02","DOIUrl":"https://doi.org/10.1088/1741-4326/ad4d02","url":null,"abstract":"A gyrokinetic threshold model for pedestal width–height scaling prediction is applied to multiple devices. A shaping and aspect ratio scan is performed on National Spherical Torus Experiment (NSTX) equilibria, finding for the wide-pedestal branch with pedestal width , aspect ratio A, elongation κ, triangularity δ, and normalized pedestal height . The width–transport scaling is found to vary significantly if the pedestal height is varied either with a fixed density or fixed temperature, showing how fueling and heating sources affect the pedestal density and temperature profiles for the kinetic-ballooning-mode (KBM) limited profiles. For an NSTX equilibrium, at fixed density, the wide branch is and at fixed temperature , where and are turbulent electron heat and particle fluxes and for an electron temperature and density . Pedestals close to the KBM limit are shown to have modified turbulent transport coefficients compared to the strongly driven KBMs. The role of flow shear is studied as a width–height scaling constraint and pedestal saturation mechanism for a standard and lithiated wide pedestal discharge. Finally, the stability, transport, and flow shear constraints are combined and examined for an NSTX experiment.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"20 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-06-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141530678","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 : 2024-06-25DOI: 10.1088/1741-4326/ad56a1
R. Ramasamy, K. Aleynikova, N. Nikulsin, F. Hindenlang, I. Holod, E. Strumberger, M. Hoelzl and the JOREK team
An important question for the outlook of stellarator reactors is their robustness against pressure driven modes, and the underlying mechanism behind experimentally observed soft β limits. Towards building a robust answer to these questions, simulation studies are presented using a recently derived reduced nonlinear MHD model. First, the initial model implementation is extended to capture fluid compression by including the influence of parallel flows. Linear benchmarks of a (2, 1) tearing mode in W7-AS geometry, and interchange modes in a finite β, net-zero current carrying stellarator with low magnetic shear are then used to demonstrate the modeling capabilities. Finally, a validation study is conducted on experimental reconstructions of finite β W7-AS discharges. In agreement with past experimental analysis, it is shown that (i) the MHD activity is resistive, (ii) a soft β limit is observed, when the plasma resistivity approaches the estimated experimental value, and (iii) low n MHD activity is observed at intermediate β values, particularly a nonlinearly dominant (2, 1) mode. The MHD activity is mild, explaining the soft β limit, because the plasma volume remains separated into distinct sub-volumes in which field lines are ergodically confined. For the assumed transport parameters, the enhanced perpendicular transport along stochastic magnetic field lines can be overcome with the experimental heating power. The limitations in the current modeling are described, alongside an outlook for characterizing the quasi-steady state operational limit in W7-AS and other devices in more detail in future work.
{"title":"Nonlinear MHD modeling of soft β limits in W7-AS","authors":"R. Ramasamy, K. Aleynikova, N. Nikulsin, F. Hindenlang, I. Holod, E. Strumberger, M. Hoelzl and the JOREK team","doi":"10.1088/1741-4326/ad56a1","DOIUrl":"https://doi.org/10.1088/1741-4326/ad56a1","url":null,"abstract":"An important question for the outlook of stellarator reactors is their robustness against pressure driven modes, and the underlying mechanism behind experimentally observed soft β limits. Towards building a robust answer to these questions, simulation studies are presented using a recently derived reduced nonlinear MHD model. First, the initial model implementation is extended to capture fluid compression by including the influence of parallel flows. Linear benchmarks of a (2, 1) tearing mode in W7-AS geometry, and interchange modes in a finite β, net-zero current carrying stellarator with low magnetic shear are then used to demonstrate the modeling capabilities. Finally, a validation study is conducted on experimental reconstructions of finite β W7-AS discharges. In agreement with past experimental analysis, it is shown that (i) the MHD activity is resistive, (ii) a soft β limit is observed, when the plasma resistivity approaches the estimated experimental value, and (iii) low n MHD activity is observed at intermediate β values, particularly a nonlinearly dominant (2, 1) mode. The MHD activity is mild, explaining the soft β limit, because the plasma volume remains separated into distinct sub-volumes in which field lines are ergodically confined. For the assumed transport parameters, the enhanced perpendicular transport along stochastic magnetic field lines can be overcome with the experimental heating power. The limitations in the current modeling are described, alongside an outlook for characterizing the quasi-steady state operational limit in W7-AS and other devices in more detail in future work.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"36 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-06-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141530677","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 : 2024-06-24DOI: 10.1088/1741-4326/ad52a5
M. Zlobinski, G. Sergienko, I. Jepu, C. Rowley, A. Widdowson, R. Ellis, D. Kos, I. Coffey, M. Fortune, D. Kinna, M. Beldishevski, A. Krimmer, H.T. Lambertz, A. Terra, A. Huber, S. Brezinsek, T. Dittmar, M. Flebbe, R. Yi, R. Rayaprolu, J. Figueiredo, P. Blatchford, S. Silburn, E. Tsitrone, E. Joffrin, K. Krieger, Y. Corre, A. Hakola, J. Likonen, the Eurofusion Tokamak Exploitation Team and JET Contributors
The paper reports the first demonstration of in situ laser-induced desorption — quadrupole mass spectrometry (LID-QMS) application on a large scale fusion device performed in summer 2023. LID-QMS allows direct measurements of the fuel inventory of plasma facing components without retrieving them from the fusion device. The diagnostic desorbs the retained gases by heating a 3 mm diameter spot on the wall using a 1 ms long laser pulse and detects them by QMS. Thus, it can measure the gas content at any wall position accessible to the laser. The successful LID-QMS application in laboratory scale and on medium size fusion devices has now been demonstrated on the larger scale and it is already foreseen as tritium monitor diagnostic in ITER. This in situ diagnostic gives direct access to retention physics on a short timescale instead of campaign-integrated measurements and can assess the space-resolvedefficacy of detritation methods. LID-QMS can be applied on many materials: on Be deposits like in JET, B deposits like in TEXTOR, C based materials or on bulk-W.
本文报告了 2023 年夏季在大型聚变装置上首次演示的原位激光诱导解吸-四极杆质谱（LID-QMS）应用。激光诱导解吸-四极杆质谱仪可以直接测量等离子体中的燃料库存，而无需从聚变装置中取出。诊断仪通过使用 1 毫秒长的激光脉冲加热壁上直径为 3 毫米的光斑来解吸保留的气体，并通过 QMS 对其进行检测。因此，它可以在激光可及的任何墙壁位置测量气体含量。LID-QMS 在实验室和中型核聚变装置上的成功应用现已在大型装置上得到验证，并已预计在国际热核聚变实验堆中用作氚监测诊断仪。这种原位诊断方法可以在短时间内直接获取滞留物理学信息，而不是进行运动集成测量，并可以评估空间分辨脱核方法的有效性。LID-QMS 可应用于多种材料：如 JET 中的 Be 沉积物、TEXTOR 中的 B 沉积物、C 基材料或 bulk-W。
{"title":"First results of laser-induced desorption - quadrupole mass spectrometry (LID-QMS) at JET","authors":"M. Zlobinski, G. Sergienko, I. Jepu, C. Rowley, A. Widdowson, R. Ellis, D. Kos, I. Coffey, M. Fortune, D. Kinna, M. Beldishevski, A. Krimmer, H.T. Lambertz, A. Terra, A. Huber, S. Brezinsek, T. Dittmar, M. Flebbe, R. Yi, R. Rayaprolu, J. Figueiredo, P. Blatchford, S. Silburn, E. Tsitrone, E. Joffrin, K. Krieger, Y. Corre, A. Hakola, J. Likonen, the Eurofusion Tokamak Exploitation Team and JET Contributors","doi":"10.1088/1741-4326/ad52a5","DOIUrl":"https://doi.org/10.1088/1741-4326/ad52a5","url":null,"abstract":"The paper reports the first demonstration of in situ laser-induced desorption — quadrupole mass spectrometry (LID-QMS) application on a large scale fusion device performed in summer 2023. LID-QMS allows direct measurements of the fuel inventory of plasma facing components without retrieving them from the fusion device. The diagnostic desorbs the retained gases by heating a 3 mm diameter spot on the wall using a 1 ms long laser pulse and detects them by QMS. Thus, it can measure the gas content at any wall position accessible to the laser. The successful LID-QMS application in laboratory scale and on medium size fusion devices has now been demonstrated on the larger scale and it is already foreseen as tritium monitor diagnostic in ITER. This in situ diagnostic gives direct access to retention physics on a short timescale instead of campaign-integrated measurements and can assess the space-resolvedefficacy of detritation methods. LID-QMS can be applied on many materials: on Be deposits like in JET, B deposits like in TEXTOR, C based materials or on bulk-W.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"35 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-06-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141530814","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 : 2024-06-24DOI: 10.1088/1741-4326/ad54d7
O. Vallhagen, L. Hanebring, F.J. Artola, M. Lehnen, E. Nardon, T. Fülöp, M. Hoppe, S.L. Newton and I. Pusztai
This study systematically explores the parameter space of disruption mitigation through shattered pellet injection in ITER with a focus on runaway electron (RE) dynamics, using the disruption modeling tool Dream. The physics fidelity is considerably increased compared to previous studies, by e.g. using realistic magnetic geometry, resistive wall configuration, thermal quench onset criteria, as well as including additional effects, such as ion transport and enhanced RE transport during the thermal quench. The work aims to provide a fairly comprehensive coverage of experimentally feasible scenarios, considering plasmas representative of both non-activated and high-performance DT operation, different thermal quench onset criteria and transport levels, a wide range of hydrogen and neon quantities injected in one or two stages, and pellets with various characteristic shard sizes. Using a staggered injection scheme, with a pure hydrogen injection preceding a mixed hydrogen-neon injection, we find injection parameters leading to acceptable RE currents in all investigated discharges without activated runaway sources. Dividing the injection into two stages is found to significantly enhance the assimilation and minimize RE generation due to the hot-tail mechanism. However, while a staggered injection outperforms a single stage injection also in cases with radioactive RE sources, no cases with acceptable RE currents are found for a DT-plasma with a plasma current.
本研究利用破坏建模工具 Dream 系统地探索了通过在热核实验堆中注入碎裂颗粒来减缓破坏的参数空间，重点关注失控电子（RE）动力学。与之前的研究相比，该研究的物理保真度有了显著提高，例如采用了现实的磁几何形状、电阻壁配置、热淬火起始标准，并纳入了额外的效应，例如热淬火期间的离子传输和增强的RE传输。这项工作旨在提供相当全面的实验可行方案，考虑到非活化和高性能 DT 运行的代表性等离子体、不同的热骤冷起始标准和传输水平、在一个或两个阶段注入的大量氢和氖，以及具有各种特征碎片尺寸的颗粒。采用交错注入方案，在氢氖混合注入之前注入纯氢，我们发现注入参数可在所有研究的放电中产生可接受的 RE 电流，且不会激活失控源。我们发现，将注入分为两个阶段可显著提高同化效果，并将热尾机制导致的可再生能源产生降至最低。然而，虽然在有放射性可再生能源的情况下，交错注入比单级注入效果更好，但在有等离子体电流的 DT 等离子体中，没有发现可接受可再生能源电流的情况。
{"title":"Runaway electron dynamics in ITER disruptions with shattered pellet injections","authors":"O. Vallhagen, L. Hanebring, F.J. Artola, M. Lehnen, E. Nardon, T. Fülöp, M. Hoppe, S.L. Newton and I. Pusztai","doi":"10.1088/1741-4326/ad54d7","DOIUrl":"https://doi.org/10.1088/1741-4326/ad54d7","url":null,"abstract":"This study systematically explores the parameter space of disruption mitigation through shattered pellet injection in ITER with a focus on runaway electron (RE) dynamics, using the disruption modeling tool Dream. The physics fidelity is considerably increased compared to previous studies, by e.g. using realistic magnetic geometry, resistive wall configuration, thermal quench onset criteria, as well as including additional effects, such as ion transport and enhanced RE transport during the thermal quench. The work aims to provide a fairly comprehensive coverage of experimentally feasible scenarios, considering plasmas representative of both non-activated and high-performance DT operation, different thermal quench onset criteria and transport levels, a wide range of hydrogen and neon quantities injected in one or two stages, and pellets with various characteristic shard sizes. Using a staggered injection scheme, with a pure hydrogen injection preceding a mixed hydrogen-neon injection, we find injection parameters leading to acceptable RE currents in all investigated discharges without activated runaway sources. Dividing the injection into two stages is found to significantly enhance the assimilation and minimize RE generation due to the hot-tail mechanism. However, while a staggered injection outperforms a single stage injection also in cases with radioactive RE sources, no cases with acceptable RE currents are found for a DT-plasma with a plasma current.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"24 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-06-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141530679","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 : 2024-06-23DOI: 10.1088/1741-4326/ad521a
M.W. Lee, S.-H. Hahn, D. Kim, J. Kang, W.H. Ko, J. Jang, W. Lee and C. Sung
We have observed a stationary high confinement regime with a double transport barrier (DTB), including both internal and edge transport barriers (ITB and ETB) but without edge-localized modes (ELMs), in KSTAR. The ELM-free DTB phase has high thermal confinement comparable to typical H-mode operation in KSTAR. We investigated the characteristics of the DTB phase through various analyses. Transport analysis shows a reduction of ion heat diffusivity to near neoclassical level after the transition from the ELMy H-mode phase to the DTB phase. This result supports the formation of an ion ITB during the DTB phase. Furthermore, we observed that the DTB phase had an edge thermal transport barrier in the ion temperature profile, comparable to that of the H-mode, without a particle transport barrier at the edge. Peeling-ballooning stability analysis indicates that a lower pressure gradient due to density decrease in the DTB phase is mainly responsible for the ELM-free operation. Linear gyrokinetic analysis shows that the real frequency of the most unstable mode in the core region ( = 0.32–0.47) is in the ion diamagnetic direction at both H-mode and DTB phases. At the DTB phase, the linear growth rate inside the ITB is reduced by 50% compared to the ITB foot, while the reduction is not shown at the H-mode phase. Further investigation including nonlinear effects will be needed to better understand the unique operation mode, which can contribute to applying the physical mechanism to fusion reactors in the future.
我们在 KSTAR 中观测到了一种具有双传输势垒 (DTB) 的静态高约束机制，其中包括内部和边缘传输势垒（ITB 和 ETB），但没有边缘定位模式 (ELM)。无 ELM 的 DTB 相具有很高的热约束性，可与 KSTAR 中典型的 H 模式运行相媲美。我们通过各种分析研究了 DTB 阶段的特性。传输分析表明，从 ELMy H 模式阶段过渡到 DTB 阶段后，离子热扩散率降低到接近新古典水平。这一结果支持在 DTB 阶段形成离子 ITB。此外，我们还观察到 DTB 阶段的离子温度曲线中存在边缘热传输障碍，与 H 模式相类似，但边缘没有粒子传输障碍。剥离-气球稳定性分析表明，DTB 相中密度下降导致的较低压力梯度是无 ELM 运行的主要原因。线性陀螺动力学分析表明，在 H 模式和 DTB 阶段，核心区域最不稳定模式的实际频率（= 0.32-0.47）都位于离子二磁方向。在 DTB 阶段，与 ITB 脚相比，ITB 内部的线性增长率降低了 50%，而在 H 模式阶段则没有降低。为了更好地理解这种独特的运行模式，还需要进行包括非线性效应在内的进一步研究，这有助于将来将物理机制应用于聚变反应堆。
{"title":"Observation of a stationary double transport barrier in KSTAR","authors":"M.W. Lee, S.-H. Hahn, D. Kim, J. Kang, W.H. Ko, J. Jang, W. Lee and C. Sung","doi":"10.1088/1741-4326/ad521a","DOIUrl":"https://doi.org/10.1088/1741-4326/ad521a","url":null,"abstract":"We have observed a stationary high confinement regime with a double transport barrier (DTB), including both internal and edge transport barriers (ITB and ETB) but without edge-localized modes (ELMs), in KSTAR. The ELM-free DTB phase has high thermal confinement comparable to typical H-mode operation in KSTAR. We investigated the characteristics of the DTB phase through various analyses. Transport analysis shows a reduction of ion heat diffusivity to near neoclassical level after the transition from the ELMy H-mode phase to the DTB phase. This result supports the formation of an ion ITB during the DTB phase. Furthermore, we observed that the DTB phase had an edge thermal transport barrier in the ion temperature profile, comparable to that of the H-mode, without a particle transport barrier at the edge. Peeling-ballooning stability analysis indicates that a lower pressure gradient due to density decrease in the DTB phase is mainly responsible for the ELM-free operation. Linear gyrokinetic analysis shows that the real frequency of the most unstable mode in the core region ( = 0.32–0.47) is in the ion diamagnetic direction at both H-mode and DTB phases. At the DTB phase, the linear growth rate inside the ITB is reduced by 50% compared to the ITB foot, while the reduction is not shown at the H-mode phase. Further investigation including nonlinear effects will be needed to better understand the unique operation mode, which can contribute to applying the physical mechanism to fusion reactors in the future.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"32 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-06-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141530815","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 : 2024-06-19DOI: 10.1088/1741-4326/ad52a9
N. Tsujii, A. Ejiri, Y. Ko, Y. Peng, K. Iwasaki, Y. Lin, K. Shinohara, O. Watanabe, S. Jang, T. Hidano, Y. Shirasawa, Y. Tian, F. Adachi and C.P. Moeller
Establishment of an efficient central solenoid (CS) free tokamak plasma start-up method may lead to an economical fusion reactor. CS-free start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Plasma current of about a quarter of CS-driven discharges has been obtained fully non-inductively using the outer-midplane and top LH launchers. Recently, an outer-off-midplane LH launcher was developed to achieve higher plasma current by optimizing for core absorption and minimal fast electron losses. Using the (outer-)off-midplane launcher, fully non-inductive plasma current start-up up to about 8 kA was achieved. Coupled ray-tracing and Fokker–Planck simulation was performed on equilibria reconstructed with an extended MHD model. It was found that the experimentally observed plasma current was in reasonable agreement with the numerical simulation. The simulation predicted appreciable orbit losses for the off-midplane launcher driven discharge at the present parameters, which was consistent with the experimentally observed x-ray radiation characteristics. The simulation showed that the current density was saturated for the present off-midplane launcher discharges and higher density and higher LH power was necessary to achieve higher plasma current.
{"title":"Studies of the outer-off-midplane lower hybrid wave launch scenario for plasma start-up on the TST-2 spherical tokamak","authors":"N. Tsujii, A. Ejiri, Y. Ko, Y. Peng, K. Iwasaki, Y. Lin, K. Shinohara, O. Watanabe, S. Jang, T. Hidano, Y. Shirasawa, Y. Tian, F. Adachi and C.P. Moeller","doi":"10.1088/1741-4326/ad52a9","DOIUrl":"https://doi.org/10.1088/1741-4326/ad52a9","url":null,"abstract":"Establishment of an efficient central solenoid (CS) free tokamak plasma start-up method may lead to an economical fusion reactor. CS-free start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Plasma current of about a quarter of CS-driven discharges has been obtained fully non-inductively using the outer-midplane and top LH launchers. Recently, an outer-off-midplane LH launcher was developed to achieve higher plasma current by optimizing for core absorption and minimal fast electron losses. Using the (outer-)off-midplane launcher, fully non-inductive plasma current start-up up to about 8 kA was achieved. Coupled ray-tracing and Fokker–Planck simulation was performed on equilibria reconstructed with an extended MHD model. It was found that the experimentally observed plasma current was in reasonable agreement with the numerical simulation. The simulation predicted appreciable orbit losses for the off-midplane launcher driven discharge at the present parameters, which was consistent with the experimentally observed x-ray radiation characteristics. The simulation showed that the current density was saturated for the present off-midplane launcher discharges and higher density and higher LH power was necessary to achieve higher plasma current.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"66 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-06-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141530816","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 : 2024-06-17DOI: 10.1088/1741-4326/ad52a4
Kenji Imadera, Yasuaki Kishimoto and Akihiro Ishizawa
Aiming at a fuel supply through particle pinch effects, turbulent particle transport is studied by gyrokinetic flux-driven Ion-Temperature-Gradient/Trapped-Electron-Mode (ITG/TEM) simulations. It is found that ITG/TEM turbulence can drive ion particle pinch by E × B drift (n ≠ 0) when the ion temperature gradient is steep enough. Electron particle pinch is also driven by E × B drift (n ≠ 0) in the case with the steep electron temperature gradient. Such an electron particle pinch can trigger an ambipolar electric field, leading to additional ion particle pinch by not only magnetic drift but also E × B drift (n = 0). These results suggest that a density peaking of bulk ions due to turbulent fluctuations can be achieved by sufficiently strong both ion and electron heating.
为了通过粒子夹持效应实现燃料供应，我们通过陀螺动量驱动的离子温度梯度/俘获电子模式（ITG/TEM）模拟研究了湍流粒子输运。研究发现，当离子温度梯度足够陡峭时，ITG/TEM湍流可以通过E × B漂移（n ≠ 0）驱动离子粒子挤压。在电子温度梯度陡峭的情况下，电子粒子夹也会受到 E × B 漂移（n ≠ 0）的驱动。这样的电子粒子挤压会引发环极性电场，导致不仅磁漂移而且 E × B 漂移（n = 0）产生额外的离子粒子挤压。这些结果表明，通过足够强的离子和电子加热，可以实现由湍流波动引起的体离子密度峰值。
{"title":"Turbulent particle pinch in gyrokinetic flux-driven ITG/TEM turbulence","authors":"Kenji Imadera, Yasuaki Kishimoto and Akihiro Ishizawa","doi":"10.1088/1741-4326/ad52a4","DOIUrl":"https://doi.org/10.1088/1741-4326/ad52a4","url":null,"abstract":"Aiming at a fuel supply through particle pinch effects, turbulent particle transport is studied by gyrokinetic flux-driven Ion-Temperature-Gradient/Trapped-Electron-Mode (ITG/TEM) simulations. It is found that ITG/TEM turbulence can drive ion particle pinch by E × B drift (n ≠ 0) when the ion temperature gradient is steep enough. Electron particle pinch is also driven by E × B drift (n ≠ 0) in the case with the steep electron temperature gradient. Such an electron particle pinch can trigger an ambipolar electric field, leading to additional ion particle pinch by not only magnetic drift but also E × B drift (n = 0). These results suggest that a density peaking of bulk ions due to turbulent fluctuations can be achieved by sufficiently strong both ion and electron heating.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"18 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141530747","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 : 2024-06-17DOI: 10.1088/1741-4326/ad521b
Vincent Graber and Eugenio Schuster
For ITER operations, the range of desirable burning-plasma regimes with high fusion power output will be restricted by various operational constraints. These constraints include the saturation of ITER’s various heating and fueling actuators such as the neutral beam injectors, the ion and electron cyclotron heating systems, the gas puffing system, and the deuterium–tritium pellet injectors. In addition to these actuator constraints, the H-mode power threshold, divertor detachment, and the heat load on the divertor targets may apply limitations to ITER’s operational space. In this work, Plasma Operation Contour (POPCON) plots that map the aforementioned constraints to the temperature-density space are used to investigate which constraints are most limiting towards accessing regimes with high fusion power output. The presented POPCON plots are based on a control-oriented core-edge model that couples the nonlinear density and energy response models for the core-plasma region with SOLPS4.3 parameterizations for conditions in the edge-plasma regions (scrape-off-layer and divertor). Using this control-oriented core-edge model, a nonlinear burn controller, which aims to regulate the plasma temperature and density in the core-plasma region, is constructed in this work. This controller is augmented with an online optimization scheme that governs the control references such that the plasma can be guided towards regimes with high fusion powers while protecting the divertor targets from dangerously high heat loads. A closed-loop simulation study illustrates the capability of this burn control scheme.
{"title":"Divertor-safe nonlinear burn control based on a SOLPS parameterized core-edge model for ITER","authors":"Vincent Graber and Eugenio Schuster","doi":"10.1088/1741-4326/ad521b","DOIUrl":"https://doi.org/10.1088/1741-4326/ad521b","url":null,"abstract":"For ITER operations, the range of desirable burning-plasma regimes with high fusion power output will be restricted by various operational constraints. These constraints include the saturation of ITER’s various heating and fueling actuators such as the neutral beam injectors, the ion and electron cyclotron heating systems, the gas puffing system, and the deuterium–tritium pellet injectors. In addition to these actuator constraints, the H-mode power threshold, divertor detachment, and the heat load on the divertor targets may apply limitations to ITER’s operational space. In this work, Plasma Operation Contour (POPCON) plots that map the aforementioned constraints to the temperature-density space are used to investigate which constraints are most limiting towards accessing regimes with high fusion power output. The presented POPCON plots are based on a control-oriented core-edge model that couples the nonlinear density and energy response models for the core-plasma region with SOLPS4.3 parameterizations for conditions in the edge-plasma regions (scrape-off-layer and divertor). Using this control-oriented core-edge model, a nonlinear burn controller, which aims to regulate the plasma temperature and density in the core-plasma region, is constructed in this work. This controller is augmented with an online optimization scheme that governs the control references such that the plasma can be guided towards regimes with high fusion powers while protecting the divertor targets from dangerously high heat loads. A closed-loop simulation study illustrates the capability of this burn control scheme.","PeriodicalId":19379,"journal":{"name":"Nuclear Fusion","volume":"24 1","pages":""},"PeriodicalIF":3.3,"publicationDate":"2024-06-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141530745","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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