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A novel reciprocating probe design for ion temperature measurements at the WEST tokamak lower divertor
Measurement Science and TechnologyAbstract A reciprocating retarding field analyzer has been installed in the lower divertor of the WEST tokamak to measure the local ion temperature in ITER-like tungsten divertor conditions. The paper presents the design of the analyzer, the robust magnetic…
INFLUENCE OF DRIFTS AND CURRENTS ON THE MAIN OPERATING PARAMETERS OF THE TOKAMAK T-15MD DIVERTOR
Физика плазмыThe work presents the results of the first calculations of the T-15MD tokamak near-wall plasma in the SOLPS-ITER code taking into account the effects of drifts and currents. The modes with the power passing through the separatrix PSOL = 6 MW and different H…
Self-consistent modeling of tokamak edge plasma transport with lithium sources
Physics of PlasmasMagnetic confinement fusion devices require effective heat and particle exhaust solutions on the divertor plates to operate sustainably, especially under reactor-relevant conditions. Liquid lithium divertors have been proposed to address two major challenge…
Application of ECWR for tin removal with prospective use in tokamaks with liquid metal divertor
Plasma Physics and Controlled FusionAbstract One of the problems of the liquid metal tokamak divertor technology is the redeposition of the released metal on the tokamak vessel walls. In this work an investigation of the removal of the redeposited metals from grounded conducting walls of toka…
Progress and innovations in the TCV tokamak research programme
Nuclear FusionAbstract Research on the Tokamak à Configuration Variable addresses a wide range of key questions relevant to ITER and future fusion power plants. Over the past two years, highly productive experimental campaigns have led to major advances across several ar…
Simulations of SMBI on high-field and low-field sides of the HL-2A tokamak
Plasma Physics and Controlled FusionAbstract Enhancing plasma injection efficiency and penetration depth is crucial for achieving high-performance steady-state fusion confinement in ITER, the next generation magnetic confinement fusion device. Based on the trans-neut module within the BOUT++…
Feasibility of Negative Triangularity Equilibria in the SPARC Tokamak
arXiv (Cornell University)We investigate the feasibility of negative triangularity (NT) plasma configurations in the SPARC tokamak, a compact, high-field device (up to 12.2 T) designed for positive triangularity (PT) operation. Using the FreeGS free-boundary equilibrium solver, we s…
Hermes-3 simulation of the low-n X-point mode driven by impurity in tokamak edge plasmas
Nuclear Materials and EnergyHermes-3, a multi-fluid plasma simulation tool is applied to investigate the low-n X-point mode driven by coupling between drift wave and impurity radiation condensation instability in tokamak edge plasmas. The radiation loss function of impurities under no…
High-throughput full-f gyrokinetics of the tokamak boundary
arXiv (Cornell University)Full-f global gyrokinetic simulations of the plasma boundary have until now required heroic computational efforts and case-by-case expert intervention, precluding systematic parameter scans. Here we demonstrate a paradigm shift: hundreds of independent, con…
Constructing field-aligned coordinate systems for gyrokinetic simulations of tokamaks in X-point geometries
Journal of Plasma PhysicsStructures in tokamak plasmas are elongated along the direction of the magnetic field and short in the directions perpendicular to the magnetic field. Many tokamak simulation codes take advantage of this by using a field-aligned coordinate system. However,…
Self-consistent plasma-neutrals fluid modeling of edge and scrape-off layer turbulence in diverted tokamaks
Plasma Physics and Controlled FusionAbstract We report on developments in the edge and scrape-off layer turbulence code GRILLIX to self-consistently simulate plasma turbulence coupled to a fluid neutral gas model. The neutrals model is extended from a single fluid moment (neutrals density) to…
Lithium Droplet Transport in Tokamak Edge Plasmas
arXiv (Cornell University)A lithium droplet transport and evaporation model has been developed within the Direct Simulation Monte Carlo code OpenEdge. This model integrates gravity, collisional ion drag, orbital-motion-limited charging, energy-balance evaporation, and an anisotropic…
First simultaneous observation of co- and counter-current fast-ion losses in the ASDEX Upgrade tokamak
Nuclear FusionAbstract In ITER and future fusion power plants, the source of the fusion born alpha particles is almost isotropic in pitch angle, thus having co- and counter-current populations. For trapped ions, the co-current side of the orbit corresponds to its outer l…
High-resolution visible spectroscopy for ion temperature and flow velocity measurements of the TCV divertor plasma
Review of Scientific InstrumentsA high-resolution spectroscopic system was installed on the Tokamak à Configuration Variable (TCV), whose toroidal lines-of-sight (LOS) cross the divertor region. This system simultaneously measures the Doppler broadening and Doppler shift of several spectr…
Engineering concept of the VNS - a beam-driven tokamak for component testing
Fusion Engineering and DesignThe qualification of in-vessel components for a fusion power plant requires a test environment with a high flux of 14 MeV neutrons over a sufficiently large surface and volume. Performance testing and qualification of the complex design and technologies of…
The impact of non-local parallel electron transport on plasma-impurity reaction rates in tokamak scrape-off layer plasmas
Plasma Physics and Controlled FusionAbstract Plasma-impurity reaction rates are a crucial part of modelling tokamak scrape-off layer (SOL) plasmas. To avoid calculating the full set of rates for the large number of important processes involved, a set of effective rates are typically derived w…
数据集CSV / XLSX / TXT多装置偏滤器脱靶供气数据库
Gauss Fusion / UKAEA / Max Planck IPP / KIT汇集托卡马克与仿星器脱靶研究中的燃料供气、杂质注入、装置几何和等离子体参数，用于比较脱靶运行空间与标度关系。
SOLPS-NN代理模型训练数据
Forschungszentrum Jülich由SOLPS-ITER参数扫描生成的训练与测试数据，覆盖边缘等离子体温度、密度、粒子与热通量、辐射及偏滤器靶板积分量，用于粒子与功率排散代理模型研究。
SOLPS-ITER 文档
SOLPS-ITER Community面向托卡马克边缘等离子体与中性粒子建模的软件文档入口。
Divertor detachment and heat exhaust mitigation control in KSTAR with tungsten divertor
IOP PublishingAbstract KSTAR has recently undergone an upgrade to use a new tungsten divertor to run experiments in ITER-relevant scenarios. Even with a high melting point of tungsten, it is important to control the heat flux impinging on tungsten divertor targets to minimize sputtering and contamination of the core plasma. Heat flux on the divertor is often controlled by increasing the degree of detachment of scrape-off layer plasma from the target plates. In this work, we have demonstrated successful divertor detachment and heat exhaust dissipation control experiments using two different methods. The first method uses attachment fraction as a control variable which is estimated using ion saturation current measurements from embedded Langmuir probes in the divertor. The second method uses a novel machine-learning-based surrogate model of 2D UEDGE simulation database, DivControlNN. We demonstrated running inference operation of DivControlNN in realtime to estimate heat flux at the divertor and use it as the control variable in a feedback loop with impurity gas flow. We present interesting insights from these experiments including a systematic approach to tuning controllers and discuss future improvements in the control infrastructure and control variables for future burning plasma experiments.
Power and particle exhaust for the ARC fusion power plant
Cambridge University PressTo successfully show that fusion is an attractive energy source, the ARCTM fusion power plant will need to operate with a robust, integrated power and particle exhaust solution. To maximise ARC’s fusion power output while avoiding excessive erosion of the plasma-facing components, we will need to radiatively dissipate most of the power crossing the last-closed flux surface, injecting radiating impurities such as argon or neon to access divertor detachment. Divertor detachment will need to be integrated with a high-performance core plasma, and with efficient impurity pumping to prevent the accumulation of helium ash in the core. To access and control detachment in high-performance plasmas, we have designed ARC with up–down-symmetric divertors, with secondary X-points in long, tightly baffled outer legs. Using a core-edge modelling workflow, we predict that with this divertor design, ARC will access detachment with modest argon seeding in the divertor (c<sub>Ar,div</sub> ∼0.9%), which should have minimal impact on the core ( Z<sub>eff,core</sub> <0.5) for reasonable argon enrichment (c<sub>Ar,div</sub>/c<sub>Ar,core</sub> =6). Due to the high predicted divertor neutral pressure (p<sub>div </sub>∼20 Pa), we predict that ARC will sufficiently pump helium to limit ash accumulation in the core (c<sub>He,core</sub> <2%) for a helium enrichment of c<sub>He,div</sub>/c<sub>He,core </sub>=0.4. ARC’s divertor design is expected to increase the stability of a detachment front in the outer divertor leg, helping to prevent divertor reattachment during smaller heat-flux transients such as scrape-off-layer filaments associated with the quasi-continuous exhaust regime. However, this buffering will not be sufficient to prevent divertor reattachment during large type-I edge-localised modes (ELMs), and as such these will need to be avoided on ARC. Experiments on SPARC will be used to select an integrated scenario which avoids or mitigates type-I-ELMs while maintaining access to detachment, good core fusion performance and sufficient impurity exhaust. SPARC experiments will also be used to finalise ARC’s divertor design, by studying the impact of magnetic and first-wall geometry on detachment stability, impurity enrichment and neutral baffling under conditions similar to those expected for ARC. In conclusion, our analysis finds that ARC will have a viable power and particle exhaust solution which is compatible with high-power operations, and this solution will be validated in experiments on SPARC.
Heat pipe technology based divertor plasma facing component concept for European DEMO
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Review of magnetic islands from the divertor perspective and a simplified heat transport model for the island divertor
IOP PublishingAbstract Magnetic islands in toroidal confinement devices are reviewed from the viewpoint of their divertor potential. Divertor-relevant geometric parameters are derived analytically, and the relationships among them are revealed. We explain how the island geometry limits the target length and demonstrate the importance of an appropriate numerical tool to minimize the risk of thermal overload of plasma-facing components in the divertor design. The currently available three-dimensional (3D) models are briefly discussed, and their strengths and weaknesses are evaluated. The highlight will be the introduction of a new energy transport model recently developed within the framework of the EMC3 code ( Feng et al 2004 Contrib. Plasma Phys. 44 57)—the so-called EMC3-Lite version—primarily for the design and optimization of 3D divertors involving thermal overload concerns. While still undergoing experimental validation with the current graphite divertor of W7-X, it is already being used to develop a subsequent tungsten divertor for W7-X.
First Wall Design of a Tokamak Pilot Plant Using a Monte Carlo Model for 3-D Heat Flux Deposition
DOE OSTIWe present a method for calculating the heat fluxes deposited on nonaxisymmetric tokamak first wall components, allowing for a first-of-its-kind model for power handling in the tokamak far scrape-off layer (SOL). The DIV3D Monte Carlo model features strict global power conservation and can calculate the finite cross-field plasma transport into magnetically-shadowed regions, which is significant when dealing with meter-scale shadows introduced by components such as poloidal limiters or antennas. As a case study, we apply the DIV3D model to inform the distribution of first wall poloidal limiters in an ARC-class reactor device. We demonstrate that discrete protection limiters can efficiently reduce peak heat fluxes on recessed breeder wall components in the presence of significant far-SOL plasma fluxes. By varying the toroidal periodicity and radial standoff depth of the limiters, we demonstrate one of the tradeoffs that must be considered in first wall design: more limiters provide greater protection, but at the cost of reduced breeding performance. We also present the impact that radial misalignments between limiters would have on first wall power loading.