聚变科学与工程资源库
检索物理代码、实验数据、代理模型、工程几何、技术文档与验证资料，并核对来源、版本和使用条件。
Bonding of Tungsten and Graphite Using Spark Plasma Sintering for Divertor Component in LHD
Japan Society of Plasma Science and Nuclear Fusion Research已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
FIREFLY: heat load and particle exhaust approximations for rapid evaluation of divertor designs
IOP PublishingAbstract The divertor in a magnetic confinement fusion reactor is an essential
component for power dissipation and particle removal. This article introduces the
FIREFLY package for rapid evaluation of divertor designs based on an extension of
the FLARE code for field line reconstruction from an unstructured flux tube mesh. First, divertor heat
loads are approximated with a simplified heat transport model. Neutralized particles
are then sampled from the resulting distribution, and the EIRENE code is used to
track molecules and atoms in a plasma background while accounting for dissociation,
charge exchange and ionization. Particles are removed on pumping surfaces in order
to estimate the exhaust efficiency for a given divertor geometry. Optimization of the
divertor geometry for more efficient particle exhaust is explored by using W7-X as an
example, and the sensitivity to model parameters for the plasma background in the
proxy calculations is evaluated.
Heat Transfer Performance of EVAPORON-3 Developed for an Enlarged Heat Transfer Surface of Divertor
Japan Society of Plasma Science and Nuclear Fusion Research已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
6.7 Plasma Facing Component (Divertor)
Japan Radioisotope Association已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Heat pipe technology based divertor plasma facing component concept for European DEMO
Elsevier BV已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Impact of the plasma geometry on the divertor power exhaust in a magnetic fusion reactor
Agence Bibliographique de l'Enseignement SupérieurImpact de la géométrie du plasma sur l'extraction de puissance au divertor d'un réacteur à fusion magnétique Une compréhension profonde du transport du plasma au bord d'un réacteur à fusion par confinement magnétique est obligatoire pour gérer l'extraction de puissance. Dans les dispositifs de fusion de nouvelle génération, des limites technologiques contraignent le flux de chaleur maximal au divertor. Pour une puissance d'échappement donnée le flux de chaleur maximal est déterminé par l'amplitude de l'empreinte du plasma au mur. Les profils de flux de chaleur au divertor peuvent être paramétrés par deux échelles de longueur du transport. Nous remettons en question l'interprétation actuelle de ces deux échelles de longueur en étudiant l'impact de la géométrie du divertor sur l'échappement. En particulier, un élargissement des profils de flux de chaleur avec la longueur de la jambe du divertor externe est diagnostiqué. Des efforts de modélisation ont montré que les simulations diffusives reproduisent les profils expérimentaux de flux de chaleur pour les plasmas à jambes courtes. Inversement, l'étalement du flux de chaleur pour une longe jambe du divertor est reproduit par un modèle turbulent, soulignant l'importance de la turbulence aussi dans le divertor. Ces résultats remettent en question l'interprétation de la largeur du flux de chaleur comme grandeur liée a la main SOL uniquement. Les configurations magnétiques avec une longe jambe du divertor mettent en évidence l'importance du transport asymétrique dans le divertor. Nous concluons que le transport dans la main SOL et celui dans le divertor ne sont pas à découpler et nous soulignons l'importance de la géométrie magnétique sur le transport turbulent avec l'avantage potentiel d'un inattendu étalement du dépôt de puissance.
Physics basis for the reference flat-top plasma scenario in the ST–E1 fusion power plant
DOE OSTI已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Tokamak Energy’s pre-concept design for a fusion power plant: an overview of ST-E1
DOE OSTI已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Analytic 1D approximation of the divertor broadening S in the divertor region for conductive heat transport
IOP PublishingAbstract An analytic 1D approximation for the divertor broadening S is introduced, depending only on the electron temperature between X-point and target. It is compared to simulations solving the 2D heat diffusion equation, in order to describe the divertor broadening along a field line solely by the ratio of the perpendicular to the parallel diffusivities. By assuming the temperature dependence of these two diffusivities an integral form of S is derived for the area along the separatrix between X-point and target. Integration along the separatrix results in an approximation for S , being in agreement with the 2D simulations. This approximation is furthermore compared to recent studies, which find a power law with negative exponent to describe S in terms of target temperature. This dependence is not reproduced in a pure conductive description, which instead shows a finite S for zero target temperature. This points to other mechanisms changing the shape of the heat flux profile—by additional widening or radiation losses—not included in the presented reduced approximation.
Achievement of a high-density, high-confinement, and high-beta tokamak plasma regime in DIII-D, and implications for a lower-current path for ITER and FPP
IOP PublishingExperiments on DIII-D have demonstrated a density-confinement synergy that enables sustainment of high performance in a previously unattained parameter regime of simultaneous very high energy confinement quality (H<sub>98y2</sub> ≥ 1.5), very high line-average density Greenwald fraction (ƒ<sub>Gr</sub> = πa<sup>2</sup>< n >/I<sub>P</sub> ≥ 1.4), and high toroidal beta (β<sub>T</sub> ≥ 3%). Tokamak operation in this regime is essential for a compact steady-state FPP, as well as for Q=10 with 500 MW of fusion power in ITER at I<sub>P</sub> << 15 MA. These experiments leveraged the knowledge that, in the high-poloidal-beta (β<sub>P</sub>) regime, impurity and density gradients can enhance turbulence stabilization caused by high α<sub>MHD</sub> (α<sub>MHD</sub>~(dβ<sub>P</sub>)⁄dr). This was described by theoretical predictions and gyrokinetic transport simulations [M.T. Kotschenreuther et al, 2024 Nucl. Fusion, 64 076033], and later confirmed by experiments on DIII-D [S. Ding et al, 2024 Nature 629 555]. To increase both β<sub>P</sub> and β<sub>T</sub>, the new experiments increased the ideal-wall stability β<sub>N</sub>-limit by using a smaller plasma-outer wall distance and higher triangularity in the plasma cross section (top/bottom average δ~0.9), enabled by the recent “shape & volume rise” (SVR) modification to the DIII-D divertor. The higher triangularity also contributed to achieving higher ƒ<sub>Gr</sub> by enabling higher pedestal density. At high density, the pedestal is ballooning limited and exhibits small and frequent ELMs, while the divertor is near detachment even without any impurity seeding. High plasma performance was attained and sustained reproducibly, with the eventual terminations brought about by an MHD mode destabilized as the current profile slowly continued to evolve. A path to stationary fully noninductive operation might include ECH injection to reduce both core impurity accumulation and the electron collisionality, thus increasing the bootstrap current. These experiments provide the first experimental demonstration of the ƒ<sub>Gr</sub>, H<sub>98y2</sub>, and β<sub>T</sub> values required simultaneously for ITER Q = 10 at I<sub>P</sub> < 10 MA, pointing to practical ways to improve the energy confinement in a fusion reactor.
Thermal stress intensity factor of interfacial cracks of a plasma facing component under high heat flux loading
Elsevier BV已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Recent advances in plasma control and physics research in the Large Helical Device
IOP Science - IAEAThe Large Helical Device (LHD), the largest superconducting helical system in the world, is equipped with advanced heating and diagnostic tools, facilitating plasma control and physics research. Data assimilation was employed for electron temperature control using a real-time Thomson scattering system and real time prediction code. A virtual LHD environment enabled visualization of escaping high-energy tritium ions and demonstrated that these ions impact the rear side of the divertor plate. Pioneering results crucial to plasma control have also been achieved. Real-time wall conditioning using Lithium granule dropping improved bulk ion energy and particle transport while simultaneously enhancing the heavy impurity transport. Progress has also been made in the investigation of turbulence-driven transport. At the confinement bifurcation, ion-scale turbulence decreased, while electron-scale turbulence increased. A change in the anisotropy of turbulent eddies was also observed at the confinement bifurcation. Coexistence of local and non-local turbulence was identified in electron-scale turbulence. Non-local turbulence exhibited the rapid spatial propagation of perturbations throughout the plasma, while local turbulence followed the temperature gradient. A transition between drift-wave turbulence and magnetohydrodynamics (MHD) turbulence was observed with the turbulence minimized at the transition condition. Machine learning analysis was employed to evaluate the temperate and density conditions of this turbulence transition. Then, real-time control of fueling and heating was applied to maintain the turbulence transition condition, improving the energy confinement enhancement factor by 20%. In addition, evidence was obtained for collisionless ion heating by energetic-ion-driven geodesic acoustic modes and MHD bursts. These achievements represent unique contributions to the development of fusion reactors.
Comparison between FEM and high heat flux thermal fatigue testing results of ITER divertor plasma facing mock-ups
Elsevier BV已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Material ejection and surface morphology changes during transient heat loading of tungsten as plasma-facing component in fusion devices
IOP Publishing已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Advances in 3D transient plasma dynamics and control through MHD and hybrid fluid-kinetic simulations with JOREK
DOE OSTI已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Time-dependent scenario modeling for the ST-E1 fusion power plant
DOE OSTI已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
209Chapter 7 Manufacture and evaluation of tungsten-copper modules for fusion reactor divertor
De Gruyter已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
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.
(Critical topics of plasma facing materials/plasma facing component data for the next step fusion devices)
Office of Scientific and Technical Information (OSTI)已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Activation analysis for the plasma facing component of ITER divertor
IEEE已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Assessing plasma face component thermal response to rotating 3D magnetic fields for SPARC tokamak
American Institute of Physics (AIP)Thermal response simulations of plasma-facing components (PFCs) in the SPARC tokamak, performed with the HEAT code, show that three-dimensional (3D) heat loads resulting from stationary n=1 perturbations require highly radiative scenarios, with up to 95% of the power crossing the separatrix (P<sub>SOL</sub>) being radiated, to maintain PFC temperatures within acceptable operational limits, whereas the application of slowly rotating 3D fields substantially reduces the thermal loads. The HEAT module, developed to predict heat loads from non-axisymmetric plasmas, is extended to model time-dependent heat flux patterns generated by rotating 3D fields, and a comprehensive thermal analysis is performed on PFCs subjected to both the maximum and minimum power loads, as well as to rotating heat flux distributions, to evaluate the temperature evolution for varying perturbation amplitudes and rotation frequencies. The extension of this analysis to 3D fields with toroidal mode number n=2 shows that this configuration leads to weaker localized heat flux peaks relative to the n=1 case, enabling safe operation with less than 80% of the power radiated when static 3D fields of low amplitude are applied, while using slowly rotating fields at higher amplitudes. These results indicate that n=2 perturbations are generally less detrimental to divertor power exhaust, emphasizing the strong dependence of divertor power exhaust on the characteristics of the applied 3D fields.
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.
Ultrasonic techniques for quality assessment of ITER Divertor plasma facing component
Elsevier BV已收录公开题录与来源记录，正文内容及使用条件请通过原始发布页面核对。
Autoregressive long-horizon prediction of plasma edge dynamics<sup>*</sup>
IOP Science - IAEAAccurate modeling of scrape-off layer (SOL) and divertor-edge dynamics is vital for designing plasma-facing components in fusion devices. High-fidelity edge fluid/neutral codes such as SOLPS-ITER capture SOL physics with high accuracy, but their computational cost limits broad parameter scans and long transient studies. We present transformer-based, autoregressive surrogates for efficient prediction of 2D, time-dependent plasma edge state fields. Trained on SOLPS-ITER spatiotemporal data for the KSTAR tokamak, the surrogates forecast electron temperature, electron density, and radiated power over extended horizons. We evaluate model variants trained with increasing autoregressive horizons (1–100 steps) on short- and long-horizon prediction tasks. Longer-horizon training systematically improves rollout stability and mitigates error accumulation, enabling stable predictions over hundreds to thousands of steps and reproducing key dynamical features such as the motion of high-radiation regions. Measured end-to-end wall-clock times show the surrogate is orders of magnitude faster than SOLPS-ITER, enabling rapid parameter exploration. Prediction accuracy degrades when the surrogate enters physical regimes not represented in the training dataset, motivating future work on data enrichment and physics-informed constraints. Overall, this approach provides a fast, accurate surrogate for computationally intensive plasma edge simulations, supporting rapid scenario exploration, control-oriented studies, and progress toward real-time applications in fusion devices.