Understanding and predicting divertor heat-load width $λ_q$ is a critically important problem for an easier and more robust operation of ITER with high fusion gain. Previous predictive simulation data for $λ_q$ using the extreme-scale edge gyrokinetic code XGC1 in the electrostatic limit under attached divertor plasma conditions in three major US tokamaks [C.S. Chang et al., Nucl. Fusion 57, 116023 (2017)] reproduced the Eich and Goldston attached-divertor formula results [formula #14 in T. Eich et al., Nucl. Fusion 53, 093031 (2013); R.J. Goldston, Nucl. Fusion 52, 013009 (2012)], and furthermore predicted over six times wider $λ_q$ than the maximal Eich and Goldston formula predictions on a full-power (Q = 10) scenario ITER plasma. After adding data from further predictive simulations on a highest current JET and highest-current Alcator C-Mod, a machine learning program is used to identify a new scaling formula for $λ_q$ as a simple modification to the Eich formula #14, which reproduces the Eich scaling formula for the present tokamaks and which embraces the wide $λ_q^X{GC}$ for the full-current Q = 10 ITER plasma. The new formula is then successfully tested on three more ITER plasmas: two corresponding to long burning scenarios with Q = 5 and one at low plasma current to be explored in the initial phases of ITER operation. The new physics that gives rise to the wider $λq_^{XGC} is identified to be the weakly-collisional, trapped-electron-mode turbulence across the magnetic separatrix, which is known to be an efficient transporter of the electron heat and mass. Electromagnetic turbulence and high-collisionality effects on the new formula are the next study topics for XGC1.
It is found that the X-Divertor (XD) configuration [1-3] can be made with the conventional PF coil set on ITER[4], where all PF coils are outside the TF coils. Desirable configurations are possible where the PF currents are below the present maximum design limits on ITER, and where the baseline divertor cassette is used. It is possible that the XD could be used to assist in high-power operation on ITER, but some further issues need examination. Note that the increased major radius of the Super X-Divertor (SXD) [5-8] is not a feature of the XD geometry. In addition, we present an XD configurati…
The XGC1 edge gyrokinetic code is used for a high fidelity prediction for the width of the heat-flux to divertor plates in attached plasma condition. The simulation results are validated against the empirical scaling $λ_q \propto B_P^{-γ}$ obtained from present tokamak devices, where $λ_q$ is the divertor heat-flux width mapped to the outboard midplane and $γ_q=1.19$ as defined by T. Eich et al. [Nucl. Fusion 53 (2013) 093031], and $B_P$ is the magnitude of the poloidal magnetic field at outboard midplane separatrix surface. This empirical scaling predicts $λ_q \leq 1mm$ when extrapolated to I…
Investigations of parametric scaling of power exhaust in the alternative divertor configuration (ADC) SOLPS-ITER simulation database of the EU-DEMO are conducted and compared to predictions based on the Lengyel model. The Lengyel model overpredicts the necessary argon concentrations for LFS divertor detachment by about a factor of 5-10 relative to the SOLPS-ITER simulations. Therefore, while the Lengyel model predicts that plasmas with accetable divertor heat loads in EU-DEMO would exceed the tolerable upstream impurity concentrations by a large margin, there are several SOLPS-ITER solutions w…
For the ITER-like wall, the JET mirror link divertor spectroscopy system was redesigned to fully cover the tungsten horizontal strike plate with faster time resolution and improved near-UV performance. Since the ITER-like wall project involves a change in JET from a carbon dominated machine to a beryllium and tungsten dominated machine with residual carbon, the aim of the system is to provide the recycling flux, equivalent, to the impinging deuterium ion flux, the impurity fluxes (C, Be, O) and tungsten sputtering fluxes and hence give information on the tungsten divertor source. In order to d…
The lithium vapor box divertor aims to detach the divertor plasma via evaporating and condensing lithium surfaces. By evaporating lithium near or at the divertor plate and condensing it closer to the main chamber, a lithium vapor density gradient can be created. This density gradient ties energy losses to poloidal distance between the target and the detachment point. The radiation zone is then prevented from reaching the X-point as the lithium ionization rate decreases when the detachment front moves away from the divertor target. Here we present Scrape Off Layer Plasma Simulator (SOLPS) simul…
It is planned to use atomic processes to spread out most of the heating power over the first wall and side walls to reduce the heat loads on the plasma facing components in ITER to ~ 50 MW. Calculations indicate that there will be 100 MW in bremstrahlung radiation from the plasma center, 50 MW of radiation from the plasma edge inside the separatrix and 100 MW of radiation from the scrape-off layer and divertor plasma, leaving 50 MW of power to be deposited on the divertor plates. The radiation losses are enhanced by the injection of impurities such as Neon or Argon at acceptably low levels (~0…
Reduction of the peak heat loads on the plasma facing components is essential for the success of the next generation of high fusion power tokamaks such as the International Thermonuclear Experimental Reactor (ITER) 1 . Many present concepts for accomplishing this involve the use of atomic processes to transfer the heat from the plasma to the main chamber and divertor chamber walls and much of the experimental and theoretical physics research in the fusion program is directed toward this issue. The results of these experiments and calculations are the result of a complex interplay of many proce…
The free-streaming plus recycling model (FSRM) has recently been developed to understand and predict tungsten gross erosion rates from the divertor during edge localized modes (ELMs). In this work, the FSRM was tested against experimental measurements of W sputtering during ELMs, conducted via fast WI spectroscopy. Good agreement is observed using a variety of controlling techniques, including gas puffing, neutral beam heating, and plasma shaping to modify the pedestal stability boundary and thus the ELM behavior. ELM mitigation by pellet pacing was observed to strongly reduce W sputtering by …
Harnessing Toroidal Neutral Flows to Enhance Divertor Particle Exhaust
Gauss Fusion GmbH（高斯聚变有限公司）；Oak Ridge National Laboratory（橡树岭国家实验室）；United Kingdom Atomic Energy Authority（英国原子能管理局）；University of York（约克大学）；RI Research Instruments GmbH（RI研究仪器有限公司）
The guiding-center kinetic neoclassical transport code, XGC0, [C.S. Chang et. al, Phys. Plasmas 11, 2649 (2004)] is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that width of the XGC0 neoclassical heat-loa…
Edge localised modes (ELMs) are a concern for future devices, such as ITER, due to the large transient heat loads they generate on the divertor surfaces which could limit the operational time of the device. This paper discusses the application of resonant magnetic perturbations (RMPs) as a mechanism for ELM control on MAST. Experiments have been performed using an n=3 toroidal mode number perturbation and measurements of the strike point splitting performed. The measurements have been made using both infrared and visible imaging to measure the heat and particle flux to the divertor. The measur…
The power load to the divertor surfaces is a key concern for future devices such as ITER, due to the thermal limits on the material surface. One factor that characterizes the heat flux to the divertor is the fall off length in the scrape-off layer (SOL), which recent empirical scalings have shown could be as small as 1 mm. These predictions are based on a multi-machine scaling of the heat flux width fitted using an expression for the divertor heat flux profile which includes a term for the exponential decay in the SOL and diffusion about the last closed flux surface (LCFS) in the private flux …
Divertor plasma detachment is likely needed for the function of magnetically confined nuclear fusion. It greatly reduces the particle and heat flux incident on a target, and thus reduces the sputtering and heat loading on the target. It is therefore advantageous to have accurate simulations of plasma behaviour in the divertor to design future Tokamak divertors. A common simulation package used for this purpose is SOLPS-ITER, which has previously been observed to be underestimating the contribution of molecular effects to plasma detachment. To correct this, its reaction rate for molecular charg…
High-power TCV scenario for conventional and alternative divertor studies
Max Planck Institute for Plasma Physics（马克斯·普朗克等离子体物理研究所）；Eindhoven University of Technology（埃因霍温理工大学）；DIFFER - Dutch Institute for Fundamental Energy Research（荷兰基础能源研究院）
For understanding carbon erosion and redeposition in nuclear fusion devices, it is important to understand the transport and chemical break-up of hydrocarbon molecules in edge plasmas, often diagnosed by emission of the CH A^2Δ- X^2ΠGerö band around 430 nm. The CH A-level can be excited either by electron-impact or by dissociative recombination (D.R.) of hydrocarbon ions. These processes were included in the 3D Monte Carlo impurity transport code ERO. A series of methane injection experiments was performed in the high-density, low-temperature linear plasma generator Pilot-PSI, and simulated em…
MAST-U has recently started operating with a Super-X divertor, designed to increase total flux expansion and neutral trapping, both predicted through simple analytic models and SOLPS calculations to reduce the plasma and impurity density detachment thresholds. In this study, utilising the SOLPS-ITER code, we are quantifying the possible gain allowed by the MAST-U Super-X and neutral baffling geometry, in terms of access to detachment. We show that a significant reduction of the upstream density detachment threshold (up to a factor 1.6) could be achieved in MAST-U, for the Super-X, as opposed t…
The Mega Amp Spherical Tokamak (MAST) is the centre piece of the UK fusion research programme. In 2010, a MAST Upgrade programme was initiated with three primary objectives, to contribute to: 1) Testing reactor concepts (in particular exhaust solutions via a flexible divertor allowing Super-X and other extended leg configurations); 2) Adding to the knowledge base for ITER (by addressing important plasma physics questions and developing predictive models to help optimise ITER performance of ITER) and 3) Exploring the feasibility of using a spherical tokamak as the basis for a fusion Component T…
Initial evaluation of miniature ultra-high-field commercial stellarator reactors with breeding external to resistive coils
Laboratorio Nacional de Fusión, CIEMAT（国家聚变实验室，CIEMAT）；Departamento de Física, Universidad Carlos III de Madrid（马德里卡洛斯三世大学物理系）；Oak Ridge National Laboratory (ORNL)（橡树岭国家实验室）；University of Texas, Austin（德克萨斯大学奥斯汀分校）
The role of impurity radiation in the reduction of heat loads on divertor plates in present experiments such as DIII-D, JET, JT-60, ASDEX, and Alcator C-Mod, and in planned experiments such as ITER and TPX places a new degree of importance on the accuracy of impurity radiation emission rates for electron temperatures below 250 eV for ITER and below 150 eV for present experiments. We have calculated the radiated power loss using a collisional radiative model for Be, B, C, Ne and Ar using a multiple configuration interaction model which includes density dependent effects, as well as a very detai…
First mirrors will be the plasma facing components of optical diagnostic systems in ITER. Mirror surfaces will undergo modification caused by erosion and re-deposition processes [1,2]. As a consequence, the mirror performance may be changed and may deteriorate [3,4]. In the divertor region it may also be obscured by deposition [5-7]. The limited access to in-vessel components of ITER calls for testing the mirror materials in present day devices in order to gather information on the material damage and degradation of the mirror performance, i.e. reflectivity. A dedicated experimental programme,…
Tokamak plasmas are confined by a magnetic field that limits the particle and heat transport perpendicular to the field. Parallel to the field the ionised particles can move freely, so to obtain confinement the field lines are "closed" (ie. form closed surfaces of constant poloidal flux) in the core of a tokamak. Towards, the edge, however, the field lines begin to intersect physical surfaces, leading to interaction between neutral and ionised particles, and the potential melting of the material surface. Simulation of this interaction is important for predicting the performance and lifetime of…
The scrape-off layer power width (λ_q) is an important parameter for characterizing the divertor heat loads. Many experimental, theoretical, and numerical studies have been performed in recent years. In this paper, a 2D electrostatic turbulence code, BOUT-HESEL, has been upgraded to simulate H-mode plasmas for the first time. The code is validated against the previous implementation and the experiments. The simulated λ_q is found to agree quite well with the Eich scaling for the EAST H-mode discharge. The code is utilized to simulate the ITER 15MA baseline scenario. The ITER simulation reveals…
The aim of the JET ITER-like Wall Project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk red…
Neutral tungsten is the primary candidate as a wall material in the divertor region of the International Thermonuclear Experimental Reactor (ITER). The efficient operation of ITER depends heavily on precise atomic physics calculations for the determination of reliable erosion diagnostics, helping to characterise the influx of tungsten impurities into the core plasma. The following paper presents detailed calculations of the atomic structure of neutral tungsten using the multiconfigurational Dirac-Fock method, drawing comparisons with experimental measurements where available, and includes a cr…
A coherent structure transport model for scrape-off layer turbulence
Renewable and Sustainable Energy Institute, University of Colorado（科罗拉多大学可再生与可持续能源研究所）；Lodestar Research Corporation（洛德斯特研究公司）；Center for Integrated Plasma Studies, University of Colorado（科罗拉多大学综合等离子体研究中心）；General Atomics（通用原子能公司）；Massachusetts Institute of Technology（麻省理工学院）
FAST (Fusion Advanced Studies Torus), the Italian proposal of a Satellite Facility to ITER, is a compact tokamak (R$_0$ = 1.82 m, a = 0.64 m, triangularity $δ$ = 0.4) able to investigate non-linear dynamics effects of $α$-particle behavior in burning plasmas and to test technical solutions for the first wall/divertor directly relevant for ITER and DEMO. Currently, ENEA is investigating the feasibility of a superconducting solution for the magnet system. This paper focuses on the analysis of the CS (Central Solenoid) magnet thermal behavior. In particular, considering a superconducting solution…
Impurities cause radiation losses and plasma dilution, and in stellarator plasmas the neoclassical ambipolar radial electric field is often unfavorable for avoiding strong impurity peaking. In this work we use a new continuum drift-kinetic solver, the SFINCS code (the Stellarator Fokker-Planck Iterative Neoclassical Conservative Solver) [M. Landreman et al., Phys. Plasmas 21 (2014) 042503] which employs the full linearized Fokker-Planck-Landau operator, to calculate neoclassical impurity transport coefficients for a Wendelstein 7-X (W7-X) magnetic configuration. We compare SFINCS calculations …
Autoregressive long-horizon prediction of plasma edge dynamics
Computational Sciences and Engineering Division, Oak Ridge National Laboratory（橡树岭国家实验室计算科学与工程部）；Fusion Energy Division, Oak Ridge National Laboratory（橡树岭国家实验室等离子体能源部）；Computer Science and Mathematics Division, Oak Ridge National Laboratory（橡树岭国家实验室计算机科学与数学部）
