Bulletin of the American Physical Society
61st Annual Meeting of the APS Division of Plasma Physics
Volume 64, Number 11
Monday–Friday, October 21–25, 2019; Fort Lauderdale, Florida
Session PO6: MF: Divertor Physics and Plasma-Material Interactions |
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Chair: Mike Mauel, Columbia University Room: Grand D |
Wednesday, October 23, 2019 2:00PM - 2:12PM |
PO6.00001: Present status of the divertor tokamak test facility Piero MARTIN, Raffaele Albanese, Flavio Crisanti, Aldo Pizzuto The Divertor Tokamak Test facility (DTT )is a superconducting tokamak with 6 T on-axis maximum toroidal magnetic field carrying plasma current up to 5.5 MA in pulses up to 100 s. The D-shaped device is up-down symmetric, with major radius R=2.14 m, minor radius a=0.64 m and average triangularity 0.3. The auxiliary heating power coupled to the plasma at maximum performance is 45 MW, shared between ECRH and ICRH and negative ion beams. DTT is designed with a high level of flexibility, in particular as far as divertor scenarios are concerned. The external coils together with a set of four internal coils will allow to control and optimize the local magnetic configuration in the vicinity of the divertor target. The main divertor magnetic topologies, which can be produced in DTT are the reference single null, double null and snowflake configurations. These can be produced at (or close) to the maximum target plasma current of 5.5 MA, while double super-X may be feasible only at significantly lower current. The DTT coil system also allows for the realization of scenarios with negative triangularity. A 5 MA single null scenario with delta =-0.13 and a double null scenario at 3.5 MA with delta=-0.38 can be produced. This paper will describe the present status of DTT realisation. [Preview Abstract] |
Wednesday, October 23, 2019 2:12PM - 2:24PM |
PO6.00002: The physical design of the advanced divertor for EAST and CFETR. Chaofeng Sang, Guosheng Xu, Liang Wang, Rui Ding, Xiaoju Liu The steady-state operation of next-step fusion devices requires both the deposited heat flux density on the divertor target below 10 MW/m$^{\mathrm{2}}$ and plasma temperature at the target below 5 eV to ensure adequate lifetime. Therefore, it will be essential to achieve highly dissipative or detached divertor conditions for the control of heat flux and erosion in a fusion reactor. One of the most effective methods to promote the achievement of detachment is to improve neutral trapping and impurity screening in the divertor by changing the divertor structure. In this work, the physical design of the lower tungsten divertor of EAST has been performed by using SOLPS simulation. A systematic analysis of the target shape and closure effects on the plasma detachment will be presented. Moreover, the application of the QSF equilibrium on the designed divertor shape is evaluated. Finally, preliminary CFETR divertor design is illustrated. *Supported by National Key R{\&}D Program of China 2017YFA0402500. [Preview Abstract] |
Wednesday, October 23, 2019 2:24PM - 2:36PM |
PO6.00003: Challenging current alternative divertor concepts Roberto Maurizio, Basil Duval, Benoit Labit, Holger Reimerdes, Christian Theiler, James Harrison, Nicolas Fedorczak An unprecedented range of innovative magnetic divertor concepts were realized on TCV to investigate their ability to enhance the target wetted area in attached conditions, as this reduces the power exhaust challenge by lowering the upstream SOL density required for detachment. Unexpectedly, the results reveal advantageous aspects of the standard low flux expansion Single Null for power exhaust. With the target field line grazing angle fixed by engineering limits, innovative concepts can enhance the wetted area by increasing either target major radius or the heat channel width. At fixed target radius, TCV experiments show two ways of increasing the wetted area. First the heat channel width scales with the square root of the divertor leg length. A long-legged divertor benefits, thus, of enhanced wetted area but has higher realization costs and aggravates heat handling at the inner plate, as a longer outer leg is seen to redistribute heat from the outer to the inner plate. Second and surprisingly, the heat spreading in the divertor and the heat channel width during ELMs scale with the inverse of target flux expansion. A small flux expansion divertor benefits, thus, of enhanced wetted area with increased tolerance to ELMs at usual realization costs. [Preview Abstract] |
Wednesday, October 23, 2019 2:36PM - 2:48PM |
PO6.00004: Real-time feedback control of the radiation front location in the TCV tokamak Matthijs Van Berkel, Tim Ravensbergen, Artur Perek, Cristian Galperti, Ricky Van Kampen, Joost Lammers, Olivier Fevrier, Stuart Henderson, Michael Komm, Dominik Brida, Christian Theiler, Basil Duval, Bryan Linehan, Marco De Baar In the detached regime in tokamak divertors, an actuator, often local gas puffing, induces plasma power loss mechanisms, leading to a significant particle and heat flux reduction at the divertor target. Such a detached regime requires real-time plasma monitoring and control of the gas puff actuator. Many existing plasma diagnostics have a low signal-to-noise ratio in detached conditions and/or are not real-time. In this work, we apply real-time acquisition and processing techniques to C-III filtered images from multi-spectral imaging diagnostic MANTIS. The detected location of the C-III emission front is mapped to the poloidal plane using a non-tomographic approach. This emission front is a proxy of divertor cooling, making its deduced position a spatially distributed and attractive controllable quantity. For a dedicated discharge on TCV, we apply system identification to generate a data-driven dynamic model, describing the effect of divertor gas fueling on the emission front location. From this model, a feedback controller was synthesized off-line, and tested on TCV. [Preview Abstract] |
Wednesday, October 23, 2019 2:48PM - 3:00PM |
PO6.00005: Scaling of L-mode heat flux for ITER and COMPASS-U divertors, based on five tokamaks J Horacek, J Seidl, P Vondracek, M Komm, K Jirakova, M Hron, F Jaulmes, J Adamek, G.F. Matthews, S Elmore, A Thornton, G Deng, X Gao, L Wang, R Ding, J Olsen, J.J. Rasmussen, A.H. Nielsen, V Naulin, D Brunner, B LaBombard, A Jardin, M Ezzat, K Camacho, Ch Guillemaut Based on {\em Nuclear Fusion} paper, we improve scalings of divertor L-mode power decay length. We employ data from tokamaks JET, EAST, MAST, Alcator C-mod and COMPASS and validate it against 2D turbulence simulation HESEL. The analysis covers 500 divertor heat flux profiles with 11 varying global plasma parameters. We see that two previously published scalings describe well only part of the database. We therefore derive 20 new scalings describing 86-93~\% of the measured decay length variability. We so-predict for attached highest current L-mode in ITER: 10-20~MW/m$^2$ surface perpendicular heat flux and twice that for COMPASS-Upgrade. [Preview Abstract] |
Wednesday, October 23, 2019 3:00PM - 3:12PM |
PO6.00006: Generalization of the Heuristic Drift Model of the SOL for Finite Collisionality Robert Goldston The “Heuristic Drift” (HD) model for the scrape off layer (SOL) power flux width \footnote{R.J. Goldston, Nucl. Fusion 52, 013009 (2012)} was explicitly formulated for low-gas-puff H-Mode conditions. Experimental results in these conditions have shown good agreement with the model. In 2015, however, ASDEX-Upgrade (AUG) data showed that the scrape-off width broadens as the collisionality increases \footnote{H.J. Sun et al., Plasma Phys. Control. Fusion 57, 125011 (2015)}, which is inconsistent with the HD model as formulated. We hypothesize that this broadening is due to enhanced residence time of heat in the scrape-off-layer (SOL) at higher collisionality, due to higher classical parallel thermal resistance. This allows more time for cross-field drifts to broaden the SOL. We find reasonable agreement with more extensive recent AUG data \footnote{T. Eich et al, EPS-DPP Conference, 2019} for SOL broadening at high collisionality. This broadening may play a synergistic role with turbulence in degrading global energy confinement. [Preview Abstract] |
Wednesday, October 23, 2019 3:12PM - 3:24PM |
PO6.00007: Characterizing heat flux widths in a closed divertor using Surface Eroding Thermocouples Jun Ren, David Donovan, Huiqian Wang, Jon Watkins, Chris Chrobak, Chris Murphy, Dan Thomas, Rejean Boivin The peak heat flux on the divertor target is largely determined by the heat flux width. To characterize the heat flux width in a closed divertor, an array of surface eroding thermocouples (SETCs) were installed in the Small Angle Slot (SAS) divertor in DIII-D. Using the Eich fitting equation, a heat flux width \textasciitilde 2.5mm was derived from experimental heat flux profiles measured by SETC in SAS experiments, similar to the heat flux width predicted by a heuristic drift-based model and the Eich scaling law. The heat flux width in the SAS divertor and on the lower divertor have been compared for the same plasma current in the unfavorable Bt direction. Wider heat flux width was found in Helium plasma than it in Deuterium plasma discharges, possibly due to larger ion gyro-radius of Helium. The upgraded SETCs provide a clearer picture of heat flux in SAS divertor than ever before. The dependence of heat fluxes on divertor closure, plasma current, plasma density, heating power and drift effect will be further investigated in SAS by using SETC. [Preview Abstract] |
Wednesday, October 23, 2019 3:24PM - 3:36PM |
PO6.00008: Impact of drift direction on near-SOL tungsten impurity accumulation in DIII-D Shawn Zamperini, David Donovan, Zeke Unterberg, Peter Stangeby, David Elder, Jonah Duran, Jacob Nichols, Mike Zach Double-sided collector probes inserted into the far-SOL midplane of DIII-D during the 2016 metal rings campaign collected measureable amounts of tungsten (W) along their surfaces. One aim of the campaign was to seek experimental evidence for long-hypothesized, near-SOL impurity accumulation in the upstream crown region caused by the $\nabla $par Ti force on the impurity ions. Upstream accumulation was inferred from the fact that the upstream facing side of the probes measured more W than the downstream facing side, despite the fact that the W-rings are on the downstream side (lower divertor). The collector probe data for two discharges that differed primarily in Bt-direction alone were analyzed in depth, showing that the upstream facing side of the probes measured twice as much W for ion $\nabla $B-drift up vs down. This is explained by drift-dependent fuel-plasma flows measured on a number of tokamaks, including DIII-D. For ion $\nabla $B-drift up, the fuel-plasma is roughly stagnant near the crown, thus ion-ion friction does not compete with the $\nabla $par Ti force favoring accumulation, but for ion $\nabla $B-drift down fast fuel-ion flow towards the inner target can ``flush out'' any accumulating impurities. The near-SOL accumulation understanding gained here is important as any impurity content in this region would act as a boundary value on the core impurity density, potentially creating unacceptable levels of core contamination. [Preview Abstract] |
Wednesday, October 23, 2019 3:36PM - 3:48PM |
PO6.00009: Analysis and model validation of tungsten prompt redeposition in tokamak divertor Jerome Guterl, Tyler Abrams, David Ennis, Curt Johnson, Stuart Loch, Dmitry Rudakov, William Wampler, Philip Snyder Mechanisms governing W prompt redeposition in divertor are analyzed. At the divertor targets, the width of the Chodura sheath is comparable to the sputtered W ionization mean-free path. As a result, W prompt redeposition and net erosion are strongly related to the sheath properties. When 90{\%} or more of sputtered W are ionized within the sheath, the fraction of promptly redeposited tungsten impurity is determined by the multiple ionizations of W impurities outside of the sheath, and weakly depends on other sheath features. Furthermore, we show that the SXB coefficient used to spectroscopically estimate W gross erosion is significantly reduced due to the ionization of W within the sheath when the plasma density is large. The validity of the model describing W prompt redeposition is examined using W net erosion measurements in DIII-D. A reduced model is presented to quantify redeposition and net erosion on W samples exposed to uniform attached divertor plasma conditions. The ratio of W net erosion rates measured experimentally from W samples of different sizes exposed to the same attached plasma conditions [1] are well reproduced with this reduced model. [1] D. L. Rudakov, et al. Physica Scripta~T159 (2014) [Preview Abstract] |
Wednesday, October 23, 2019 3:48PM - 4:00PM |
PO6.00010: Helium and hydrogen interactions with tungsten surfaces as a plasma-facing material Chun-Shang Wong, Robert Kolasinski, Josh Whaley Tungsten is a leading plasma-facing-material candidate for divertors in fusion tokamaks due to its favorable properties, such as a high melting temperature and a low sputter yield. However, plasma-W interactions can complicate matters. Undesirable hydrogen-W interactions include tritium retention and H embrittlement. Helium-W interactions can drive W fuzz growth, leading to material degradation. To investigate these interactions, we performed two experiments for H and He interactions with W surfaces. First, we characterized the W(111)+H(ads) system with multi-angle scattering and recoil maps. Backscattering maps provided crystallographic information of the W(111) substrate, including surface relaxation. Forward-scattering and recoil maps were used with MD simulations to determine H-adsorption positions. Second, we investigated the effect of high-temperature annealing on the surface morphology of He-induced W fuzz. W samples were exposed to a high-flux He plasma under identical conditions to grow fuzz layers. Samples were then annealed for different durations at different temperatures. He ion microscopy and spectroscopic ellipsometry revealed that the W fuzz morphology changed dramatically at temperatures as low as 1273 K. [Preview Abstract] |
Wednesday, October 23, 2019 4:00PM - 4:12PM |
PO6.00011: Particle-particle simulations of plasma-material interactions Jon Drobny, Davide Curreli Plasma-material interactions (PMI) are key to the operation of plasma devices, from DC glow to fusion. Many computational device models rely on assumptions such as perfectly absorbing walls, specular particle reflection, or semi-empirical sputtering yield formulas. To correctly simulate the plasma-material interface, however, a complete PMI model must be developed, remaining as close to first-principles models as possible. To this end, we propose a fully ion-kinetic, particle-in-cell binary-collision-approximation (PIC-BCA) model with one-to-one, single-timestep particle coupling. Each timestep, every PIC superparticle that impacts the wall will generate a BCA superparticle trajectory, and emit any reflected or sputtered particles into the PIC domain, where neutrals can be ionized through a Monte Carlo ionization model. This coupling scheme permits modeling of PMI far from steady-state, without reliance on time-step separation arguments and without the implicit filtering of producing ion energy-angle distributions at the wall. For this presentation, we will present the methodology of the particle-particle coupling and results using pre-existing PIC and BCA codes. [Preview Abstract] |
Wednesday, October 23, 2019 4:12PM - 4:24PM |
PO6.00012: Uncertainty quantification workflows in fusion plasma surface interaction modeling Tiernan Casey, Khachik Sargsyan, Habib Najm Numerical simulations of plasma surface interactions at the device level involve multi-scale processes represented through complex coupled models. These models are defined by parameters that are estimated from typically noisy experimental measurement data, or from higher fidelity simulations that themselves contain multiple parameters that may be physical or related to numerical convergence or tuning. As such the uncertain nature of these parameters and their consequence on the uncertainty of predictions is of interest in order to glean meaningful conclusions with respect to design and operational optimization. We present a variety of uncertain quantification methodologies to enable quantification of parameter uncertainty, model selection, and efficient uncertainty propagation and describe their deployment in the context of a fusion reactor plasma surface interaction modeling hierarchy. [Preview Abstract] |
Wednesday, October 23, 2019 4:24PM - 4:36PM |
PO6.00013: The effects of lithium coating for long pulse and high parameters plasmas in EAST Wei Xu, Jiansheng Hu, Ragesh Maingi, Zhen Sun, Guizhong Zuo, Yuzhong Qian, Xiancai Meng, Chenglong Li, Ming Huang Impurity and fuel recycling are both two key issues for the achievement of long pulse and high stored energy plasmas in tokamak. The EAST has successfully obtained 101s H-mode discharge with low impurity concentration and low wall recycling under long-term lithium coatings. Before the 101s H-mode pulse was achieved, it was found that impurities, e.g. carbon, molybdenum and tungsten, always increased after 40s during long pulse operation, restricting the pulse length due to increasing edge and core radiation. It was observed over the course of the campaign, however, that C, Mo and W impurities gradually decreased with accumulated lithium. In the 101s H-mode discharge, the tungsten core impurity concentration was maintained below 10-5, with the help of strong lithium evaporative coating and real-time lithium powder injection. Moreover, the plasma density and a low wall recycling were maintained during the entire 101s H-mode. And the recycling level even decreased gradually in the discharge later phase due to the increase of Li-II emission, which inducing by the increase of first wall temperature. Furthermore, the long-term lithium coating is found to reduce the HD ratio to 5{\%}, which improved the efficiency of ICRF minority heating. [Preview Abstract] |
Wednesday, October 23, 2019 4:36PM - 4:48PM |
PO6.00014: Experimental development of a lithium vapor box for Magnum-PSI J. A. Schwartz, E. D. Emdee, R. J. Goldston The lithium vapor box divertor is a possible solution for the fusion power exhaust challenge. The extreme but narrow heat flux of a divertor plasma would be dissipated by a localized gas of lithium, which ionizes and radiates to cool the plasma until recombination. Before implementation on a tokamak, the concept will be tested using the linear plasma facility Magnum-PSI, which produces a 3 cm diameter (FWHM) beam of 5 eV plasma that can deliver a continuous 10 MW per meter squared heat flux to a target. In these experiments, the beam will be passed through a cylindrical box filled with lithium vapor at up to 20 Pa and 650 degrees Celsius. As the amount of lithium vapor is varied, the power deposited at the target and profiles of heat flux to the box will be measured. Concepts for the design of the box, including methods of limiting the lithium efflux to the main chamber, details of experimental procedures, and diagnostics are presented. [Preview Abstract] |
Wednesday, October 23, 2019 4:48PM - 5:00PM |
PO6.00015: Predictions For a Simplified Lithium Vapor Box Divertor Using SOLPS-ITER Eric Emdee, Robert Goldston The heat flux impinging on the divertor in future fusion power plants is predicted to be beyond the capabilities of a solid target, attached plasma divertor. Stable detachment, whereby the plasma pressure drops significantly along a magnetic field line as it approaches the divertor target, will thus be necessary. A relevant issue is confining this pressure drop to the divertor region. In past experiments, divertor detachment has typically been followed by strong radiation at the X-point, reducing pedestal performance. The lithium vapor box aims to radiate power via lithium vapor contained within the divertor region. Lithium vapor localization would occur via evaporation near the divertor target and condensation closer to the main chamber. In this way, the detachment of the plasma could be kept stable by a strong dependence of the lithium ionization rate on the poloidal location of the detachment point. New work will be presented on lithium vapor box modeling using SOLPS-ITER. Realistic PFC geometry and EFIT equilibrium for EAST are used. The lithium fraction is shown to be heavily reduced with the addition of a neutral deuterium puff. Effects of deuterium puff location, core density, and wall recycling coefficient on radiated power and impurity concentration are also explored. [Preview Abstract] |
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