Bulletin of the American Physical Society
60th Annual Meeting of the APS Division of Plasma Physics
Volume 63, Number 11
Monday–Friday, November 5–9, 2018; Portland, Oregon
Session PO5: Boundary, Plasma Materials Interactions, Stellarators |
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Chair: Oliver Schmitz, University of Wisconsin Room: OCC B113-114 |
Wednesday, November 7, 2018 2:00PM - 2:12PM |
PO5.00001: The Role of Perturbative Fields in Wendelstein 7-X Samuel Aaron Lazerson, Sergey Bozhenkov, Matthias Otte, Kian Rahbarnia, Marcin Jakubowski, Yu Gao, David Gates The application of perturbative and corrective magnetic fields in Wendelstein 7-X plays a central role in the operation and physics exploitation of the experiment. In the initial limiter campaign, the set of five copper trim coils were utilized to confirm the small magnitude of the n=1 error field (S.A. Lazerson et al. Nuclear Fusion 56 106005, T.S. Pedersen et al. Nature Comm. 7). The first divertor campaign allowed direct assessment of the n/m=1/1 error field through flux surface mapping, allowing for symmetrization of divertor heat loads. Application of symmetrizing fields allowed achievement of pulse lengths in excess of 25 seconds and injected energies of 75 MJ. The trim coil perturbative fields were also used to calibrate and verify the magnetic diagnostics on W7-X. In-vessel control coils were used to modify the n/m=5/5 divertor island chain allowing for divertor strike line actuation and assessment of island width on divertor performance. In this presentation, we review measurements made in previous campaigns and examine the possibility of future scientific exploitation of perturbative fields in W7-X. |
Wednesday, November 7, 2018 2:12PM - 2:24PM |
PO5.00002: Electron cyclotron heating scenarios for long pulse and high performance operation at the stellarator Wendelstein 7‑X Torsten Stange, Heinrich Peter Laqua, Sergey Bozhenkov, Kai Jakob Brunner, Golo Fuchert, Udo Hoefel, Walter Kasparek, Yevgen Kazakov, Stefan Marsen, Dmitry Moseev, Burkhard Plaum, Robert Wolf, Marco Zanini, W7-X Team Electron cyclotron resonance heating (ECRH) with an absorbed power of 7.5 MW provided by 10 gyrotrons is the dominating heating system at the stellarator Wendelstein 7-X. As a very versatile technique, ECRH provides plasma start-up, wall conditioning, heating & electron cyclotron current drive (ECCD). Even though a stellarator is operated without a toroidally driven current, the magnetic island divertor requires “fine tuning” of the edge rotational transform by ECCD at least during transitional phases. A not vanishing bootstrap current can be compensated or mimicked during plasma ramp-up to prohibit an unwanted sweep over the divertor. In preparation for long pulse and high performance operation with an actively cooled divertor beyond 2020, the necessary ECRH scenarios were already demonstrated in the operational phase 1.2 with an uncooled divertor & an energy limit of about 80 MJ per pulse. Up to densities of 1020m-3 a standard X2-heating scenario is used which has to be optimized with regard to finite beta & Shafranov-shift. For densities close and beyond the X2-cutoff density, an advanced O2-heating scenario was developed which makes use of holographic reflection tiles to allow 3 beam paths. Central densities of 1.4∙1020 m-3 were achieved at a central temperature of 3 keV. |
Wednesday, November 7, 2018 2:24PM - 2:36PM |
PO5.00003: Overview of first NBI fast ion results from W7-X Simppa Äkäslompolo, Sergey Bozhenkov, Yevgen Kazakov, Samuel A Lazerson, Dirk Hartmann, Benedikt Geiger, Glen A Wurden, Oliver Ford, Kunihiro Ogawa, Mitsutaka Isobe, Christoph Slaby, Kian Rahbarnia, Henning Thomsen, Dmitry Moseev, Robert Wolf, W7-X Team We present an overview of first fast ion experiments in the optimized stellarator Wendelstein 7-X (W7-X). The improved fast-ion behaviour is essential for applicability of stellarators as a fusion reactor and is one of the optimization criteria for the W7-X field geometry. The neutral beam injection (NBI) system of the Wendelstein 7-X (W7-X) stellarator is commissioned during the 2018 experimental campaign. The NBI system consists of two boxes with two sources each. The four sources inject 55 keV hydrogen with approximately 1.7MW power, each. The injection geometry is quite radial and the two boxes produce stellarator-antisymmetric initial fast ion populations in phase space. The confined fast ion population is measured with fast ion hydrogen alpha spectroscopy (FIDA) and ion cyclotron emissions probes (ICE) as well as via the excited MHD activity using magnetic pick-up coils. The lost fast ions are measured with a fast ion loss diagnostics probe (FILD) as well as monitored with infrared and visible light cameras viewing the plasma facing components. |
Wednesday, November 7, 2018 2:36PM - 2:48PM |
PO5.00004: Stellarator coil optimization towards higher engineering tolerances Jim-Felix Lobsien, Michael Drevlak, Thomas Sunn Pedersen Recently designed optimized stellarator experiments have suffered from very tight construction tolerances, but some level of deviation of the coil system is unavoidable during fabrication of the coils and assembly of the coil system. We present a new approach that incorporates reduced sensitivity to construction tolerances of the coil system into the optimization sequence. The approach was tested within the framework of the existing coil optimization scheme for Wendelstein 7-X. The results are compared with those of a coil set obtained by the original optimization. The result is a more optimal coil system with less stringent tolerances, such that small deviations cause reduced deterioration of the properties important for fusion performance. |
Wednesday, November 7, 2018 2:48PM - 3:00PM |
PO5.00005: Sensitivity of Energetic Particle Transport in 3D Toroidal Magnetic Confinement Configurations to Changes in Boundary Shape Thomas Antonsen, Elizabeth J Paul, Matt Landreman Cross-field transport of energetic particles in 3D magnetic field configurations due to magnetic drifts depends sensitively on the spatial dependence of the confining magnetic field. In optimizing a configuration, it is desirable to know how changes in the shape of the outermost flux surface will impact the transport. We present here a method for determining this sensitivity for arbitrary displacements of the outer flux surface based on adjoint methods. The method compares the value of a transport metric for two neighboring MHD equilibria that differ by a small displacement of the magnetic field, but have the same flux dependent pressure profile and rotational transform. The method involves a single solution of the linearized MHD equations for fixed boundary shape but subject to the addition of a judiciously chosen bulk force. The method can be implemented in a standard MHD stability code. Examples will be presented. |
Wednesday, November 7, 2018 3:00PM - 3:12PM |
PO5.00006: Plasma density blobs and drift wave dynamics with shear flow Yanzeng Zhang, Sergei Krasheninnikov, Haotian Mao, Alexey Knyazev Blobs, propagating in tokamak edge toward outer wall, play an important role in the scrape-off layer plasma transport. Although the blobs are studied for about 15 years, the mechanism(-s) of blob formation is still under debates. Recently, the Hasegawa-Mima equation was generalized by considering Boltzmann electrons and keeping all nonlinearities. It follows that the amplitude of wave-packet propagating in the direction of decreasing background plasma density will increase exponentially with the distance travelled until nonlinear effects become important. Nonlinear modification of time averaged plasma density profile results in the formation of large amplitude modes locked in x-direction but still propagating in y-direction, which resembles experimentally observed chain of blobs propagating in poloidal direction. Such specific density profiles, causing the locking of drift waves can form naturally at the edge of tokamak due to neutral ionization source. The impact of the shear flow on the drift wave dynamics is studied to be a factor impeding the development of the large density structure. |
Wednesday, November 7, 2018 3:12PM - 3:24PM |
PO5.00007: The divertor tokamak test facility project Raffaele Albanese, Flavio Crisanti, Piero Martin, ALdo Pizzuto The Divertor Tokamak Test Facility (DTT) is a new tokamak whose construction has recently been approved by the Italian government. DTT will be a high field superconducting toroidal device (6 T) carrying plasma current up to 6 MA in pulses with length up to 100s, with an up-down symmetrical D-shape defined by major radius R=2.15 m, minor radius a=0.7 m and average triangularity 0.3. The main role of DTT is to contribute to the development of a reliable solution for the power and particle exhaust in a reactor, a challenge commonly recognised as one of the major issues in the road map towards the realisation of a nuclear fusion power plant. This paper will illustrate the main features of the new DTT experiment. |
Wednesday, November 7, 2018 3:24PM - 3:36PM |
PO5.00008: A Simplified Lithium Vapor Box Divertor Design Robert J. Goldston, Eric D. Emdee, Michael A. Jaworski, Marvin E. Rensink, Thomas D. Rognlien UEDGE simulations give the remarkable result that a divertor plasma acts nearly like a mirror for lithium vapor. In lithium-detached plasmas the rates of lithium ionization and recombination at the cold ends of each flux tube are nearly in local balance. This suggests the possibility of simplifying the lithium vapor box divertor concept by using the plasma itself as one of its side walls, and perhaps forgoing distinct baffles. Since there is much less heat flux on the private-flux side of a divertor leg, it should be easier to detach the divertor leg all the way to the separatrix from that side, so we choose to evaporate lithium into the private-flux side of the divertor leg. This creates a vapor-box-like configuration where the lithium vapor is emitted from near the target end of the divertor leg, on the private flux side, and is recondensed along the “dome” closer to the main plasma as well as on the outer wall of the divertor channel. Configurations of this sort, without material baffles, may be easier to test on existing experiments and may be geometrically attractive for future devices. We will examine engineering parameters, including lithium liquid flow rates and cooling rates, for practical examples. |
Wednesday, November 7, 2018 3:36PM - 3:48PM |
PO5.00009: Advanced Divertor Studies in EAST/DIII-D by SOLPS Hang Si, Houyang Guo Divertor is one of the key components in Tokamak. The design, construction and operation of advanced divertors have been the main topics of tokamak research during the last decade. In general, developing an advanced divertor configuration requires: (1)Optimizing magnetic configuration to extend the plasma-wetted area through flux expansion, and increasing the divertor volume by increasing the field-line length;(2)Increasing divertor closure by divertor baffling to improve divertor screening for recycling neutrals and impurities, hence increasing divertor neutral pressure, thus enhancing divertor particle and power exhaust. In the present work, some advanced divertor solutions are developed and validated in EAST and DIII-D respectively. In EAST an alternative advanced divertor configuration, i.e., quasi snowflake (QSF), aka X-divertor and in DIII-D an alternative advanced divertor coupling divertor closure with advanced magnetic configuration(X-divertor) have been attempted respectively. The findings indicate that the advanced divertor solutions in EAST and DIII-D can significantly reduce the peak heat flux density at the divertor target and the density threshold for detachment, which provides a promising means for the design of advanced divertors in the next-step fusion devices. |
Wednesday, November 7, 2018 3:48PM - 4:00PM |
PO5.00010: First simulations of dissipative divertor conditions with resonant magnetic perturbations evaluated for the non-active phase in ITER Heinke Frerichs, Xavier P Bonnin, Yühe Feng, Yueqiang Liu, Alberto Loarte, Richard A Pitts, Detlev Reiter, Oliver Schmitz The application of external, symmetry breaking resonant magnetic perturbations (RMPs) is required for control of edge localized modes in ITER, yet its divertor has been designed under axisymmetric assumptions. Non-axisymmetric divertor conditions arise under RMP application where the perturbed separatrix guides field lines from the bulk plasma onto divertor plates, and are here modeled with the 3D edge plasma and neutral gas code EMC3-EIRENE. A linearization of the electron energy loss term from interactions with neutral gas has been implemented, providing numerical stable access to dissipative divertor conditions. Comparisons with SOLPS-ITER simulations for hydrogenic plasmas without impurities during the initial non-active low power phase show good agreement for a roll-over of the peak particle flux from increased gas-puffing. Subsequent simulations for RMP application are sensitive to assumptions on rotation in plasma response calculations by MARS-F, and show that detachment at the original strike zone can occur at lower gas-puff rates and lower peak particle flux distinct from the behavior at the non-axisymmetric strike point further away. |
Wednesday, November 7, 2018 4:00PM - 4:12PM |
PO5.00011: Study of edge transport near density limit with BOUT++ simulations* Ben Zhu, Xueqiao Xu As a first step to understand edge plasma behaviors, especially near the density limit before disruptions, a series of BOUT++ simulations are carried out to study the edge turbulence and transport prior disruptions. We start with a simulation of a DIII-D lower single-null density-limited shot right before density ramping. Strongly inboard-outboard asymmetric fluctuation indicates that the dominate turbulence driver is the resistive ballooning mode. As the density gradually increases turbulence and radial transport are dramatically enhanced likely due to the increasing resistivity. The broadened heat flux width on outboard mid-plane are observed in our simulations. We also find that radial transport starts to dominate parallel transport near the density limit which could potentially retract edge plasma profile, induce detachment and x-point MARFE formation, and eventually lead to disruption. |
Wednesday, November 7, 2018 4:12PM - 4:24PM |
PO5.00012: Simulation of the Lithium pellet injection on EAST tokamak using the BOUT++ transport code Yumin Wang, X.Q. Xu, Zhehui Wang, Nami Li, Zhen Sun, Tianyang Xia, Jiansheng Hu, Xiang Gao A novel model predicting the Lithium pellet injected into the plasmas has been developed to simulate the pellet effects to the background plasma profiles, as well as the Lithium impurity density and temperature profiles. The pellet ablates due to the heat from the background plasma, and the neutral gas shielding (NGS) model has been used to describe this physical process. The shielding factor of the model have been identified by comparing the simulated neutral gas density and the experimental neutral gas measurements using CCD camera. In the new BOUT++ transport code, the ionization, charge exchange and recombination rates of the lithium ions use the data from the ADAS database. The transport equations have been obtained by reduction of the Braginskii equations with source and sink terms due to the collisional interactions. The scan of the pellet parameters such as the pellet size, injection velocity has been conducted. The comparison of the simulation results and the experimental measurements from both the reflectometry and the POINT interferometer have been done on EAST tokamak. |
Wednesday, November 7, 2018 4:24PM - 4:36PM |
PO5.00013: Modeling of nano-meter scale dust grains in tokamak Zhuang Liu, Xiaotao Xiao, Nami Li, Xueqiao Xu, Dezhen Wang In this work, the NDS (Nano-meter Dust Simulation), which evaluates the charging and ablation processes of the dust grains, has been done under BOUT++ framework. The guiding-center orbits of dust particles are tracked in tokamak plasmas, whose parameters are obtained from BOUT++, a highly desirable C++ code package for performing parallel plasma fluid simulations with an arbitrary number of equations in 3D curvilinear coordinates. The Hamiltonian guiding center equation of motion is utilized to describe the velocity and acceleration of dust grains at arbitrary time step. Calculations with the coupled NDS-BOUT++ codes provide results such as trajectories and distributions of dust particles with different components, sizes, and velocities for different tokamak geometries. The distribution of tungsten impurity, resulted from ablated tungsten dust grains for several typical scenarios, is also assessed. Preliminary results show in some cases, tungsten dust grains can cross the seperatrix and survive for several ms before ablated completely, which will significantly contribute to core contamination. |
Wednesday, November 7, 2018 4:36PM - 4:48PM |
PO5.00014: A Study on Effective Parameters of Radial Electric Field in IR-T1 Tokamak Pejman Khorshid, Kowthar Noori The radial electric field is calculated from the radial component of the force balance equation. It has been shown that three major driving forces have contributed radial electric field including radial pressure gradient, poloidal and toroidal rotation. In this study, we have considered the contribution of these terms, thereby calculating the profiles of plasma parameters by using fluids equations especially in the edge of plasma. Then we have obtained profiles of terms and compared them. The results have shown that the gradient of pressure, which is a negative term, makes a small contribution to the radial electric field; on the other hand, the rotational term more specifically, poloidal one makes a great contribution to it. Therefore, it seems true to say that, the sign of the field is dependent on the poloidal term. |
Wednesday, November 7, 2018 4:48PM - 5:00PM |
PO5.00015: Predicting Equilibrium Electron Density Production on Proto-MPEX Pawel Andrzej Piotrowicz, Juan F Caneses, Nischal Kafle, Richard H Goulding, David L Green, John B Caughman, Cornwall Hong Man Lau, David Neil Ruzic The helicon antenna plasma source installed on Proto-MPEX has been able to produce electron densities of ne >6×1019 m-3 in the “helicon-mode”. The existing power availability of the helicon antenna is P≈100 kW at 13.56 MHz driving frequency. The “helicon-mode” plasma on Proto-MPEX is only observed up to a magnetic field of BH < 0.08 T. This low operating magnetic field strength restricts the plasma diameter at the target, which is typically operated at B0 ≈ 1.0 T, to dp ≈ 4 cm. A larger plasma diameter would be beneficial to Proto-MPEX both for electron and ion heating demonstration as well as future PMI experiments. A full-wave electromagnetic model that was used to explain the mechanisms that lead to the observation of the “helicon-mode” plasma on Proto-MPEX is used to predict power deposition in the helicon region. Power losses are calculated and used in a power balance to predict an equilibrium electron density. Comparison with experimentally measured electron density variation with magnetic field strength at a power level of P = 100 kW is presented. This model is then used to predict the expected electron density at higher magnetic field values with an available power of P = 200 kW. |
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