Bulletin of the American Physical Society
58th Annual Meeting of the APS Division of Plasma Physics
Volume 61, Number 18
Monday–Friday, October 31–November 4 2016; San Jose, California
Session JI3: Edge PhysicsInvited
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Chair: Jeremy Lore, Oak Ridge National Laboratory Room: 210 ABEF |
Tuesday, November 1, 2016 2:00PM - 2:30PM |
JI3.00001: Control of neutral particle fueling and exhaust by plasma edge topology optimization in Wendelstein 7-X and HSX Invited Speaker: Laurie Stephey Comparative experiments at the HSX and Wendelstein 7-X stellarators are being performed. At W7-X it was shown that fine control of the edge magnetic structure in W7-X is a feasible actuator to control global particle confinement. During the startup campaign of W7-X, the edge magnetic structure is defined by five poloidal limiters. Inside of the last closed flux surface in the standard magnetic configuration, the 5/6 resonance and corresponding magnetic island are located directly inside of the plasma source region. Inward movement of the island in a predominantly electron-root transport regime has been found to increase the effective helium confinement time $\tau_p^*$ He, a critical metric for plasma purity control in future burning plasmas, by a factor of two. The experimental analysis is supported by fully 3-D fluid plasma and kinetic neutral modeling using the EMC3-EIRENE code and will be compared to these experimental results from both devices. A single reservoir, single species particle balance will be extracted from experimental measurements aided by the fully 3-D modeling analysis from EMC3-EIRENE to quantify the causal link established above based on measured parameters. At HSX, similar investigations to those performed at W7-X are ongoing. HSX has substantial flexibility in both its edge magnetic configuration and also in edge connection lengths via limiter insertion. Both are being examined to study any resulting changes global particle confinement and provide insight into the physics of the underlying mechanism. Together with the results from W7-X, both experiments will provide information on the link between the plasma edge topology and the global particle confinement. [Preview Abstract] |
Tuesday, November 1, 2016 2:30PM - 3:00PM |
JI3.00002: The role of atomic and molecular physics for dissipative divertor operation in helium and deuterium plasmas Invited Speaker: J.M. Canik Recent experiments in DIII-D helium plasmas are examined to resolve the role of atomic and molecular physics in major discrepancies between experiment and modeling of dissipative divertor operation. Helium operation removes the complicated molecular processes of deuterium plasmas that are a prime candidate for the inability of standard fluid models (SOLPS, UEDGE) to reproduce dissipative divertor operation, primarily the consistent under-prediction of radiated power. With helium fueling, a high-recycling divertor was established with divertor densities increasing to n$_{e,div} \geq 3\times10$^{20} m^{-3}$ and temperatures decreasing to T$_{e,div} \leq$ 2 eV as measured by divertor Thomson scattering (DTS). The electron pressure, $p_{e,div}$ decreased gradually with increasing density to less than 30$\%$ of the low density value. However, the ion flux to the divertor target did not decrease until the highest densities and lowest temperatures, T$_{e,div} \leq$ 2 eV. In contrast, with deuterium operation, increasing density leads to a rapid transition from T$_{e,div} \geq$ 10 eV to T$_{e,div} \leq$ 3 eV, though both $p_{e,div}$ and ion flux do not decrease until T$_{e,div} \leq$ 2 eV. These differences indicate an important role for molecular and atomic physics in the dynamics of divertor dissipation. Initial SOLPS modeling has reproduced $n_e$ and $T_e$ profiles at the midplane and divertor target, as well as the spatial structure of radiation patterns measured in moderate density helium plasmas. However, the modeled divertor radiation is less than measured, similar to deuterium simulations, suggesting processes more universal than species-specific atomic or molecular physics may be the source of radiation deficit. Detailed assessments of $n_e$, $T_e$ profiles in the divertor volume, uniquely determined at DIII-D using DTS, are made along with analysis of measured and modeled line radiation to shed more light on these intriguing findings. [Preview Abstract] |
Tuesday, November 1, 2016 3:00PM - 3:30PM |
JI3.00003: Attainment of a stable, fully detached plasma state in innovative divertor configurations. Invited Speaker: Maxim Umansky The heat load on plasma facing components is a critical engineering constraint for future tokamaks, which has stimulated the community to consider innovative magnetic divertor geometries for future high power devices. Present-day advanced divertor scenarios generally rely on partially detached regimes, also planned for ITER; a fully detached state would usually lead to MARFE and degradation of core confinement [1]. Modeling reveals that novel magnetic geometries can have a major impact on plasma detachment and power handling. Using the UEDGE tokamak edge transport model for configurations with tightly baffled long divertor legs, extended radially [2], or vertically, we find stable, fully detached divertor operation. Including a secondary X-point in the outer leg volume [3] extends the attainment of a stable detached state to the highest power. As the input power is reduced to a threshold value, the outer leg transitions to a fully detached state with the detachment front localized at the secondary X-point or in the leg volume; reducing the power further results in the detachment front steady-state location shifting upstream. As the power is reduced, the detachment front eventually moves to the primary X-point, which sets the lower power limit for the range of stable operation. Still, for a long-legged divertor, a fully detached, stable divertor regime is maintained over an order-of-magnitude variation in exhaust power. In contrast, a standard divertor has a much smaller detachment operational window. These results suggest that stable fully detached divertor operation can be realized in tokamaks with extended divertor legs. [1]Matthews, J. Nucl. Mater. 220-222 (1995) 104. [2]Valanju et al., Phys. Plasmas 16, 056110, 2009. [3] LaBombard et al., Nucl. Fusion 55, 053020, 2015. [Preview Abstract] |
Tuesday, November 1, 2016 3:30PM - 4:00PM |
JI3.00004: A fluid modeling perspective on the tokamak power scrape-off width using SOLPS-ITER Invited Speaker: Eric Meier SOLPS-ITER, a 2D fluid code, is used to conduct the first fluid modeling study of the physics behind the power scrape-off width ($\lambda_q$). When drift physics are activated in the code, $\lambda_q$ is insensitive to changes in toroidal magnetic field ($B_t$), as predicted by the 0D heuristic drift (HD) model developed by Goldston. Using the HD model, which quantitatively agrees with regression analysis of a multi-tokamak database, $\lambda_q$ in ITER is projected to be 1 mm instead of the previously assumed 4 mm, magnifying the challenge of maintaining the peak divertor target heat flux below the technological limit. These simulations, which use DIII-D H-mode experimental conditions as input, and reproduce the observed high-recycling, attached outer target plasma, allow insights into the scrape-off layer (SOL) physics that set $\lambda_q$. Independence of $\lambda_q$ with respect to $B_t$ suggests that SOLPS-ITER captures basic HD physics: the effect of $B_t$ on the particle dwell time ($\sim$$B_t$) cancels with the effect on drift speed ($\sim$$1/B_t$), fixing the SOL plasma density width, and dictating $\lambda_q$. Scaling with plasma current ($I_p$), however, is much weaker than the roughly $1/I_p$ dependence predicted by the HD model. Simulated net cross-separatrix particle flux due to magnetic drifts exceeds the anomalous particle transport, and a Pfirsch-Schluter-like SOL flow pattern is established. Up-down ion pressure asymmetry enables the net magnetic drift flux. Drifts establish in-out temperature asymmetry, and an associated thermoelectric current carries significant heat flux to the outer target. The density fall-off length in the SOL is similar to the electron temperature fall-off length, as observed experimentally. Finally, opportunities and challenges foreseen in ongoing work to extrapolate SOLPS-ITER and the HD model to ITER and future machines will be discussed. [Preview Abstract] |
Tuesday, November 1, 2016 4:00PM - 4:30PM |
JI3.00005: Compatibility of lithium plasma-facing surfaces with high edge temperatures in the Lithium Tokamak Experiment (LTX) Invited Speaker: Dick Majeski High edge electron temperatures (200 eV or greater) have been measured at the wall-limited plasma boundary in the Lithium Tokamak eXperiment (LTX). High edge temperatures, with flat electron temperature profiles, are a long-predicted consequence of low recycling boundary conditions. The temperature profile in LTX, measured by Thomson scattering, varies by as little as 10{\%} from the plasma axis to the boundary, determined by the lithium-coated high field-side wall. The hydrogen plasma density in the outer scrape-off layer is very low, 2-3 x 10$^{\mathrm{17\thinspace }}$m$^{\mathrm{-3}}$, consistent with a low recycling metallic lithium boundary. The plasma surface interaction in LTX is characterized by a low flux of high energy protons to the lithium PFC, with an estimated Debye sheath potential approaching 1 kV. Plasma-material interactions in LTX are consequently in a novel regime, where the impacting proton energy exceeds the peak in the sputtering yield for the lithium wall. In this regime, further increases in the edge temperature will decrease, rather than increase, the sputtering yield. Despite the high edge temperature, the core impurity content is low. Z$_{\mathrm{eff}}$ is 1.2 -- 1.5, with a very modest contribution (\textless 0.1) from lithium. So far experiments are transient. Gas puffing is used to increase the plasma density. After gas injection stops, the discharge density is allowed to drop, and the edge is pumped by the low recycling lithium wall. An upgrade to LTX which includes a 35A, 20 kV neutral beam injector to provide core fueling to maintain constant density, as well as auxiliary heating, is underway. Two beam systems have been loaned to LTX by Tri Alpha Energy. Additional results from LTX, as well as progress on the upgrade -- LTX-$\beta $ -- will be discussed. [Preview Abstract] |
Tuesday, November 1, 2016 4:30PM - 5:00PM |
JI3.00006: Global 3D Two-Fluid Simulations of Turbulent Transport in the Tokamak Edge Region: Turbulence, Profile Evaluation and Spontaneous E $\times$ B rotation Invited Speaker: Ben Zhu A new global 3D two-fluid code, GDBM, based on the drift-reduced Braginskii model[1] is developed to study the turbulent transport across the entire tokamak edge region: from plasma sources in the inner core to plasma sinks in the outer-most scrape-off layer (SOL). In this code, profiles of plasma density, electron and ion temperature, electric potential, magnetic flux and parallel flow are evolved self-consistentlty. As a first step, the current code implements a flux-driven source zone near the core and a limiter configuration with the volume penalization method on the edge. In the L-mode regime, simulations indicates the predominant driver of edge turbulence is the resistive ballooning instability. The simulations show that, in agreement with experimental observations, as the simulation moves towards density limit regime by ramping up the density profile, the turbulent transport is enhanced; on the other hand, as the simulation approaches to the H-mode regime by ramping up the temperature profile, the turbulent transport is suppressed. In all cases, spontaneous formation of the E $\times$ B drift in the opposite direction of the ion diamagnetic drift is observed. The E $\times$ B shear gets stronger as the temperature is increased. These findings seem largely consistent with previous local flux-tube simulation results[2]. Simulations also show that in L-mode plasma, the Boussinesq approximation has limited impact on the turbulence structure, fluctuation level and the global profile evolution. More detailed results will be presented at the meeting. \\ $[1]$ A. Zeiler, J. F. Drake and B. N. Rogers, Phys. Plasmas, 4, 2134 (1997) \\ $[2]$ B. N. Rogers, J. F. Drake and A. Zeiler, Phys. Rev. Lett., 81,4396 (1998) [Preview Abstract] |
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