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 JI2: Boundary, Plasma Material Interactions, RF Heating |
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Chair: Dennis Whyte, Massachusetts Institute of Technology Room: OCC Oregon Ballroom 203 |
Tuesday, November 6, 2018 2:00PM - 2:30PM |
JI2.00001: An improved understanding of the roles of atomic processes and power balance in target ion current loss during detachment Invited Speaker: Kevin Verhaegh The process of divertor detachment, whereby heat and particle fluxes to divertor surfaces are strongly reduced, is required to reduce heat loading and erosion in a magnetic fusion reactor. In this talk the physics leading to the decrease of the TCV tokamak divertor ion flux, or ‘roll-over’ (corresponding to partial detachment), is experimentally explored through characterisation of the location, magnitude and role of the various divertor ion sinks and sources including a complete measure of particle and power balance. This is achieved using a novel, validated, experimental probabilistic analysis of the Balmer line spectra from multi-chord visible spectroscopy. Over a range in core plasma conditions (current, impurity-seeding, density) the divertor target ion current (It) loss, measured by Langmuir probes, is caused by a drop in the measured divertor ion source; volumetric recombination remains small to negligible. ‘Power limitation’ results in ion source reduction when 50% or more of the power entering the recycling region (Precl) is spent on ionisation, in agreement with analytical predictions. This pivot point corresponded to target temperatures ~ 4-6 eV and to a critical upstream pressure to recycling heat flux, estimated using IR heat deposition profiles and Precl, ratio (pu/qrecl): both in agreement with analytical predictions. The profiles of the various ion sources and sinks together with power losses are tracked experimentally as detachment proceeds and were quantitatively consistent with SOLPS modelling. As the power reaching the near-target region decreases, the ionisation and impurity radiation regions move towards the x-point, leaving behind a region with strong molecular (Dα) emission and elevated charge exchange to ionisation ratios. From those ratios momentum losses (up to 70%) are implied that develop concurrently with power limitation of the ion source. |
Tuesday, November 6, 2018 2:30PM - 3:00PM |
JI2.00002: Validating Divertor Power Exhaust Models with Vacuum Ultraviolet Spectroscopy in DIII-D Invited Speaker: Adam Mclean The unique combination of vacuum ultraviolet spectroscopy, impurity flow imaging, and Thomson scattering in the DIII-D divertor has enabled direct measurement of the components of radiated power, significantly advancing our understanding of dissipative processes and the root causes of an under-prediction of radiative power by fluid simulations compared to experiments. Radiative exhaust experiments in H-mode at DIII-D, corroborated with multi-fluid simulations, show that the intrinsic carbon impurity, CIV (1550 Å), line dominates the radiated power and peaks close to the X-point in detached divertor conditions with ion B×∇B drift toward the X-point (fwd. BT). In contrast operating with ion drift away from the X-point (rev. BT), the Deuterium Ly-a (1215 Å) dominates over CIV in detached conditions and peaks in front of the outer target revealing the important role of cross-field drifts for detachment onset and detachment front characteristics. UEDGE simulations with drifts qualitatively capture the dominant radiating lines in fwd. and rev. BT. The simulations, however, both under-predict the total radiated power and predict the profile to be a factor of three more localized than measured, indicating models are not capturing mechanisms that expand the radiating volume. Cross-field drifts and parallel flows measured in 2D with coherence imaging are examined as possible root causes for the radiation shortfall. Divertor exhaust predictions for ITER and design efforts for future fusion devices rely on simulations with 2D fluid codes, such as SOLPS. This work provides a significant contribution in the effort to improve confidence of the code predictions. |
Tuesday, November 6, 2018 3:00PM - 3:30PM |
JI2.00003: Application of a Kinetic-Analytic Model for Edge Localized Modes to Fast Tungsten Sputtering Measurements in the DIII-D Divertor Invited Speaker: Tyler Abrams The roles of deuterium and low-Z impurities in tungsten sourcing during ELMs in an ITER-like mixed materials environment have been quantified in the DIII-D divertor with high temporal and spatial resolution. These findings are consistent with the Fundamenski-Moulton 'free-streaming' (FMFS) model predictions of how the W source scales with ELM deposited energy density. The modified FMFS model utilizes plasma conditions at the pedestal top as input and accounts for enhanced target electron densities and ion fluxes due to neutral recycling during ELMs to calculate ion impact energies and flux densities on the divertor targets. This model shows that the energetic 'free-streaming' D+ and C6+ ions originating inside the confined plasma volume dominate W sourcing during ELMs, despite comprising a relatively small fraction of the total ELM ion flux, because the high impact energy of these ions causes substantial physical sputtering of tungsten. Quantitative agreement across a range of ELM frequencies and sizes is observed between spectroscopic measurements of intra-ELM gross erosion of tungsten and this model when coupled to SDTrim.SP sputtering calculations. Interpretive modeling for the spatial profile of tungsten erosion during and between ELMs was also developed via OEDGE. It includes full mixed-material effects, W self-sputtering and both carbon and main ion sputtering. In contrast to the JET-ILW environment, the inter-ELM phase dominates the time-averaged W divertor source due to the more efficient physical sputtering process of C on W, relative to Be on W, in the sub-keV ion impact energy regime. This work represents major progress towards a predictive model linking pedestal conditions to the ELM-induced high-Z divertor source, essential for ITER and beyond where W sputtering due to ELMs may dominate the tungsten divertor source and overall high-Z core impurity accumulation. |
Tuesday, November 6, 2018 3:30PM - 4:00PM |
JI2.00004: Modeling of plasma-wall interaction in tokamak experiments with high-Z materials Invited Speaker: Rui Ding Since high-Z material W will be used for ITER divertor, the understanding of related plasma-wall interaction processes is indispensable and only can be achieved in a combined effort of experiment and modelling. The 3D Monte Carlo code ERO taking into account a material mixing surface model has been used to simulate W erosion and re-deposition on DIII-D with toroidally continuous W rings embedded in the divertor and EAST with an upper full W divertor. Modeling shows that the transport of C impurities not only dominates the W sputtering but also determines the overall erosion and deposition balance in the mixed materials surface. With a self-consistent calculation of C impurity distribution, W gross erosion rates measured by WI spectroscopy can be well reproduced by the modeling. The ExB drift and lower electron temperature at the radial outboard side lead to a net deposition zone where W and C are accumulated. In the net erosion zone closer to the outer strike point, the W coverage on C is very low and saturated independent of exposure time, agreeing with the measurements by collector probes. Strong sheath effects on material erosion rates have been observed using external biasing samples, which have been simulated by the 2D PIC code SPICE2 and the ERO code. Both the PIC simulation and the measured Dα emission reveal that with increasing biasing voltage the ion flux decreases at the biased area while increases at the adjacent downstream tile, although the biased sample potential is far below the plasma potential. The reason indicated by modeling is the strong gradient of the electric field in the sheath, which results in different magnitude of the polarization drift above the biased and non-biased surface. More than an order of magnitude reduction in erosion with slight positive voltage biasing in the experiments is due to the reduced incident energy and ion flux. |
Tuesday, November 6, 2018 4:00PM - 4:30PM |
JI2.00005: Electron and ion heating physics for the Material Plasma Exposure eXperiment plasma source concept Invited Speaker: Cornwall Lau The Material Plasma Exposure eXperiment (MPEX) is a proposed high-intensity linear divertor simulator based on radio frequency (RF) physics and technology to address the fusion plasma-material interaction difficulty. RF heating has never been demonstrated at these unexplored high density, low temperature, and high collisionality regime in linear plasma machines, so new RF heating schemes, challenges, results, and understanding have emerged. We report recent electron and ion heating physics highlights on the Proto-MPEX experiment as a demonstration of various heating concepts for MPEX:
We have observed and modeled these results for a variety of different heating schemes: magnetic beach ion cyclotron heating, electron Bernstein waves (EBW), upper hybrid (UH), and electron cyclotron heating (ECH). For observed core ion cyclotron heating, finite electron temperature needs to be included in the model to match experimental observations of core ion temperature. The model also illustrates possible electron heating mechanisms at the Alfven resonance. EBW modeling and experiments show that beam focusing and reduction in magnetic mirror trapping are necessary for efficient core heating. UH collisional damping can be controlled both experimentally and numerically by magnetic field or helicon power to obtain core or edge power deposition and may reduce effects of magnetic mirror trapping. ECH may have low single pass absorption, and both UH and ECH require higher frequencies to access higher density plasmas. Neutral pressure control is important for all these electron heating schemes. |
Tuesday, November 6, 2018 4:30PM - 5:00PM |
JI2.00006: First demonstration of efficient lower hybrid current drive at reactor relevant densities in a tokamak Invited Speaker: Seung-Gyou Baek For the first time, lower hybrid current drive (LHCD) is demonstrated in a reactor relevant diverted configuration at densities exceeding the previously reported “LH density limit” at the line-averaged density of ne ≈ 1.0x1020 m-3. By achieving efficient LHCD beyond the LH density limit (above which anomalous loss of current drive efficiency is usually observed), these new experimental results show that current drive at a reactor relevant density can be attained by management of the scrape-off-layer (SOL) plasma at high Bt and Ip. In Alcator C-Mod, detailed studies have identified parasitic wave interactions with the diverted SOL plasma as a dominant contributor to the LH density limit. To avoid this, the new experiments focused on minimizing the SOL width and its associated turbulence level, based on the previously identified scaling of the SOL width with the Greenwald fraction (ne/nG, where nG ≡ Ip/(πa2)) [B. LaBombard, et al., Phys. Plasmas 15 (2008) 056106]. Operation at high Bt (7.8 T) allows increased plasma current up to 1.4 MA at ne ≈ 1.4x1020 m-3. At this low Greenwald fraction (ne/nG < 0.2), parasitic interactions are largely eliminated, evidenced by minimal broadening in the measured wave frequency spectrum. LHCD efficiency in these high density plasmas is found to match that achieved at low density: η0=neIpR0/PLH ≈ 2.5 (1019 MA MW-1 m-2). Further evidence of efficient current drive comes from increased non-thermal Bremsstrahlung emission rates at high Ip, whereas the emission remains low at low Ip. These new experimental results provide further motivation to place the LH antenna at the high-field-side (HFS) of the tokamak in a double null configuration. In this case, the HFS SOL exhibits a narrow SOL width and low levels of SOL turbulence, which is expected to provide an optimum SOL condition in attaining efficient current drive in a reactor plasma. |
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