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Stangeby, Peter C.

Publications and source records attributed to Stangeby, Peter C..

10 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Design and physics basis for the upcoming DIII-D SAS-VW campaign to quantify tungsten leakage and transport in a new slot divertor geometry

A set of experiments are planned to exploit the high SOL collisionality enabled by a tightly baffled slot divertor geometry to suppress tungsten leakage in DIII-D. A toroidal row of graphite tiles from the Small Angle Slot (SAS) divertor is being coated with 10–15 μ m of tungsten. New spectroscopic viewing chords with in-vacuo optics will measure the W gross erosion source from the divertor surface with high spatial and temporal resolution. In parallel, the bottom of the SAS divertor is changed from a flat to a ‘V’ shape. New SOLPS-ITER/DIVIMP simulations conducted with drifts using the planned ‘V’ shape predict a substantial reduction in W sourcing and SOL accumulation in either B × ∇B direction relative to either the old SAS divertor shape or the open, lower divertor. Dedicated studies are planned to carefully characterize the level of W sourcing, leakage, and scrape-off-layer (SOL) accumulation in DIII-D over a wide range of plasma scenarios. Various actuators will be assessed for their efficacy in further reducing high-Z impurity sources and leakage from the slot divertor geometry. Finally, this coupled code-experiment validation effort will be used to stress-test physics models and build confidence in extrapolations to advanced, high-Z divertor geometries for next-step devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

First evidence of dominant influence of E × B drifts on plasma cooling in an advanced slot divertor for tokamak power exhaust

Addressing power exhaust in tokamaks is presently recognized as one of the major remaining open issues for the development of fusion reactors. At the forefront of this endeavor is the effort to develop an advanced divertor by maximizing dissipation of plasma power and momentum inside the divertor. Here, we demonstrate, for the first time, that the electromagnetic (E × B) drifts exert a key influence on plasma and gas dynamics in a new advanced slot divertor in the DIII-D tokamak, named the small angle slot (SAS). SAS leverages the effect of drifts to achieve a highly dissipative divertor with electron temperature T e ≲ 10 eV over a wide range of plasma densities, for ion B × ∇B away from the divertor, as used for advanced tokamak operation on DIII-D. Modeling with the SOLPS-ITER code shows that for this drift direction, the E × B flow carries particles toward the outer common flux region (CFR) via the private flux region (PFR), reinforcing neutral recycling and enhancing divertor dissipation. In contrast, for the opposite field direction, the E × B flow carries particles away from the outer CFR into the PFR, offsetting the anticipated SAS geometric effects. Furthermore, this finding is an important step in the understanding of the behavior of advanced divertors for achieving a power exhaust solution for fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Model-based real-time surface heat flux and temperature estimation for the DIII-D tokamak

A control-oriented model for monitoring of wall power flux densities on the DIII-D tokamak has been successfully implemented and validated experimentally. Future reactors will have to withstand severe steady state high heat flux loads on plasma-facing components (PFCs). Due to the difficulty of directly-measuring local heat fluxes on these components, monitoring and protection of PFCs during the plasma discharge can benefit from simplified physics-based real-time (RT) functional models to estimate and guide heat load control. As a first step into the development, a control-oriented model for monitoring of wall power flux densities and temperatures on DIII-D tokamak has been successfully implemented. The paper discusses the experimental demonstration and comparison of the 2-D model-based wall heat flux algorithm on the DIII-Dinner wall limiter (IWL) against infra-red camera heat flux measurements for limited plasma configurations. The paper also reports on the benchmarking of the field line tracing environment, SMITER, developed at ITER organization on DIII-D tokamak against experimental IR diagnostic data and the derivation of the component shaping weighting factors for the 2-D model-based approach. Here, the extension of the model-based approach for surface temperature estimation on the DIII-D IWL is also presented.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Innovative approaches towards an economic fusion reactor

Nuclear fusion potentially offers a clean, environmentally friendly and intrinsically safe energy source with an abundant fuel supply. Magnetic fusion energy research is approaching a new era of fusion power reactor design and cons-truction planning. New physics understanding and powerful predictive tools have become available for improving fusion performance, developing and optimizing various magnetic confinement concepts. Emerging transformative enabling technologies can potentially mitigate and transform some present physics challenges from the possibly insurmountable to the potentially solvable.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Impact of ELM control techniques on tungsten sputtering in the DIII-D divertor and extrapolations to ITER

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 the experimental measurements of W sputtering during ELMs, conducted via fast neutral tungsten (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 flushing C impurities from the pedestal and reducing the divertor target electron temperature. No reduction of W sputtering was observed during the application of resonant magnetic perturbations (RMPs), in contrast to the prediction of the FSRM. Potential sources of this discrepancy are discussed. Lastly, the framework of the FSRM is utilized to predict intra-ELM W sputtering rates in ITER. It is concluded that W erosion during ELMs in ITER will be caused mainly by free-streaming fuel ions, but free-streaming seeded impurities (N or Ne) may increase the erosion rate significantly if present in the pedestal at even the 1% level. Impurity recycling is not expected to cause critical W erosion in ITER due to the very low target electron temperature.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Identifying the location of the OMP separatrix in DIII-D using power accounting

In order to identify reliable scalings for the scrape-off layer (SOL) power width it is necessary to know the location of the separatrix in divertor tokamaks as accurately as possible, specifically its location at the outside midplane (OMP) the standard reference location. Two methods are described which use power accounting to improve the accuracy of identifying the location of the OMP separatrix. The first uses the infrared-measured deposited power profile at the outer target as the primary input, the ' $P_{{\rm SOL}}^{{\rm exhaust}}$ method'. The second uses the measured power input to the SOL, obtained by subtracting the power radiated from inside the separatrix from the total heating power, the ' $P_{{\rm SOL}}^{{\rm input}}$ method'. Furthermore, these two power accounting methods are illustrated with the examples of 21 H-mode DIII-D discharges. High spatial resolution Thomson scattering measured profiles of ne and Te for the main SOL near the OMP are also used as primary input to the analysis; only between-edge localized mode data are used here. The Thomson profiles are used to calculate the electron parallel conducted heat flux profiles which are then matched to the measured $P_{{\rm SOL}}^{{\rm exhaust}}$ and $P_{{\rm SOL}}^{{\rm input}}$ by adjusting the location of the OMP separatrix relative to that of the Thomson data. For these attached discharges, it is found that the values of $R_{{\rm sep}}^{{\rm omp}}$ given by the two power accounting methods agree to within ~1 mm of each other and also to within ~1 mm of the values given by the 'standard DIII-D method' described by Porter et al (1998 Phys. Plasmas 5 1410). Lastly, the shifted $R_{{\rm sep}}^{{\rm omp}}$ results in only modest changes to the values of ne and Te at the OMP separatrix relative to the 'standard' values, increasing $n_{{\rm e}}^{{\rm sep}}$ by 8% and $T_{{\rm e}}^{{\rm sep}}$ by 20%.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Analysis of a tungsten sputtering experiment in DIII-D and code/data validation of high redeposition/reduced erosion

We analyze a DIII-D tokamak experiment where two tungsten spots on the removable DiMES divertor probe were exposed to 12 s of attached plasma conditions, with moderate strike point temperature and density (~20 eV, ~4.5 × 10 19 m –3 ), and 3% carbon impurity content. Both very small (1 mm diameter) and small (1 cm diameter) deposited samples were used for assessing gross and net tungsten sputtering erosion. The analysis uses a 3-D erosion/redeposition code package (REDEP/WBC), with input from a diagnostic-calibrated near-surface plasma code (OEDGE), and with focus on charge state resolved impinging carbon ion flux and energy. The tungsten surfaces are primarily sputtered by the carbon, in charge states +1 to +4. We predict high redeposition (~75%) of sputtered tungsten on the 1 cm spot—with consequent reduced net erosion—and this agrees well with post-exposure DiMES probe RBS analysis data. As a result, this study and recent related work is encouraging for erosion lifetime and non-contamination performance of tokamak reactor high-Z plasma facing components.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗