ORCID Profile
0000-0002-5266-4269
Current Organisation
University of California, San Diego
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Publisher: Elsevier BV
Date: 12-2023
Publisher: IOP Publishing
Date: 12-04-2022
Abstract: It is estimated that pilot plants and reactors may experience rates of net erosion and deposition of solid plasma facing component (PFC) material of 10 3 –10 5 kg yr −1 . Even if the net erosion (wear) problem can be solved, the redeposition of so much material has the potential for major interference with operation, including disruptions due to so-called ‘unidentified flying objects (UFOs)’ and unsafe dust levels. The potential implications appear to be no less serious than for plasma contact with the ertor target: a dust explosion or a major UFO-disruption could be as damaging for an actively-cooled deuterium-tritium (DT) tokamak as target failure. It will therefore be necessary to manage material deposits to prevent their fouling operation. This situation appears to require a fundamental paradigm shift with regard to meeting the challenge of taming the plasma–material interface: it appears that any acceptable solid PFC material will in effect be flow-through , like liquid–metal PFCs, although at far lower mass flow rates. Solid PFC material will have to be treated as a consumable , like brake pads in cars. ITER will use high-Z (tungsten) armor on the ertor targets and low-Z (beryllium) on the main walls. The ARIES-AT reactor design calls for a similar arrangement, but with SiC cladding on the main walls. Non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc) used as in situ replenishable, relatively thin—of order mm—claddings on a substrate which is resistant to neutron damage could provide a potential solution for the main walls, while reducing the risk of degrading the confined plasma. Separately, wall conditioning has proven essential for achieving high performance. For DT devices, however, standard methods appear to be unworkable, but recently powder droppers injecting low-Z material ∼continuously into discharges have been quite effective and may be usable in DT devices as well. The resulting massive generation of low-Z debris, however, has the same potential to seriously disrupt operation as noted above. Powder droppers provide a unique opportunity to carry out controlled studies on the management of low-Z slag in all current tokamaks, independent of whether their protection tiles use low-Z or high-Z material.
Publisher: IOP Publishing
Date: 04-04-2022
Abstract: The magnetic pre-sheath (MPS) length, L MPS , is a critical parameter to define the sheath potential, which controls the ion trajectory of low-Z species (D, T, He, and C), as well as the prompt re-deposition of high-Z species. To determine L MPS , we fabricated micro-trenches (30 × 30 × 4 μ m) via focused ion beam milling on a silicon surface and exposed them to L-mode deuterium plasmas in DIII-D via the ertor material evaluation system (DiMES) removable s le exposure probe. The areal distribution of impurity depositions, mainly consisting of carbon, was measured by energy-dispersive x-ray spectroscopy (EDS) to reveal the deuterium ion shadowing effect on the trench floors. The carbon deposition profiles showed that the erosion was maximized for the azimuthal direction of φ = −40° (referenced to the toroidal magnetic field direction) as well as the polar angle of θ = 80°. A Monte Carlo equation-of-motion (EOM) model, based on a collisionless MPS, was used to calculate the azimuthal and polar deuterium ion angle distributions (IADs) at the surface for a range of L MPS = k × ρ i , where ρ i is the ion gyro radius and k = 0.5–4. Then, gross erosion profiles were calculated by a Monte Carlo micro-patterning and roughness (MPR) code for ion sputtering using as input the calculated azimuthal and polar IADs for each value of k . Good agreement with the experimental C deposition profiles was obtained for the case k = 2.5–3.5. This result is consistent with a previous kinetic modeling prediction of k ∼ 3, as well as previous analytical investigations that predicted the L MPS to be several ion gyro radii. A validation of theoretical sheath models supports its applicability to ITER and pilot plant ertors to successfully predict plasma–materials interactions.
Publisher: IOP Publishing
Date: 05-08-2021
Publisher: IOP Publishing
Date: 27-11-2019
Publisher: IOP Publishing
Date: 28-09-2020
Publisher: IOP Publishing
Date: 18-03-2022
Abstract: The companion part A paper (Stangeby et al 2022) reports a number of independent estimates indicating that high-duty-cycle DT tokamaks starting with pilot plants will likely experience rates of net erosion and deposition of solid PFC, plasma facing component, material in the range of 10 3 to 10 4 kg yr −1 , regardless of the material used. The subsequent redeposition of such large quantities of material has the potential for major interference with tokamak operation. Similar levels and issues will be involved if ∼continuous low-Z powder dropping is used for surface conditioning of DT tokamaks, independent of the material used for the PFC armor. In Stangeby et al (2022) (part A) it is proposed that for high-duty-cycle DT tokamaks, non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc) used as in situ replenishable, relatively thin—of order mm—claddings on a substrate which is resistant to neutron damage could provide a potential solution for protecting the main walls, while reducing the risk of degrading the confined plasma. Assessment of whether such an approach is viable will require information, much of which is not available today. Section 6 of part A identifies a partial list of major physics questions that will need to be answered in order to make an informed assessment. This part B report describes R& D needed to be done in present tokamaks in order to answer many of these questions. Most of the required R& D is to establish better understanding of low-Z slag generation and to identify means to safely manage it. Powder droppers provide a unique opportunity to carry out controlled studies on the management of low-Z slag in current tokamaks, independent of whether their protection tiles use low-Z or high-Z material.
Publisher: IOP Publishing
Date: 06-12-2022
Abstract: We assess the toroidal magnetic field B t asymmetry in DIII-D due to a misalignment of the toroidal field coils with respect to the poloidal magnetic field coils and vacuum vessel. The peak-to-peak variation of the ertor strike point (SP) radius is measured to be 1 cm, with an n = 1 toroidal pattern. We use the centre of a narrow carbon deposition band on tungsten-coated ertor tiles just inside the outer strike point (OSP) as a proxy for the ertor SP location. The band occurred in a series of reverse B t discharges with the OSP positioned on the ertor inserts due to strong E × B drift transport of C from the inner to the outer SP through the private flux region. The variation in band radius (and hence the magnetic SP) is a (4.89 ± 0.31) mm shift toward (310 ± 4)° toroidal direction. These measurements agree well with previous measurements of the 3D magnetic field distribution (Luxon 2003 Nucl. Fusion 43 1813), simulations performed by the mafot field line integration code, and recent Langmuir probe measurements in the small-angle-slot (SAS) ertor (Watkins et al 2019 Nucl. Mater. Energy 18 46). Comparison of these measurements in the SAS ertor also indicates that there is the possibility of a tilt (in conjunction with the shift) of the B t coil field of (0.04 ± 0.07)° towards the toroidal angle of (215 ± 25)°. Previous measurements suggested a field misalignment of (4.6 ± 0.3) mm in the 270° toroidal direction, and a tilt of (0.06 ± 0.02)° toward the 114° toroidal direction, which is similar to the results reported here. These studies will be important for better understanding the radial variation of the toroidal strike line in DIII-D, for designing the new generation of SAS ertor, and for developing an understanding of the impact of error fields on tokamaks with tightly baffled slot ertors.
Publisher: IOP Publishing
Date: 21-04-2021
Abstract: Silicon carbide (SiC) represents a promising but largely untested plasma-facing material (PFM) for next-step fusion devices. In this work, an analytic mixed-material erosion model is developed by calculating the physical (via SDTrimSP) and chemical (via empirical scalings) sputtering yield from SiC, Si, and C. The Si content in the near-surface SiC layer is predicted to increase during D plasma bombardment due to more efficient physical and chemical sputtering of C relative to Si. Silicon erosion from SiC thereby occurs primarily from sputtering of the enriched Si layer, rather than directly from the SiC itself. SiC coatings on ATJ graphite, manufactured via chemical vapor deposition, were exposed to repeated H-mode plasma discharges in the DIII-D tokamak to test this model. The qualitative trends from analytic modeling are reproduced by the experimental measurements, obtained via spectroscopic inference using the S / XB method. Quantitatively the model slightly under-predicts measured erosion rates, which is attributed to uncertainties in the ion impact angle distribution, as well as the effect of edge-localized modes. After exposure, minimal changes to the macroscopic or microscopic surface morphology of the SiC coatings were observed. Compositional analysis reveals Si enrichment of about 10%, in line with expectations from the erosion model. Extrapolating to a DEMO-type device, an order-of-magnitude decrease in impurity sourcing, and up to a factor of 2 decrease in impurity radiation, is expected with SiC walls, relative to graphite, if low C plasma impurity content can be achieved. These favorable erosion properties motivate further investigations of SiC as a low- Z , non-metallic PFM.
Publisher: IOP Publishing
Date: 2020
Publisher: IOP Publishing
Date: 16-07-2021
Publisher: AIP Publishing
Date: 06-2019
DOI: 10.1063/1.5089895
Abstract: The free-streaming plus recycling model (FSRM) has recently been developed to understand and predict tungsten gross erosion rates from the ertor 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 ertor 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. Finally, 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 significant W erosion in ITER due to the very low target electron temperature.
Publisher: IOP Publishing
Date: 12-07-2023
Abstract: Type-I and type-II edge-localized-modes (ELMs) heat flux profiles measured at the DIII-D ertor feature a peak in the vicinity of the strike-point and a plateau in the scrape-off-layer (SOL), which extends to the first wall. The plateau is present in attached and detached ertors and it is found to originate with plasma bursts upstream in the SOL. The integrated ELM heat flux is distributed at ∼65% in the peak and ∼35% in this plateau. The parallel loss model, currently used at ITER to predict power loads to the walls, is benchmarked using these results in the primary and secondary ertors with unprecedented constraints using experimental input data for ELM size, radial velocity, energy, electron temperature and density, heat flux footprints and number of filaments. The model can reproduce the experimental near-SOL peak within ∼20%, but cannot match the SOL plateau. Employing a two-component approach for the ELM radial velocity, as guided by intermittent data, the full radial heat flux profile can be well matched. The ELM-averaged radial velocity at the separatrix, which explains profile widening, increases from ∼0.2 km s −1 in attached to ∼0.8 km s −1 in detached scenarios, as the ELM filaments’ path becomes electrically disconnected from the sheath at the target. The results presented here indicate filaments fragmentation as a possible mechanism for ELM transport to the far-SOL and provide evidence on the beneficial role of detachment to mitigate ELM flux in the ertor far-SOL. However, these findings imply that wall regions far from the strike points in future machines should be designed to withstand significant heat flux, even for small-ELM regimes.
Publisher: IOP Publishing
Date: 11-11-2019
Publisher: IOP Publishing
Date: 14-10-2020
Abstract: We analyzed recent DIII-D tokamak tungsten ertor probe experiments using advanced, coupled, sputter erosion/redeposition, plasma, and surface response code packages. Modeling is done for ELMing H-mode, and L-mode plasmas, impinging on various size tungsten deposits on Divertor Material Evaluation System (DiMES) carbon probes. The simulations compute 3D, full kinetic, sub-gyromotion, impurity sputtering and transport, including changes in tungsten surface composition and response due to mixed deuterium and carbon ions irradiation. Per our analysis, ELM (edge localized mode) plasma sputtering in DIII-D mostly involves free-streaming high energy (∼500–1000 eV) D + and C +6 ions, with high near-surface plasma density. L-Mode sputtering is due to impurity sputtering (C, W) only, with lower density. All cases show complete redeposition of tungsten on the ertor, with significant redeposition on the tungsten spots themselves, and low self-sputtering. Comparison of ELM plasma gross tungsten erosion simulation results with in-situ spectroscopic data is good, as are code/data comparisons of net erosion using post-exposure Rutherford backscattering (RBS) data for the L-mode probes. The analysis, extrapolated to a full tungsten ertor, implies low net erosion and negligible plasma contamination from sputtering. These results support the use of high-Z plasma facing surfaces in ITER and beyond.
Publisher: IOP Publishing
Date: 07-11-2019
Publisher: IOP Publishing
Date: 21-08-2019
Publisher: IOP Publishing
Date: 12-2021
Abstract: A set of experiments are planned to exploit the high SOL collisionality enabled by a tightly baffled slot ertor geometry to suppress tungsten leakage in DIII-D. A toroidal row of graphite tiles from the Small Angle Slot (SAS) ertor 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 ertor surface with high spatial and temporal resolution. In parallel, the bottom of the SAS ertor 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 ertor shape or the open, lower ertor. 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 ertor geometry. This coupled code-experiment validation effort will be used to stress-test physics models and build confidence in extrapolations to advanced, high-Z ertor geometries for next-step devices.
Publisher: Elsevier BV
Date: 12-2020
Publisher: IOP Publishing
Date: 09-02-2021
Abstract: Comprehensive upgrades to the dust transport code Dust in TOKamaks (DTOKS) that extend the plasma-dust interaction model are presented and compared with recent measurements of dust transport in DIII-D. Simulations incorporating variation in physical properties of graphite dust with temperature and size in a stationary plasma background suggest a substantial decrease in lifetimes due principally to thermal expansion. The trajectories of 53 dust grains identified from analysis of visible camera data taken across two similar shots were used to measure the dust particle velocity distributions. Dust tracks terminated mostly at the outer ertor strike point having a mean observation time of 2.1 ± 0.4 ms. Stochastic modelling of 200 graphite dust particles in the DIII-D tokamak performed with DTOKS-U using plasma simulations generated by OEDGE found similar behaviour, with particles ablating rapidly after acquiring a positive charge in the region close to the outer strike point, creating an acute source of neutral carbon atoms. The simulated mean lifetime, 11 ± 2 ms, showed approximate agreement with experimental observation when corrected by accounting for dust visibility and ignoring the longest trajectories 1.5 ± 0.2 ms. Synthetic diagnostic data generated from coupling the results of DTOKS-U with the visualisation software Calcam offers a powerful new tool for validation of simulations and predictive calculations of dust dynamics.
Publisher: IOP Publishing
Date: 24-08-2023
Abstract: Experiments have been conducted in the DIII-D tokamak to explore the in-situ growth of silicon-rich layers as a potential technique for real-time replenishment of surface coatings on plasma-facing components (PFCs) during steady-state long-pulse reactor operation. Silicon (Si) pellets of 1 mm diameter were injected into low- and high-confinement (L-mode and H-mode) plasma discharges with densities ranging from 3.9– 7.5 × 10 19 m −3 and input powers ranging from 5.5 to 9 MW. The small Si pellets were delivered with the impurity granule injector at frequencies ranging from 4 to 16 Hz corresponding to mass flow rates of 5–19 mg s −1 (1– 4.2 × 10 20 Si s −1 ) at cumulative amounts of up to 34 mg of Si per five-second discharge. Graphite s les were exposed to the scrape-off layer and private flux region plasmas through the ertor material evaluation system to evaluate the Si deposition on the ertor targets. The Si II emission at the s le correlates with silicon injection and suggests net surface Si-deposition in measurable amounts. Post-mortem analysis showed Si-rich coatings containing silicon oxides, of which SiO 2 is the dominant component. No evidence of SiC was found, which is attributed to low ertor surface temperatures. The in-situ and ex-situ analysis found that Si-rich coatings of at least 0.4–1.2 nm thickness have been deposited at 0.4–0.7 nm s −1 . The technique is estimated to coat a surface area of at least 0.94 m 2 on the outer ertor. These results demonstrate the potential of using real-time material injection to form Si-enriched layers on ertor PFCs during reactor operation.
Publisher: Elsevier BV
Date: 03-2021
Publisher: IOP Publishing
Date: 11-10-2021
Publisher: IOP Publishing
Date: 05-01-2022
Abstract: Near-separatrix impurity accumulation between the crown and the outer midplane of tokamaks is a common feature in results from codes such as SOLPS-ITER and DIVIMP however, experimental evidence of accumulation has only recently been obtained and is reported here. The codes find that the poloidal distribution of impurity ions in the scrape-off layer (SOL) depends primarily on toroidal field ( B T )-dependent parallel flow patterns of the background plasma and the parallel ion temperature gradient (∇ ‖ T ion ) force. Experimentally, Mach probes used in L-mode plasmas with favorable (for H-mode access) B T measure fast ( M ∼ 0.3–0.5) inner-target-directed (ITD) background plasma flows at the crown of single-null discharges. This study reports a set of DIVIMP simulations for two similar H-mode discharges from the DIII-D W metal rings c aign differing primarily in B T -direction to assess the effect that fast ITD flows have on the distribution of W ions in the SOL. It is found that for imposed ITD flows of M = 0.3, W ions that otherwise accumulate due to the ∇ ‖ T ion -force are largely flushed out. It is also found that doubling the radial diffusion coefficient from 0.3 to 0.6 m 2 s −1 prevents accumulation due to rapid cross-field transport into the far-SOL, where background plasma flows drain W ions to the ertors. Far-SOL W distributions from DIVIMP are then used to specify input to the impurity transport code 3DLIM, which is used to interpretively model collector probe (CP) deposition patterns measured in the ‘wall-SOL’. It is demonstrated that the deposition patterns are consistent with the DIVIMP predictions of near-SOL accumulation for the unfavorable- B T direction, and little/no accumulation for the favorable- B T direction. The wall-SOL CPs have thus provided the first experimental evidence, albeit indirect, of near-SOL W accumulation—finding it occurs for the unfavorable- B T direction only. For the favorable- B T direction, fast flows can largely prevent accumulation from occurring.
Location: United States of America
No related grants have been discovered for Dmitry Rudakov.