ORCID Profile
0000-0002-9543-7605
Current Organisation
International Islamic University Malaysia
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Publisher: Springer Science and Business Media LLC
Date: 11-07-2013
Publisher: Springer Science and Business Media LLC
Date: 04-2012
Publisher: EManuscript Technologies
Date: 2015
Publisher: Springer Science and Business Media LLC
Date: 12-2011
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: Informa UK Limited
Date: 2012
DOI: 10.1080/01635581.2012.630160
Abstract: Cat's whiskers (Orthosiphon stamineus) is commonly used as Java tea to treat kidney stones including a variety of angiogenesis-dependent diseases such as tumorous edema, rheumatism, diabetic blindness, and obesity. In the present study, antitumor potential of standardized 50% ethanol extract of O. stamineus leaves (EOS) was evaluated against colorectal tumor in athymic mice and antiangiogenic efficacy of EOS was investigated in human umbilical vein endothelial cells (HUVEC). EOS at 100 mg/kg caused 47.62 ± 6.4% suppression in tumor growth, while at 200 mg/kg it caused 83.39 ± 4.1% tumor regression. Tumor histology revealed significant reduction in extent of vascularization. Enzyme-linked immunosorbent assay showed EOS (200 mg/kg) significantly reduced the vascular endothelial growth factor (VEGF) level in vitro (211 ± 0.26 pg/ml cell lysate) as well as in vivo (90.9 ± 2 pg/g tissue homogenate) when compared to the control (378 ± 5 and 135.5 ± 4 pg, respectively). However, EOS was found to be noncytotoxic to colon cancer and endothelial cells. In vitro, EOS significantly inhibited the migration and tube formation of human umbilical vein endothelial cells (HUVECs). EOS suppressed VEGF-induced phosphorylation of VEGF receptor-2 in HUVECs. High performance liquid chromatography (HPLC) analysis of EOS showed high rosmarinic acid contents, whereas phytochemical analysis revealed high protein and phenolic contents. These results demonstrated that the antitumor activity of EOS may be due to its VEGF-targeted antiangiogenicity.
Publisher: IOP Publishing
Date: 05-08-2021
Publisher: IOP Publishing
Date: 27-11-2019
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: 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: JCDR Research and Publications
Date: 2016
Publisher: AIP Publishing
Date: 03-2011
DOI: 10.1063/1.3559492
Abstract: First measurements of the D+ parallel velocity, V∥D+, in L-mode discharges in the DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] tokamak boundary region at two poloidal locations, θ∼0° and θ∼255°, made using Mach probes, feature a peak with velocities of up to 80 km/s at the midplane last closed flux surface (LCFS), as high as ten times the charge exchange recombination C6+ toroidal velocity, VϕC6+, in the same location. The V∥D+ profiles are very asymmetric poloidally, by a factor of 8–10, and feature a local peak at the midplane. This peak, 1–2 cm wide, is located at or just inside the LCFS, and it suggests a large source of momentum in that location. This momentum source is quantified at ∼0.31 N m by using a simple momentum transport model. This is the most accurate measurement of the effects of so called “intrinsic” edge momentum source to date. The V∥D+ measurements are quantitatively consistent with a purely neoclassical computational modeling of V∥D+ by the code NEO [E. A. Belli and J. Candy, Plasma Phys. Controlled Fusion 50, 095010 (2008)], using VϕC6+ as input, for ρ∼0.7–0.95 at the two poloidal locations, where V∥D+ measurements exist. The midplane NEO-calculated V∥D+ grows larger than V∥C6+ in the steeper edge gradient region and trends to agreement with the probe-measured V∥D+ data near ρ∼1, where the local V∥D+ velocity peak exists. The measurements and computations were made in OH and L-mode discharges on an upper single null, with ion ∇BT drift away from the ertor. The rotating layer finding is similar in auxiliary heated discharges with and without external momentum input, except that at higher density the edge velocity weakens.
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: AIP Publishing
Date: 09-2016
DOI: 10.1063/1.4962683
Abstract: Bulk ion toroidal velocity profiles, V||D+, peaking at 40–60 km/s are observed with Mach probes in a narrow edge region of DIII-D discharges without external momentum input. This intrinsic rotation can be well reproduced by a first principle, collisionless kinetic loss model of thermal ion loss that predicts the existence of a loss-cone distribution in velocity space resulting in a co-Ip directed velocity. We consider two kinetic models, one of which includes turbulence-enhanced momentum transport, as well as the Pfirsch-Schluter (P-S) fluid mechanism. We measure a fine structure of the boundary radial electric field, Er, insofar ignored, featuring large (10–20 kV/m) positive peaks in the scrape off layer (SOL) at, or slightly inside, the last closed flux surface of these low power L- and H-mode discharges in DIII-D. The Er structure significantly affects the ion-loss model, extended to account for a non-uniform electric field. We also find that V||D+ is reduced when the magnetic topology is changed from lower single null to upper single null. The kinetic ion loss model containing turbulence-enhanced momentum transport can explain the reduction, as we find that the potential fluctuations decay with radius, while we need to invoke a topology-enhanced collisionality on the simpler kinetic model. The P-S mechanism fails to reproduce the d ing. We show a clear correlation between the near core V||C6+ velocity and the peak edge V||D+ in discharges with no external torque, further supporting the hypothesis that ion loss is the source for intrinsic torque in the present tokamaks. However, we also show that when external torque is injected in the core, it can complete with, and eventually overwhelm, the edge source, thus determining the near SOL flows. Finally, we show some additional evidence that the ion/electron distribution in the SOL is non-Maxwellian.
Publisher: Science Alert
Date: 15-12-2010
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: IOP Publishing
Date: 14-09-2016
DOI: 10.1088/0029-5515/56/12/126010
Abstract: A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable ertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable ertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced ertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor ertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans (2) Develop and validate key ertor design concepts and codes through innovative variations in physical structure and magnetic geometry (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced ertor for DIII-D to enable highly dissipative ertor operation at core density ( n e / n GW ), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.
Publisher: Elsevier BV
Date: 12-2020
Publisher: IOP Publishing
Date: 05-08-2019
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: No location found
Location: United States of America
No related grants have been discovered for M J Siddiqui.