Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “DIVERTORS”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Divertor characterization and access to dissipative divertor conditions in negative triangularity discharges in DIII-D

Reduced radial transport, short midplane-to-target parallel connection lengths, and a strong effect of cross-field drifts were responsible for the high densities required for detachment in strong negative triangularity (NT) configurations in DIII-D. Dissipative divertor conditions were achieved in NT discharges at different triangularity, injected power, plasma current, and toroidal field direction. Differences between negative and positive triangularity (PT) discharges are analyzed in this paper to understand the requirements for access to detached divertor conditions: power balance, geometry, radial transport and effect of cross-field particle drifts. Parametric dependencies of access to detachment on plasma current and power flowing into the scrape-off layer remained similar in negative and PT and impurity seeding was observed to reduce the density needed to detach by up to 30% at the expense of core impurity dilution. The impact of triangularity on core-edge integration was tested varying bottom triangularity at fixed top triangularity. The high density needed to detach was not intrinsic to the NT edge as shapes with positive lower triangularity and negative upper triangularity were able to detach at lower upstream densities while maintaining an ELM-free NT edge. Confinement degradation at deeper detachment levels was however observed in all NT shapes, often associated with radiation instabilities.

detachment↗
DOE OSTI2018

Effects of two-dimensional magnetic uncertainties and three-dimensional error and perturbation fields on the Small Angle Slot divertor geometry and topology [Effects of two- and three-dimensional magnetic fields on the Small Angle Slot divertor magnetic topology]

The Small Angle Slot (SAS) was recently installed on DIII-D as an advanced divertor, promising easier plasma detachment and lower temperatures across the whole target. A twofold study of the SAS magnetic topology is presented in this paper. On one hand, a twodimensional uncertainty quantification analysis is carried out through a Monte Carlo approach in order to understand the level of accuracy of two-dimensional equilibrium computations in reconstructing the strike point and angle onto the divertor. Under typical experimental conditions, the uncertainties are found to be roughly 6.8 mm and 0.56 deg, respectively. On the other hand, a three-dimensional “vacuum” analysis is carried out to understand the effects of typical external perturbation fields on the scrape-off layer topology. When the threedimensional I-coils are switched on, poloidally-localized lobes are found to appear, grow, and hit the SAS target, although barely, even for 5 kA; at the same time, the strike point modulation is found to be roughly 1.8 mm and thus negligible for most purposes. Furthermore, such results complement previous two-dimensional analyses in characterizing typical SAS equilibria and provide useful background information for planning and interpreting SAS experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Imaging divertor strike point splitting in RMP ELM suppression experiments in the DIII-D tokamak [Imaging divertor strike point splitting in RMP ELM suppression experiments]

Fast visible imaging of the lower divertor from above is used to study the structure and dynamics of lobes induced by resonant magnetic perturbations (RMPs) in Edge-Localized Mode (ELM) suppression experiments in DIII-D. The best compromise between the amount of light and sharp imaging was obtained using emission at 601 nm from Fulcher band molecular deuterium. Multiple spatially resolved peaks in the D 2 emission, taken as a proxy for the particle flux, are readily resolved during RMPs, in contrast to the heat flux measured by infrared cameras, which shows little spatial structure in ITER-like conditions. The 25 mm objective lens provides high spatial resolution (2–4 mm/pixel) from the centerpost to the outer shelf over 40° toroidally that overlaps the field of view of the IRTV that measures the divertor heat flux, allowing direct comparison in non-axisymmetric discharges. The image is coupled to a Phantom 7.3 camera using a Schott wound fiber bundle, providing high temporal resolution that allows the lobe dynamics to be resolved between ELMs and across ELM suppression onset. Furthermore, these measurements are used to study the heat and particle flux in 3D magnetic fields and to validate models for the plasma response to RMPs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Flow instabilities in non-uniformly heated helium jet arrays used for divertor PFCs

In this study, due to a lack of prototypical experimental data, little is known about the off-normal behavior of recently proposed divertor jet cooling concepts. This article describes a computational fluid dynamics (CFD) study on two jet array designs to investigate their susceptibility to parallel flow instabilities induced by non-uniform heating and large increases in the helium outlet temperature. The study compared a single 25-jet helium-cooled modular divertor (HEMJ) thimble and a micro-jet array with 116 jets. Both have pure tungsten armor and a total mass flow rate of 10 g/s at a 600 °C inlet temperature. We investigated flow perturbations caused by a 30 MW/m 2 off-normal heat flux applied over a 25 mm 2 area in addition to the nominal 5 MW/m 2 applied over a 75 mm 2 portion of the face. The micro-jet array exhibited lower temperatures and a more uniform surface temperature distribution than the HEMJ thimble. We also investigated the response of a manifolded nine-finger HEMJ assembly using the nominal heat flux and a 274 mm 2 heated area. For the 30 MW/m2 case, the micro-jet array absorbed 750 W in the helium with a maximum armor surface temperature of 1280 °C and a fluid/solid interface temperature of 801 °C. The HEMJ absorbed 750 W with a maximum armor surface temperature of 1411 °C and a fluid/solid interface temperature of 844 °C. For comparison, both the single HEMJ finger and the micro-jet array used 5-mm-thick tungsten armor. The ratio of maximum to average temperature and variations in the local heat transfer coefficient were lower for the micro-jet array compared to the HEMJ device. Although high heat flux testing is required to validate the results obtained in these simulations, the results provide important guidance in jet design and manifolding to increase heat removal while providing more even temperature distribution and minimizing non-uniformity in the gas flow and thermal stresses at the armor joint.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Validation of EDGDE2D-EIRENE predicted 2D distributions of electron temperature and density against divertor Thomson scattering measurements in the low-field side divertor leg in DIII-D

Comparisons of profiles of the electron temperature (T e ), density (n e ) and pressure (p e ) measured with Divertor Thomson Scattering in DIII-D low-confinement mode discharges to predictions from the edge fluid code EDGE2D-EIRENE show that the models implemented in EDGE2D-EIRENE predict the measurements within their collective uncertainties if the T e at the separatrix (T e,sep ) is 10 eV, or higher. The simulations do not predict, however, the peaked T e and ne profiles measured adjacent to the target plate when T e,sep , is below 10 eV, i.e., for the plasma downstream from the region of ionization of deuterium atoms. Inclusion of cross-field drifts and a five-fold reduction of radial transport cannot reconcile the discrepancy between the measurements and predictions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma

One of the main advantages of using tungsten (W) as a plasma facing material (PFM) is its low uptake and retention of tritium. However, in high purity (ITER grade) W, hydrogenic retention increases significantly with neutron-induced displacement damage in the W lattice. This experiment examines an alternative W grade PFM, ultra-fine grain (UFG) W, to compare its retention properties with ITER grade W after 12 MeV Si ion displacement damage up to 0.6 dpa (displacements per atom.) Following exposure to plasma in the DIII-D divertor, D retention was then assessed with Nuclear Reaction Analysis (NRA) depth profiling up to 3.5 µm and thermal desorption spectrometry (TDS). Undamaged specimens were also included in our test matrix for comparison. For all samples, D release peaks were observed during TDS at approximately 200 °C and 750 °C. For the ITER-grade W specimens, the intensity of the 750 °C release peak was more pronounced for specimens that had been pre-damaged. Conversely, UFG samples that had been damaged by 12 MeV Si showed enhancement of the lower temperature release peak (200 °C). NRA profiles also reveal a higher D concentration for UFG W samples up to the peak in the damage profile at a depth of 2 m. Overall, we observed that the total trapped inventory in UFG W was 20% higher than ITER grade W in the undamaged case and 10% higher in the damaged case. A comparison of NRA and TDS data indicates that a larger fraction of the total retained D is trapped near the surface (86–100%) in UFG W pre-damaged to 0.6 dpa compared with ITER grade W (39–61%). Further examination of the UFG material with microscopy is recommended for a definitive determination of the types of defects responsible for D trapping. Our results highlight some potential trade-offs associated UFG W regarding its performance from a tritium retention standpoint. That said, our TDS results indicate that this enhanced inventory can be released by baking at relatively low temperatures (<500 °C), providing an avenue for minimizing tritium retention in this material that would be practical for implementation in a tokamak.

Barton, J↗
DOE OSTI2019

Error field impact on mode locking and divertor heat flux in NSTX-U

During the 2016 NSTX-U experimental campaign, locked modes in the plasma edge presented clear evidence of the presence of error fields. Extensive metrology and plasma response modeling with IPEC and M3D-C1 have been conducted to understand the various sources of error fields in NSTX-U as built in 2016, and to determine which of these sources have the greatest effect on the plasma. In particular, modeling finds that the error field from misalignment of the toroidal field (TF) coils may have a significant effect on the plasma. The response to the TF error field is shown to depend on the presence of a q = 1 surface, in qualitative agreement with experimental observations. It is found that certain characteristics of the TF error field present new challenges for error field correction. Specifically, the error field spectrum differs significantly from that of coils on the low-field side (such as the NSTX-U error field correction coils), and does not resonate strongly with the dominant kink mode, thus potentially requiring a multi-mode correction. Moreover, to mitigate heat fluxes using poloidal flux expansion, the pitch angle at the divertor plates must be small (~1°). It is shown that uncorrected error fields may result in potentially significant local perturbation to the pitch angle. Estimates for coil alignment tolerances in NSTX-U are derived based on consideration of both heat flux and core resonant fields independently.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗