Engineering topics

McLean, Adam G.

Publications and source records attributed to McLean, Adam G..

4 records

DOE OSTI2023

Demonstration of a two-dimensional divertor Thomson scattering system on DIII-D

Recent upgrades providing two-dimensional divertor Thomson scattering (DTS-2D) measurements of T e and n e during a DIII-D plasma shot and a thorough description of system components and their functionality are presented. This system expands the capabilities of the existing single divertor Floor measurement location by introducing seven additional laser beam path options in the poloidal plane, spanning major radii from 1.062 to 1.335 m. The system redirects ~1 J, 50 Hz Nd:YAG laser pulses to the new beam paths within 20 ms, stepping through each path on the divertor Floor every 200 ms during a plasma shot. The laser is redirected using an ex-vessel, fast-steering mirror to one of eight in-vessel beam paths oriented underneath the vessel tiles. Up to eleven measurement positions per beam path, from –1.35 to –1.13 m below the machine midplane, are available by dynamically refocusing the ex-vessel collection fiber array using a high-speed linear stage. As a result, current measurement positions above the divertor Shelf are retained via a hole in the fixed, in-vessel mirror, allowing laser pass through.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2022

Determination of the characteristic magnetic pre-sheath length at divertor surfaces using micro-engineered targets on DiMES at DIII-D

The magnetic pre-sheath (MPS) width, 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. In this work, to determine L MPS , we fabricated micro-trenches (30×30×4 µm) via focused ion beam (FIB) milling on a silicon surface and exposed them to L-mode deuterium plasmas in DIII-D via the Divertor Material Evaluation System (DiMES) removable sample 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 model, based on a collisionless MPS, was used to calculate the azimuthal and polar deuterium ion angle distributions (IADs) 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 divertors to successfully predict plasma-materials interactions.

Erosion↗

DOE OSTI2021

Diverted Negative Triangularity plasmas on DIII–D: The benefit of high confinement without the liability of an edge pedestal

Diverted discharges at negative triangularity on the DIII-D tokamak sustain normalized confinement and pressure levels typical of standard H-mode scenarios (H 98y,2 ≅1, β N ≅3) without developing an edge pressure pedestal, despite the auxiliary power far exceeding the L → H power threshold expected from conventional scaling laws. The power degradation of confinement is substantially weaker than the ITER-89P scaling, resulting in a confinement factor that improves with increasing auxiliary power. Furthermore, the absence of the edge pedestal is beneficial in several aspects, such as eliminating the need for active mitigation or suppression of edge localized modes, low impurity retention and a reconstructed scrape-off layer heat flux width at the mid-plane that exceeds the ITPA multi-machine scaling law by up to 50%. Together with technological advantages granted by placing the divertor at larger radii, plasmas at Negative Triangularity without an edge pedestal feature both core confinement and power handling characteristics that are potentially suitable for operation in future fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

Understanding Plasma Divertor Detachment in Fusion Power Reactors

Key LLNL spectroscopic and modeling capabilities have been successfully applied to a range of ionizing and recombining plasmas in a magnetically confined fusion device to elucidate the roles of physical processes relevant to the ‘detached divertor’ regime essential for operation of future fusion power reactors. The primary goal of this work was improving theoretical models and the interpretability of plasmas near the ionizing/recombining balance, in order to improve predictive capability for divertor performance in state-of-the-art fluid-based modeling codes. To do so, contributors to a universal shortfall in predicted radiation, the underestimated impact of convective plasma drifts/flows, and the role of molecules in cold plasmas were identified in the DIII-D National Fusion Facility using newly installed multispectral extreme ultraviolet and visible spectroscopy, and advanced boundary modeling developed by LLNL. As a result of this LDRD, a powerful array of spectroscopic diagnostics that are each unique worldwide have been added to the DIII-D device for continued use in coming operational campaigns including research in boundary and core-edge integration to further expand operational space. Experiments scanning the detached regime have resulted in application of essential missing physics with plasma-boundary codes and a transformation in the recognition among international magnetic fusion programs of the formidable influence of convective plasma transport due to drifts in regimes including conditions necessary for future fusion power reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗