Publications
Showing 72 results for Author: Robert S. Wilcox
Aug, 2026
Journal
Understanding carbon sourcing and transport originating from the helicon antenna surfaces during high-power helicon discharge in DIII-D Tokamak
The high-power helicon wave system in the DIII-D tokamak could potentially introduce new plasma-material interaction (PMI) challenges owing to rectified radio-frequency (RF) sheath potentials that develop near the antenna and surrounding plasma-facing components. We present the first application of the STRIPE (Simulated Transport of RF Impurity Production and Emission) fra…
May, 2026
Journal
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, an…
Mar, 2026
Journal
The effectiveness of D2 pellet and gas injection in reducing intra-ELM tungsten erosion and heat flux in the DIII-D small angle slot divertor
Edge localized modes (ELMs) in H-mode plasmas can melt and erode plasma-facing components (PFCs) and lead to impurities in the core, reducing confinement. This study analyzes the use of D2 pellet and gas injection for ELM mitigation on the DIII-D tokamak during the 2022 Small Angle Slot V-shaped Tungsten (W) (SAS-VW) divertor campaign, reducing W erosion and heat flux duri…
Feb, 2026
Journal
Towards self-consistent integrated modeling of the tokamak pedestal, scrape-off layer, and divertor using SOLPS-ITER and EPED
A predictive, self-consistent integrated model of the tokamak edge plasma and neutral system has been developed and validated against experimental data from DIII-D, spanning from the divertor target and scrape-off layer to the top of the pedestal. The model uses the SOLPS-ITER fluid boundary plasma and neutral code to constrain the EPED prediction of the pedestal height an…
Jan, 2026
Journal
Taming harmful bursts and heat flux in high-confinement tokamak plasmas
A major challenge in tokamak fusion research is first-wall erosion caused by steady heat loads and sudden energy bursts known as edge-localized modes. Divertor detachment reduces steady-state heat flux, while resonant magnetic perturbations can suppress these instabilities. However, integrating the two has been difficult because they require conflicting operating condition…
Nov, 2025
Journal
Challenges and approaches to interpretive modeling of boundary plasma and neutral transport in a closed, pumped divertor
An experimental discharge from the DIII-D tokamak is modeled using the SOLPS-ITER code suite and compared against measurements in the pumped and relatively closed upper divertor. Uncertainties of boundary plasma simulations are identified by attempting to match code inputs to experimental conditions, including iteratively solving transport coefficients to match upstream ex…
Oct, 2025
Journal
Design and Implementation of Surface Eroding Thermocouples for Stage-One Modular Divertor Integration at the DIII-D Tokamak
The DIII-D has upgraded its upper divertor to a modular system using copper alloy pedestals to alter the divertor geometry without changing the vessel structure. Six new graphite tiles were designed for the shape and volume rise (SVR) divertor. These new SVR tiles facilitated the formation of a poloidal array of 27 surface eroding thermocouples (SETCs) in the upper diverto…
Sep, 2025
Journal
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, an…
Sep, 2025
Journal
Recrystallization, cracking, and erosion of dispersoid-strengthened tungsten materials during exposure to divertor plasmas
In this study, we investigated the effects of combined intense particle and heat flux exposure on advanced tungsten plasma-facing materials within the DIII-D fusion facility. Our test matrix included two types of dispersoid-strengthened tungsten (containing either 100 nm diameter TiO2 or Ni particles), along with high-purity polycrystalline tungsten as a reference. This ex…