Results 1 to 10 of about 4,697,162 (157)
ECRH/EBWH system for NSTX-U [PDF]
EPJ Web of Conferences, 2012 The National Spherical Torus Experiment Upgrade (NSTX-U) will operate at an axial toroidal field of up to 1 T, about twice the field available on NSTX. A 28 GHz electron cylotron resonance heating (ECRH) system is currently being planned for NSTX-U.
Hosea J.C. +4 more
doaj +6 more sources
Advances in Boronization on NSTX-Upgrade
Nuclear Materials and Energy, 2017 Boronization has been effective in reducing plasma impurities and enabling access to higher density, higher confinement plasmas in many magnetic fusion devices. The National Spherical Torus eXperiment, NSTX, has recently undergone a major upgrade to NSTX-
Filippo Scotti, Felipe Bedoya, Dang Cai
exaly +4 more sources
Global modeling of wall material migration following boronization in NSTX-U
Nuclear Materials and Energy, 2019 Boronization is commonly utilized in tokamaks to suppress intrinsic impurities, most notably oxygen from residual water vapor. However, this is a temporary solution, as oxygen levels typically return to pre-boronization levels following repeated plasma ...
J.H. Nichols +6 more
doaj +2 more sources
Nuclear Materials and Energy, 2019 Tokamak plasma facing components have surface roughness that can cause microscopic spatial variations in erosion and deposition and hence influence material migration, erosion lifetime, dust and tritium accumulation, and plasma contamination.
C.H. Skinner +4 more
doaj +2 more sources
Nuclear Materials and Energy, 2018 Boronization is a Plasma Facing Component (PFC) conditioning technique widely used in tokamak machines. The National Spherical Torus Experiment-Upgrade (NSTX-U) applied this conditioning, using a plasma glow with a deuterated Trimethyl-boron (d-TMB) and ...
Filippo Scotti +2 more
exaly +3 more sources
Comparing lithium vapor box designs in a high heat flux scenario using SOLPS-ITER
Nuclear Materials and Energy, 2023 SOLPS calculations of lithium vapor box divertor designs on NSTX-U are presented. Predictive high power simulations (Pheat=10 MW, qpeakunmitigated∼65 MW/m2) are used to compare and contrast two divertor designs.
E.D. Emdee, R.J. Goldston
doaj +1 more source
Predictive modeling of a lithium vapor box divertor in NSTX-U using SOLPS-ITER
Nuclear Materials and Energy, 2021 The unmitigated heat flux in attached operation of a fusion power plant is predicted to be destructive to any solid divertor surface. Detachment, whereby the plasma pressure drops significantly before reaching the divertor target thus greatly reducing ...
E.D. Emdee +3 more
doaj +1 more source
Advances in Materials Science and Engineering, Volume 2014, Issue 1, 2014., 2014 The large volume vacuum systems are used in many industrial operations and research laboratories. Accidents in these systems should have a relevant economical and safety impact. A loss of vacuum accident (LOVA) due to a failure of the main vacuum vessel can result in a fast pressurization of the vessel and consequent mobilization dispersion of ...
A. Malizia +6 more
wiley +1 more source
EPJ Web of Conferences, 2017 30 MHz fast-wave heating may be an effective tool for non-inductively ramping low-current plasmas to a level suitable for initiating up to 12 MW of neutral beam injection on the National Spherical Tokamak Experiment Upgrade (NSTX-U).
Taylor Gary +8 more
doaj +1 more source
The effect of gas injection location on a lithium vapor box divertor in NSTX-U
Nuclear Fusion, 2023 The lithium vapor box divertor is a proposed divertor design that will minimize contamination of the upstream plasma in a fusion device, while also ensuring protection of the target.
E.D. Emdee, R.J. Goldston
doaj +1 more source