Development of rugged hydrogen sensors for measuring charge-exchange neutral flux at the wall and divertor
Abstract not provided.
Publications
Development of rugged hydrogen sensors for measuring charge-exchange neutral flux at the wall and divertor
Abstract not provided.
MODEL VALIDATION ON EAST AND DIII-D EXPERIMENTS TOWARDS UNDERSTANDING OF HIGH-Z MATERIAL EROSION AND MIGRATION IN A MIXED MATERIALS ENVIRONMENT
Abstract not provided.
CHARACTERIZATION OF SOL IMPURITY TRANSPORT AND DIVERTOR LEAKAGE USING ISOTOPIC TUNGSTEN DURING THE DIII-D METAL RINGS CAMPAIGN
Abstract not provided.
Attractive retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma
Abstract not provided.
Measurements of tungsten migration in the DIII-D divertor
Physica Scripta
An experimental study of migration of tungsten in the DIII-D divertor is described, in which the outer strike point of L-mode plasmas was positioned on a toroidal ring of tungsten-coated metal inserts. Net deposition of tungsten on the divertor just outside the strike point was measured on graphite samples exposed to various plasma durations using the divertor materials evaluation system. Tungsten coverage, measured by Rutherford backscattering spectroscopy (RBS), was found to be low and nearly independent of both radius and exposure time closer to the strike point, whereas farther from the strike point the W coverage was much larger and increased with exposure time. Depth profiles from RBS show this was due to accumulation of thicker mixed-material deposits farther from the strike point where the plasma temperature is lower. These results are consistent with a low near-surface steady-state coverage on graphite undergoing net erosion, and continuing accumulation in regions of net deposition. This experiment provides data needed to validate, and further improve computational simulations of erosion and deposition of material on plasma-facing components and transport of impurities in magnetic fusion devices. Such simulations are underway and will be reported later.
Measurements of Tungsten Migration in the DIII-D Divertor
Abstract not provided.
Measurements of Tungsten Migration in the DIII-D Divertor
Abstract not provided.
The inter-ELM W erosion profile in DIII-D H-mode discharges and benchmarking with ERO modeling
Abstract not provided.
Thermal Management of Tungsten Leading Edges in DIII-D and ITER
Abstract not provided.
Thermal Management of Tungsten Leading Edges in DIII-D and ITER
Abstract not provided.
Institutional Overview - Sandia National Labs
Abstract not provided.
Control of High-Z PFC Erosion by Local Gas Injection in DIII-D
Abstract not provided.
Divertor Sheath Power Studies in DIII-D Using Fixed Langmuir Probes and 3-D Modeling of Tile Heat Flows
Abstract not provided.
Divertor Sheath Power Studies in DIII-D Using Fixed Langmuir Probes and 3-D Modeling of Tile Heat Flows
Abstract not provided.
Sheath Power Transmission Studies on DIII-D Divertor and GIXRD Measurements of Tungsten Fuzz
Abstract not provided.
Measurement of net versus gross erosion of molybdenum divertor surface in DIII-D
Abstract not provided.
Sheath Power Transmission and Divertor Probe Analysis
Abstract not provided.
Overview of PMI at SNL-CA
Abstract not provided.
OEDGE modeling of the DIII-D double null 13CH4 puffing experiment
Unbalanced double null ELMy H-mode configurations in DIII-D are used to simulate the situation in ITER high triangularity, burning plasma magnetic equilibria, where the second X-point lies close to the top of the vacuum vessel, creating a secondary divertor region at the upper blanket modules. The measured plasma conditions in the outer secondary divertor closely duplicated those projected for ITER. {sup 13}CH{sub 4} was injected into the secondary outer divertor to simulate sputtering there. The majority of the {sup 13}C found was in the secondary outer divertor. This material migration pattern is radically different than that observed for main wall {sup 13}CH{sub 4} injections into single null configurations where the deposition is primarily at the inner divertor. The implications for tritium codeposition resulting from sputtering at the secondary divertor in ITER are significant since release of tritium from Be co-deposits at the main wall bake temperature for ITER, 240 C, is incomplete. The principal features of the measured {sup 13}C deposition pattern have been replicated by the OEDGE interpretive code.
DiMES studies of temperature dependence of carbon erosion and re-deposition in the DIII-D divertor
Physica Scripta
Abstract not provided.
First tests of molybdenum mirrors for ITER diagnostics in DIII-D divertor
Review of Scientific Instruments
Abstract not provided.
Transport and deposition of 13C from methane injection into partially detached divertor H-mode plasmas in DIII-D
Abstract not provided.
Far SOL transport and mail wall plasma interaction in DIII-D
Abstract not provided.
DIVIMP modeling of the toroidally-symmetrical injection of 13 CH4 into the upper SOL of DIII-D
As part of a study of carbon-tritium co-deposition, we carried out an experiment on DIII-D involving a toroidally symmetric injection of {sup 13}CH{sub 4} at the top of a LSN discharge. A Monte Carlo code, DIVIMP-HC, which includes molecular breakup of hydrocarbons, was used to model the region near the puff. The interpretive analysis indicates a parallel flow in the SOL of M {parallel} {approx} 0.4 directed toward the inner divertor. The CH{sub 4} is ionized in the periphery of the SOL and so the particle confinement time, T{sub C}, is not high, only {approx} 5 ms, and about 4X lower than if the CH{sub 4} were ionized at the separatrix. For such a wall injection location, however, approximately 60-75% of the CH{sub 4} gets ionized to C{sup +}, C{sup 2+}, etc., and is efficiently transported along the SOL to the inner divertor, trapping hydrogen by co-deposition there.
25 Results