Expanding the role of impurity spectroscopy for investigating the physics of high-Z dissipative divertors

New techniques that attempt to more fully exploit spectroscopic diagnostics in the divertor and pedestal region during highly dissipative scenarios are demonstrated using experimental results from recent low-Z seeding experiments on Alcator C-Mod, JET and ASDEX Upgrade. To exhaust power at high parallel heat flux, q ‖ > 1 GW/m 2 , while minimizing erosion, reactors with solid, high-Z plasma facing components (PFCs) are expected to use extrinsic impurity seeding. Due to transport and atomic physics processes which impact impurity ionization balance, so-called ‘non-coronal’ effects, we do not accurately know and have yet to demonstrate the maximum q ‖ which can be mitigated in a tokamak. Radiation enhancement for nitrogen is shown to arise primarily from changes in Li- and Be-like charge states on open field lines, but also through transport-driven enhancement of H- and He-like charge states in the pedestal region. Measurements are presented from nitrogen seeded H-mode and L-mode plasmas where emission from N 1+ through N 6+ are observed. Active charge exchange spectroscopy of partially ionized low-Z impuri- ties in the plasma edge is explored to measure N 5+ and N 6+ within the confined plasma, while passive UV and visible spectroscopy is used to measure N 1+ -N 4+ in the boundary. Examples from recent JET and Alcator C-Mod experiments which employ nitrogen seeding highlight how improving spectroscopic cov- erage can be used to gain empirical insight and provide more data to validate boundary simulations.EURATOM 63305

EC62-219 Nebraska Swine Production Report

This 1962 Nebraska Swine Production Report was developed by the Animal Husbandry Department of the University of Nebraska-Lincoln. Authors from the departments of Animal Husbandry, Agricultural Economics Veterinary Science, Agricultural Engineering contributed to this publication. It covers the following areas: breeding, feeding, economics, disease control, mechanization, housing and equipment

High-field side scrape-off layer investigation: Plasma profiles and impurity screening behavior in near-double-null configurations

New experiments on Alcator C-Mod reveal that the favorable impurity screening characteristics of the high-field side (HFS) scrape-off layer (SOL), previously reported for single null geometries, is retained in double null configurations, despite the formation of an extremely thin SOL. In balanced double-null, nitrogen injected locally into the HFS SOL is better screened by a factor of 2.5 compared to the same injection into the low field side (LFS) SOL. This result is insensitive to plasma current and Greenwald fraction. Nitrogen injected into the HFS SOL is not as well screened (only a factor of 1.5 improvement over LFS) in unbalanced double-null discharges, when the primary divertor is in the direction of B×∇B. In this configuration, impurity ‘plume’ emission patterns indicate that an opposing E × B drift competes with the parallel impurity flow to the divertor. In balanced double-null plasmas, the dispersal pattern exhibits a dominant E × B motion. Unbalanced discharges with the primary divertor opposite the direction of B×∇B exhibit excellent HFS screening characteristics – a factor of 5 enhancement compared to LFS. These data support the idea that future tokamaks should locate all RF actuators and close-fitting wall structures on the HFS and employ near-double-null magnetic topologies, both to precisely control plasma conditions at the antenna/plasma interface and to maximally mitigate the impact of local impurity sources arising from plasma-material interactions. Keywords: Alcator C-Mod; Impurity screening; Double null; High field side scrape-off layerUnited States. Department of Energy (Contract DE-FC02-99ER54512

Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod’s high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes >40 MW m−2 down to 1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions

Alcator C-Mod: research in support of ITER and steps beyond

This paper presents an overview of recent highlights from research on Alcator C-Mod. Significant progress has been made across all research areas over the last two years, with particular emphasis on divertor physics and power handling, plasma–material interaction studies, edge localized mode-suppressed pedestal dynamics, core transport and turbulence, and RF heating and current drive utilizing ion cyclotron and lower hybrid tools. Specific results of particular relevance to ITER include: inner wall SOL transport studies that have led, together with results from other experiments, to the change of the detailed shape of the inner wall in ITER; runaway electron studies showing that the critical electric field required for runaway generation is much higher than predicted from collisional theory; core tungsten impurity transport studies reveal that tungsten accumulation is naturally avoided in typical C-Mod conditions.United States. Department of Energy (DE-FC02-99ER54512-CMOD)United States. Department of Energy (DE-AC02-09CH11466)United States. Department of Energy (DE-FG02-96ER-54373)United States. Department of Energy (DE-FG02-94ER54235

Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod’s high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes >40 MW/m2 down to 1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions

High-field side scrape-off layer investigation: Plasma profiles and impurity screening behavior in near-double-null configurations

New experiments on Alcator C-Mod reveal that the favorable impurity screening characteristics of the high-field side (HFS) scrape-off layer (SOL), previously reported for single null geometries, is retained in double null configurations, despite the formation of an extremely thin SOL. In balanced double-null, nitrogen injected locally into the HFS SOL is better screened by a factor of 2.5 compared to the same injection into the low field side (LFS) SOL. This result is insensitive to plasma current and Greenwald fraction. Nitrogen injected into the HFS SOL is not as well screened (only a factor of 1.5 improvement over LFS) in unbalanced double-null discharges, when the primary divertor is in the direction of B×∇B. In this configuration, impurity ‘plume’ emission patterns indicate that an opposing E×B drift competes with the parallel impurity flow to the divertor. In balanced double-null plasmas, the dispersal pattern exhibits a dominant E×B motion. Unbalanced discharges with the primary divertor opposite the direction of B×∇B exhibit excellent HFS screening characteristics – a factor of 5 enhancement compared to LFS. These data support the idea that future tokamaks should locate all RF actuators and close-fitting wall structures on the HFS and employ near-double-null magnetic topologies, both to precisely control plasma conditions at the antenna/plasma interface and to maximally mitigate the impact of local impurity sources arising from plasma-material interactions. Keywords: Alcator C-Mod, Impurity screening, Double null, High field side scrape-off laye

Core impurity transport in Alcator C-Mod L-, I- and H-mode plasmas

Core impurity transport has been investigated for a variety of confinement regimes in Alcator C-Mod plasmas from x-ray emission following injection of medium and high Z materials. In ohmic L-mode discharges, impurity transport is anomalous (D[subscript eff] ≫ D[subscript nc]) and changes very little across the LOC/SOC boundary. In ion cyclotron range of frequencies (ICRF) heated L-mode plasmas, the core impurity confinement time decreases with increasing ICRF input power (and subsequent increasing electron temperature) and increases with plasma current. Nearly identical impurity confinement characteristics are observed in I-mode plasmas. In enhanced D[subscript α] H-mode discharges the core impurity confinement times are much longer. There is a strong connection between core impurity confinement time and the edge density gradient across all confinement regimes studied here. Deduced central impurity density profiles in stationary plasmas are generally flat, in spite of large amplitude sawtooth oscillations, and there is little evidence of impurity convection inside of r/a = 0.3 when averaged over sawteeth.United States. Department of Energy (Contract DE-FC02-99ER54512)United States. Dept. of Energy. Fusion Energy Postdoctoral Research Program (Oak Ridge Institute for Science and Education