LHCD during current ramp experiments on Alcator C-Mod

The lower hybrid current drive (LHCD) system on Alcator C-Mod is capable of sustaining fully non-inductive discharges for multiple current relaxation times (τcr ∼ 200 ms) at line averaged densities in the range of 5x1019 m-3. Some of these non-inductive discharges develop unstable MHD modes that can greatly reduce current drive performance, particularly in discharges with plasma current of 0.5 MA or less [1,2]. Avoiding these unstable MHD modes motivated an experiment to test if the stable current profile shape of a higher current non-inductive discharge could be achieved in a lower current discharge. Starting from a discharge at 0.8 MA, the plasma current was ramped down to 0.5 MA over 200 ms. The surface voltage of the plasma swings negative during the ramp, with the loop voltage reversal impacting the edge fast electron measurements immediately. Little change can be seen during the Ip ramp in the core fast electron measurements, indicating that the loop voltage reversal does not penetrate fully to the magnetic axis on the timescale of the current ramp. The resulting discharge did not exhibit deleterious MHD instabilities, however the existence of this one discharge does not necessarily represent a robust solution to the problem

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

Physics research on the TCV tokamak facility: from conventional to alternative scenarios and beyond

The research program of the TCV tokamak ranges from conventional to advanced-tokamak scenarios and alternative divertor configurations, to exploratory plasmas driven by theoretical insight, exploiting the device’s unique shaping capabilities. Disruption avoidance by real-time locked mode prevention or unlocking with electron-cyclotron resonance heating (ECRH) was thoroughly documented, using magnetic and radiation triggers. Runaway generation with high-Z noble-gas injection and runaway dissipation by subsequent Ne or Ar injection were studied for model validation. The new 1 MW neutral beam injector has expanded the parameter range, now encompassing ELMy H-modes in an ITER-like shape and nearly non-inductive H-mode discharges sustained by electron cyclotron and neutral beam current drive. In the H-mode, the pedestal pressure increases modestly with nitrogen seeding while fueling moves the density pedestal outwards, but the plasma stored energy is largely uncorrelated to either seeding or fueling. High fueling at high triangularity is key to accessing the attractive small edge-localized mode (type-II) regime. Turbulence is reduced in the core at negative triangularity, consistent with increased confinement and in accord with global gyrokinetic simulations. The geodesic acoustic mode, possibly coupled with avalanche events, has been linked with particle flow to the wall in diverted plasmas. Detachment, scrape-off layer transport, and turbulence were studied in L- and H-modes in both standard and alternative configurations (snowflake, super-X, and beyond). The detachment process is caused by power ‘starvation’ reducing the ionization source, with volume recombination playing only a minor role. Partial detachment in the H-mode is obtained with impurity seeding and has shown little dependence on flux expansion in standard single-null geometry. In the attached L-mode phase, increasing the outer connection length reduces the in–out heat-flow asymmetry. A doublet plasma, featuring an internal X-point, was achieved successfully, and a transport barrier was observed in the mantle just outside the internal separatrix. In the near future variable-configuration baffles and possibly divertor pumping will be introduced to investigate the effect of divertor closure on exhaust and performance, and 3.5 MW ECRH and 1 MW neutral beam injection heating will be added

Experimental and synthetic measurements of polarized synchrotron emission from runaway electrons in Alcator C-Mod

This paper presents the first experimental analysis of polarized synchrotron emission from relativistic runaway electrons (REs) in a tokamak plasma. Importantly, we show that the polarization information of synchrotron radiation can be used to diagnose spatially-localized RE pitch angle distributions. Synchrotron-producing REs were generated during low density, Ohmic, diverted plasma discharges in the Alcator C-Mod tokamak. The ten-channel Motional Stark Effect diagnostic was used to measure spatial profiles of the polarization angle and the fraction of detected light that was linearly-polarized. Spatial transitions in the polarization angle of 90 degrees---from horizontal to vertical polarization and vice versa---are observed in experimental data and are well-explained by the gyro-motion of REs and high directionality of synchrotron radiation. Polarized synchrotron emission is modeled with the synthetic diagnostic SOFT; its output Green's (or detector response) functions reveal a critical RE pitch angle at which the polarization angle flips by 90 degrees and the polarization fraction is minimal. Using SOFT, we determine the dominant RE pitch angle which reproduces measured polarization angle and fraction values. The spatiotemporal evolutions of the polarization angle and fraction are explored in detail for one C-Mod discharge. For channels viewing REs near the magnetic axis and flux surfaces q = 1 and 4/3, disagreements between synthetic and experimental signals suggest that the sawtooth instability may be influencing RE dynamics. Furthermore, other sources of pitch angle scattering, not considered in this analysis, could help explain discrepancies between simulation and experiment

LHCD during current ramp experiments on Alcator C-Mod

The lower hybrid current drive (LHCD) system on Alcator C-Mod is capable of sustaining fully non-inductive discharges for multiple current relaxation times (τcr ∼ 200 ms) at line averaged densities in the range of 5x1019 m-3. Some of these non-inductive discharges develop unstable MHD modes that can greatly reduce current drive performance, particularly in discharges with plasma current of 0.5 MA or less [1,2]. Avoiding these unstable MHD modes motivated an experiment to test if the stable current profile shape of a higher current non-inductive discharge could be achieved in a lower current discharge. Starting from a discharge at 0.8 MA, the plasma current was ramped down to 0.5 MA over 200 ms. The surface voltage of the plasma swings negative during the ramp, with the loop voltage reversal impacting the edge fast electron measurements immediately. Little change can be seen during the Ip ramp in the core fast electron measurements, indicating that the loop voltage reversal does not penetrate fully to the magnetic axis on the timescale of the current ramp. The resulting discharge did not exhibit deleterious MHD instabilities, however the existence of this one discharge does not necessarily represent a robust solution to the problem

cfs-energy/cfspopcon: v4.0.0

<p>POPCONs (Plasma OPerating CONtours)</p&gt

Development of an 11-channel multi wavelength imaging diagnostic for divertor plasmas in MAST Upgrade

Divertor detachment and alternative divertor magnetic geometries are predicted to be promising approaches to handle the power exhaust of future fusion devices. In order to understand the detachment process caused by volumetric losses in alternative divertor magnetic geometries, a Multi-Wavelength Imaging (MWI) diagnostic has recently been designed and built for the Mega Amp Spherical Tokamak Upgrade. The MWI diagnostic will simultaneously capture 11 spectrally filtered images of the visible light emitted from divertor plasmas and provide crucial knowledge for the interpretation of observations and modeling efforts. This paper presents the optical design, mechanical design, hardware, and test results of an 11-channel MWI system with a field of view of 40°. The optical design shows better than 5 mm FWHM spatial resolution at the plasma on all 11 channels across the whole field of view. The spread of angle of incidence on the surface of each filter is also analyzed to inform the bandwidth specification of the interference filters. The results of the initial laboratory tests demonstrate that a spatial resolution of better than 5 mm FWHM is achieved for all 11 channels, meeting the specifications required for accurate tomography

Development of an 11-channel multi wavelength imaging diagnostic for divertor plasmas in MAST Upgrade

Divertor detachment and alternative divertor magnetic geometries are predicted to be promising approaches to handle the power exhaust of future fusion devices. In order to understand the detachment process caused by volumetric losses in alternative divertor magnetic geometries, a Multi-Wavelength Imaging (MWI) diagnostic has recently been designed and built for the Mega Amp Spherical Tokamak Upgrade. The MWI diagnostic will simultaneously capture 11 spectrally filtered images of the visible light emitted from divertor plasmas and provide crucial knowledge for the interpretation of observations and modeling efforts. This paper presents the optical design, mechanical design, hardware, and test results of an 11-channel MWI system with a field of view of 40°. The optical design shows better than 5 mm FWHM spatial resolution at the plasma on all 11 channels across the whole field of view. The spread of angle of incidence on the surface of each filter is also analyzed to inform the bandwidth specification of the interference filters. The results of the initial laboratory tests demonstrate that a spatial resolution of better than 5 mm FWHM is achieved for all 11 channels, meeting the specifications required for accurate tomography