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Plasma discharges with a negative triangularity (δ=-0.4) shape have been created in the DIII-D tokamak with a significant normalized beta (β_{N}=2.7) and confinement characteristic of the high confinement mode (H_{98y2}=1.2) despite the absence of an edge pressure pedestal and no edge localized modes (ELMs). These inner-wall-limited plasmas have a similar global performance as a positive triangularity (δ=+0.4) ELMing H-mode discharge with the same plasma current, elongation and cross sectional area. For cases both of dominant electron cyclotron heating with T_{e}/T_{i}>1 and dominant neutral beam injection heating with T_{e}/T_{i}=1, turbulent fluctuations over radii 0.5

In this paper, we address the challenging question of measuring turbulence levels on the high magnetic field side (HFS) of tokamak plasmas. Although turbulence measurements on the HFS can provide a stringent constraint for the turbulence model validation, to date only low magnetic field side (LFS) measured turbulence has been used in validation studies. To address this issue, an eight channel Correlation Electron Cyclotron Emission (CECE) system at DIII-D was modified to probe both LFS and HFS. In contrast to the second harmonic extraordinary mode electron cyclotron resonance emission that is typically used in CECE, we show that it is possible to probe the HFS using fundamental O-mode electron cyclotron resonance emission. The required hardware modifications for the HFS measurements are presented here, and the potential issues in this measurement are discussed.

A new eight-channel correlation electron cyclotron emission diagnostic has recently been installed on the DIII-D tokamak to study both turbulent and coherent electron temperature fluctuations under various plasma conditions and locations. This unique system is designed to cover a broad range of operation space on DIII-D (1.6-2.1 T, detection frequency: 72-108 GHz) via four remotely selected local oscillators (80, 88, 96, and 104 GHz). Eight radial locations are measured simultaneously in a single discharge covering as much as half the minor radius. In this paper, we present design details of the quasi-optical system, the receiver, as well as representative data illustrating operation of the system.

A new spherical torus called VEST (Versatile Experiment Spherical Torus) is designed, constructed and successfully commissioned at Seoul National University. A unique design feature of the VEST is two partial solenoid coils installed at both vertical ends of a center stack, which can provide sufficient magnetic fluxes to initiate tokamak plasmas while keeping a low aspect ratio configuration in the central region. According to initial double null merging start-up scenario using the partial solenoid coils, appropriate power supplies for driving a toroidal field coil, outer poloidal field coils, and the partial solenoid coils are fabricated and successfully commissioned. For reliable start-up, a pre-ionization system with two cost-effective homemade magnetron power supplies is also prepared. In addition, magnetic and spectroscopic diagnostics with appropriate data acquisition and control systems are well prepared for initial operation of the device. The VEST is ready for tokamak plasma operation by completing and commissioning most of the designed components.

The affordable, robust, compact (ARC) reactor is the product of a conceptual design study aimed at reducing the size, cost, and complexity of a combined fusion nuclear science facility (FNSF) and demonstration fusion Pilot power plant. ARC is a ∼200–250 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has rare earth barium copper oxide (REBCO) superconducting toroidal field coils, which have joints to enable disassembly. This allows the vacuum vessel to be replaced quickly, mitigating first wall survivability concerns, and permits a single device to test many vacuum vessel designs and divertor materials. The design point has a plasma fusion gain of Q[subscript p] ≈ 13.6, yet is fully non-inductive, with a modest bootstrap fraction of only ∼63%. Thus ARC offers a high power gain with relatively large external control of the current profile. This highly attractive combination is enabled by the ∼23 T peak field on coil achievable with newly available REBCO superconductor technology. External current drive is provided by two innovative inboard RF launchers using 25 MW of lower hybrid and 13.6 MW of ion cyclotron fast wave power. The resulting efficient current drive provides a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing fluorine lithium beryllium (FLiBe) molten salt. The liquid blanket is low-risk technology and provides effective neutron moderation and shielding, excellent heat removal, and a tritium breeding ratio ≥ 1.1. The large temperature range over which FLiBe is liquid permits an output blanket temperature of 900 K, single phase fluid cooling, and a high efficiency helium Brayton cycle, which allows for net electricity generation when operating ARC as a Pilot power plant., United States. Department of Energy (Grant DE-FG02-94ER54235), United States. Department of Energy (Grant DE-SC008435), United States. Department of Energy. Office of Fusion Energy Sciences (Grant DE-FC02-93ER54186), National Science Foundation (U.S.) (Grant 1122374)

The affordable, robust, compact (ARC) reactor conceptual design study aims to reduce the size, cost, and complexity of a combined fusion nuclear science facility (FNSF) and demonstration fusion pilot power plant. ARC is a 200 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC is designed to use rare earth barium copper oxide (REBCO), a type of high-temperature superconductor (HTS), for its toroidal field coils. The use of HTS technology offers many advantages over traditional superconductors when applied to tokamak designs. REBCO superconductors in particular have orders of magnitude higher critical current density than traditional superconductors such as Nb3Sn at local fields greater than 20 T, enabling much higher fields to be used in the tokamak. The large allowable temperature range (up to ∼90 K) of HTS allows the use of coolants other than helium and makes possible the design of joints in the toroidal field coils. This allows the vacuum vessel to be replaced quickly, lowering first wall survivability concerns and reducing the cost and operational implications of vessel failure during the experimental phase of the reactor. External current drive for ARC is provided by two inboard (high-field side) RF launchers using 25 MW of lower hybrid and 13.6 MW of ion cyclotron fast wave power. The resulting efficient current drive provides a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing fluorine lithium beryllium (FLiBe) molten salt. The liquid blanket is low-risk technology and provides effective neutron moderation and shielding, excellent heat removal, and a tritium breeding ratio ≥ 1.1. The large temperature range over which FLiBe is liquid permits blanket operation at ∼900 K with single phase fluid cooling and a high-efficiency Brayton cycle, allowing for net electricity generation when operating ARC as a pilot power plant. When coupled with a demountable compact reactor design, the immersion blanket allows the vacuum vessel to be a replaceable component, eliminating the need for complex sector maintenance. The modular design of ARC allows a single machine to initially serve as an experiment and then transition to a demonstration commercial reactor.

New results on the I-mode regime of operation on the Alcator C-Mod tokamak are reported. This ELM-free regime features high energy confinement and a steep temperature pedestal, while particle confinement remains at L-mode levels, giving stationary density and avoiding impurity accumulation. I-mode has now been obtained over nearly all of the magnetic fields and currents possible in this high field tokamak (I-p 0.55-1.7 MA, B-T 2.8-8 T) using a configuration with B x del B drift away from the X-point. Results at 8 T confirm that the L-I power threshold varies only weakly with B-T, and that the power range for I-mode increases with B-T; no 8 T discharges transitioned to H-mode. Parameter dependences of energy confinement are investigated. Core transport simulations are giving insight into the observed turbulence reduction, profile stiffness and confinement improvement. Pedestal models explain the observed stability to ELMs, and can simulate the observed weakly coherent mode. Conditions for I-H transitions have complex dependences on density as well as power. I-modes have now been maintained in near-DN configurations, leading to improved divertor power flux sharing. Prospects for I-mode on future fusion devices such as ITER and ARC are encouraging. Further experiments on other tokamaks are needed to improve confidence in extrapolation.

We present the first localized measurements of long and intermediate wavelength turbulent density fluctuations ( n) and long wavelength turbulent electron temperature fluctuations ( Te) modified by m/n=2/1 Neoclassical Tearing Mode (NTM) islands (m and n are the poloidal and toroidal mode numbers, respectively). These long and intermediate wavelengths correspond to the expected Ion Temperature Gradient and Trapped Electron Mode scales, respectively. Two regimes have been observed when tracking n during NTM evolution: (1) small islands are characterized by a steep Te radial profile and turbulence levels comparable to those of the background; (2) large islands have a flat Te profile and reduced turbulence level at the O-point. Radially outside the large island, the Te profile is steeper and the turbulence level increased compared to the no or small island case. Reduced turbulence at the O-point compared to the X-point leads to a 15% modulation of n2 across the island that is nearly in phase with the Te ...

Negative magnetic shear has been demonstrated in DIII-D and JT-60U to mitigate the confinement degradation typically observed with increasing the electron to ion temperature ratio (T e/T i). In recent experiments in DIII-D negative central magnetic shear (NCS) discharges, the thermal transport in the internal transport barrier formed around the radius of the minimum safety factor (q min) remained almost constant and modestly increased in the region outside of q min compared to the positive shear (PS) case, when T e/T i increased from about 0.8 to 1.1 through electron cyclotron heating (ECH). The benefit of NCS extending into the region outside of q min can be explained by the lower magnetic shear in the NCS plasma over the plasma radius relative to the PS plasma. Reduced confinement degradation at high T e/T i with NCS plasmas was commonly observed in DIII-D and JT-60U. The mechanism of the different transport responses between the NCS and PS plasmas has been assessed in terms of fluctuation measurements and gyrokinetic simulations in DIII-D; NCS gave a smaller rise in the low-wavenumber broadband turbulent fluctuations with the increase in T e/T i compared with the PS case. This is consistent with gyrokinetic simulations, which show a smaller rise in the growth rates of the ion temperature gradient mode in the NCS plasmas, with increasing T e/T i. Gyrokinetic simulations also showed a change in the stability of the electron modes with ECH applied, consistent with higher-wavenumber fluctuation measurements, although more detailed simulations are needed to give a quantitative explanation for the experimental observations. Control of q-profile and magnetic shear will allow confinement improvement in future machines with dominant electron heating.

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental observation of ion cyclotron range of frequency (ICRF) mode-conversion, ICRF flow drive, demonstration of lower-hybrid current drive at ITER-like densities and fields and, using a set of novel diagnostics, extensive validation of advanced RF codes. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. A summary of important achievements and discoveries are included., United States. Dept. of Energy (Cooperative Agreement DE-FC02-99ER54512), United States. Dept. of Energy (Cooperative Agreement DE-FG03-94ER-54241), United States. Dept. of Energy (Cooperative Agreement DE-AC02-78ET- 51013), United States. Dept. of Energy (Cooperative Agreement DE-AC02-09CH11466), United States. Dept. of Energy (Cooperative Agreement DE-FG02-95ER54309), United States. Dept. of Energy (Cooperative Agreement DE-AC02-05CH11231), United States. Dept. of Energy (Cooperative Agreement DE-AC52-07NA27344), United States. Dept. of Energy (Cooperative Agreement DE-FG02- 97ER54392), United States. Dept. of Energy (Cooperative Agreement DE-SC00-02060)

A tunable correlation electron cyclotron (CECE) system was recently installed on the Alcator C-Mod tokamak to provide local, quantitative measurement of electron temperature fluctuations in the tokamak core. This system represents a significant upgrade from the original CECE system, expanding the measurement capabilities from 4 to 8 total channels, including 2 remotely tunable YIG filters (6-18 GHz; 200 MHz bandwidth). Additional upgrades were made to the optical system to provide enhanced poloidal resolution and allow for measurement of turbulent fluctuations below kθρs < 0.3. These expanded capabilities allow for single shot measurement of partial temperature fluctuation profiles in the region ρ = 0.7 - 0.9 (square root of normalized toroidal flux) in a wide variety of plasma conditions. These measurements are currently being used to provide stringent tests of the gyrokinetic model in ongoing model validation efforts. Details of the hardware upgrades, turbulent fluctuation measurements, and ongoing comparisons with simulations are presented.

Recent research on the Alcator C-Mod tokamak has focused on a range of scientific issues with particular emphasis on ITER needs and on detailed comparisons between experimental measurements and predictive models. Research on ICRF (ion cyclotron range of frequencies) heating emphasized the origins and mitigation of metallic impurities while work on lower hybrid current drive experiments have focused on linear and nonlinear wave interactions that limit efficiency at high densities in regimes with low single pass absorption. Experiments in core turbulence and transport focused on quantitative, multi-field comparisons between nonlinear gyro-kinetics simulations and experimental measurements of profiles, fluxes and fluctuations. Experiments into self-generated rotation observed spontaneous flow reversal at a critical density identical to the transition density between linear ohmic confinement and saturated ohmic confinement regimes. H-mode studies have measured pedestal widths consistent with kinetic-ballooning-mode-like instabilities, while the pedestal heights quantitatively match the EPED code predictions. Experiments with I-mode have increased the operating window for this promising edge-localized-mode-free regime. Extrapolation of I-mode to ITER suggests that the fusion gain Q ~ 10 could be possible in ITER. Investigations into the physics and scaling of the power exhaust channel width in attached enhanced D-alpha H-mode and L-mode plasma showed a direct connection between the midplane pressure-folding length and the outer divertor target footprint. The width was found to scale inversely with IP, while being independent of conducted power, BT or q95 and insensitive to the scrape-off layer connection length - a behaviour that suggests critical-gradient physics sets both pressure and heat-flux profiles.

The first measurements of long wavelength (ky?s?0.3) electron temperature fluctuations in Alcator C-Mod made with a new correlation electron cyclotron emission diagnostic support a long-standing hypothesis regarding the confinement transition from linear ohmic confinement (LOC) to saturated ohmic confinement (SOC). Electron temperature fluctuations decrease significantly (?40%) crossing from LOC to SOC, consistent with a change from trapped electron mode (TEM) turbulence domination to ion temperature gradient (ITG) turbulence as the density is increased. Linear stability analysis performed with the GYRO code (Candy and Waltz 2003 J. Comput. Phys. 186 545) shows that TEMs are dominant for long wavelength turbulence in the LOC regime and ITG modes are dominant in the SOC regime at the radial location (????0.8) where the changes in electron temperature fluctuations are measured. In contrast, deeper in the core (??

A correlation electron cyclotron emission (CECE) diagnostic has been installed in Alcator C-Mod. In order to measure electron temperature fluctuations, this diagnostic uses a spectral decorrelation technique. Constraints obtained with nonlinear gyrokinetic simulations guided the design of the optical system and receiver. The CECE diagnostic is designed to measure temperature fluctuations which have k(θ) ≤ 4.8 cm(-1) (k(θ)ρ(s)0.5) using a well-focused beam pattern. Because the CECE diagnostic is a dedicated turbulence diagnostic, the optical system is also flexible, which allows for various collimating lenses and antenna to be used. The system overview and the demonstration of its operability as designed are presented in this paper.

For the first time, nonlinear gyrokinetic simulations of I-mode plasmas are performed and compared with experiment. I-mode is a high confinement regime, featuring energy confinement similar to H-mode, but without enhanced particle and impurity particle confinement [D. G. Whyte et al., Nucl. Fusion 50, 105005 (2010)]. As a consequence of the separation between heat and particle transport, I-mode exhibits several favorable characteristics compared to H-mode. The nonlinear gyrokinetic code GYRO [J. Candy and R. E. Waltz, J Comput. Phys. 186, 545 (2003)] is used to explore the effects of E x B shear and profile stiffness in I-mode and compare with L-mode. The nonlinear GYRO simulations show that I-mode core ion temperature and electron temperature profiles are more stiff than L-mode core plasmas. Scans of the input E x B shear in GYRO simulations show that E x B shearing of turbulence is a stronger effect in the core of I-mode than L-mode. The nonlinear simulations match the observed reductions in long wavelength density fluctuation levels across the L-I transition but underestimate the reduction of long wavelength electron temperature fluctuation levels. The comparisons between experiment and gyrokinetic simulations for I-mode suggest that increased E x B shearing of turbulence combined with increased profile stiffness are responsible for the reductions in core turbulence observed in the experiment, and that I-mode resembles H-mode plasmas more than L-mode plasmas with regards to marginal stability and temperature profile stiffness. (C) 2015 AIP Publishing LLC.

Long wavelength turbulent electron temperature fluctuations (k(y)rho(s)< 0.3) are measured in the outer core region (r/a> 0.8) of Ohmic L-mode plasmas at Alcator C-Mod [E. S. Marmar et al., Nucl. Fusion 49, 104014 (2009)] with a correlation electron cyclotron emission diagnostic. The relative amplitude and frequency spectrum of the fluctuations are compared quantitatively with nonlinear gyrokinetic simulations using the GYRO code [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] in two different confinement regimes: linear Ohmic confinement (LOC) regime and saturated Ohmic confinement (SOC) regime. When comparing experiment with nonlinear simulations, it is found that local, electrostatic ion-scale simulations (k(y)rho(s) less than or similar to 1.7) performed at r/a similar to 0.85 reproduce the experimental ion heat flux levels, electron temperature fluctuation levels, and frequency spectra within experimental error bars. In contrast, the electron heat flux is robustly under-predicted and cannot be recovered by using scans of the simulation inputs within error bars or by using global simulations. If both the ion heat flux and the measured temperature fluctuations are attributed predominantly to long-wavelength turbulence, then under-prediction of electron heat flux strongly suggests that electron scale turbulence is important for transport in C-Mod Ohmic L-mode discharges. In addition, no evidence is found from linear or nonlinear simulations for a clear transition from trapped electron mode to ion temperature gradient turbulence across the LOC/SOC transition, and also there is no evidence in these Ohmic L-mode plasmas of the "Transport Shortfall" [C. Holland et al., Phys. Plasmas 16, 052301 (2009)]. (c) 2016 AIP Publishing LLC.