Circular corrugated waveguides are often used in fusion applications at single, multiple, or broadband millimeter frequencies due to their low ohmic loss, expected large frequency bandwidth, and direct coupling to free-space Gaussian modes. For single-frequency corrugated waveguide systems, transmission line components can be optimized to the desired frequency. For broadband or multiple-frequency applications, this is not possible. The goal of this paper is to demonstrate that the frequency bandwidth of circular corrugated waveguides can be compromised by diffraction losses to miter bends and gaps. It is shown that if the corrugation depth differs significantly from lambda/4, a theory can substantially underpredict the gap and miter bend losses. The simulations are also shown to compare favorably to experimental measurements. To improve the transmission line performance in large frequency bandwidth systems, such as 33-165 GHz reflectometry, reducing the number of miter bends may, therefore, be necessary. For the improvement of performance in narrower frequency bandwidth or multiple-frequency systems, especially high-powered systems, wavelength-dependent techniques may be applicable. One such application is shown for a Bragg reflection technique to reduce the thermal load to insulating ceramic rings in a dc break component for multiple-frequency ITER electron cyclotron heating transmission lines.

A phase 1 dose-escalation trial assessed the chemotherapeutic potential of ammonium perfluorooctanoate (APFO). Forty-nine primarily solid-tumor cancer patients who failed standard therapy received weekly APFO doses (50-1200 mg) for 6 weeks. Clinical chemistries and plasma PFOA (anionic APFO) were measured predose and weekly thereafter. Several clinical measures including total cholesterol, high-density lipoproteins (HDLs), thyroid stimulating hormone (TSH), and free thyroxine (fT4), relative to PFOA concentrations were examined by: Standard statistical analyses using generalized estimating equations (GEE) and a probabilistic analysis using probability distribution functions (pdf) at various PFOA concentrations; and a 2-compartment pharmacokinetic/pharmacodynamic (PK/PD) model to directly estimate mean changes. Based on the GEE, the average rates of change in total cholesterol and fT4 associated with increasing PFOA were approximately -1.2×10-3 mmol/l/μM and 2.8×10-3 pmol/l/μM, respectively. The PK/PD model predicted more closely the trends observed in the data as well as the pdfs of biomarkers. A decline in total cholesterol was observed, with a clear transition in shape and range of the pdfs, manifested by the maximum value of the Kullback-Leibler (KL) divergence, that occurred at plasma PFOA between 420 and 565 μM (175 000-230 000 ng/ml). High-density lipoprotein was unchanged. An increase in fT4 was observed at a higher PFOA transition point, albeit TSH was unchanged. Our findings are consistent with some animal models and may motivate re-examination of the epidemiologic studies to PFOA at levels several orders of magnitude lower than this study. These observational studies have reported contrary associations, but currently understood biology does not support the existence of such conflicting effects.

The present design concept for the ITER low-field-side reflectometer transmission line (TL) consists of an ∼40 m long, 6.35 cm diameter helically corrugated waveguide (WG) together with ten 90° miter bends. This paper presents an evaluation of the TL performance at low frequencies (33-50 GHz) where the predicted HE11 mode ohmic and mode conversion losses start to increase significantly. Quasi-optical techniques were used to form a near Gaussian beam to efficiently couple radiation in this frequency range into the WG. It was observed that the output beams from the guide remained linearly polarized with cross-polarization power levels of ∼1.5%-3%. The polarization rotation due to the helical corrugations was in the range ∼1°-3°. The radiated beam power profiles typically show excellent Gaussian propagation characteristics at distances >20 cm from the final exit aperture. The round trip propagation loss was found to be ∼2.5 dB at 50 GHz and ∼6.5 dB at 35 GHz, showing an inverse increase with frequency. This was consistent with updated calculations of miter bend and ohmic losses. At low frequencies (33-50 GHz), the mode purity remained very good at the exit of the waveguide, and the losses are perfectly acceptable for operation in ITER. The primary challenge may come from the future addition of a Gaussian telescope and other filter components within the corrugated guide, which will likely introduce additional perturbations to the beam profile and an increase in mode-conversion loss.

Doppler backscattering system can measure the perpendicular velocity and fluctuation amplitude of the density turbulence with intermediate wavenumber. An eight-channel Doppler backscattering system has been installed in the Experimental Advanced Superconducting Tokamak (EAST), which can probe eight different radial locations simultaneously by launching eight fixed frequencies (55, 57.5, 60, 62.5, 67.5, 70, 72.5, 75 GHz) into plasma. The quasi-optical system consists of circular corrugated waveguide transmission, a fixed parabolic mirror, and a rotatable parabolic mirror which are integrated with quasi-optics front-end of the profile reflectometer inside the vacuum vessel. The incidence angle can be chosen from 5° to 12°, and the wavenumber range is 2-15/cm with the wavenumber resolution Δk/k≤0.21. Ray tracing simulations are used to calculate the scattering locations and the perpendicular wavenumber. The dynamic range of this new eight-channel Doppler backscattering system can be as large as 40 dB in the EAST. In this article, the hardware design, the ray tracing, and the preliminary experimental results in the EAST will be presented.

In this paper we show that the local density gradient in the plasma core depends on the calculated mode-frequency of the most unstable linear mode and reaches a maximum when this frequency is close to zero. Previous theoretical and experimental work on AUG has shown that the ratio of electron to ion temperature, and as such the frequency of the dominant linear gyrokinetic mode, affects the local density gradient close to rho = 0.3 (Fable et al 2010 Plasma Phys. Control. Fusion 52 015007, Angioni et al 2011 Nucl. Fusion 51 023006). On DIII-D we find that by adding electron cyclotron heating, we modify the dominant unstable linear gyrokinetic mode from an ion temperature gradient (ITG) mode to a trapped electron mode (TEM), which means that the frequency of the dominant mode changes sign (from the ion to the electron direction). Local density peaking around mid-radius increases by 50% right around the cross-over between the ITG and TEM regimes. By comparing how the particle flux changes, through the derivative of the electron density, n(e), with respect to time, partial derivative n(e)/partial derivative t, we find that the particle flux also exhibits the same trend versus mode frequency. As a result, we find that the changes in local particle transport are inversely proportional to the changes in electron density, indicating that the changes are driven by a change in thermo-diffusive pinch.

In this paper, we show that the often observed 'density pump-out' with electron cyclotron heating (ECH) (Angioni et al 2004 Nucl. Fusion 44 827, 2009 Plasma Phys. Control. Fusion 51 124017), at low density and/or collisionality is the result of an increase in turbulence drive at the plasma edge (Angioni et al 2004 Nucl. Fusion 44 827, 2005 Phys. Plasmas 12 040701, Mordijck et al 2015 Nucl. Fusion 43 113025). Prior results were limited to comparison of steady-state conditions, before and after the ECH was applied, and thus failed to capture the dynamics of the density pump-out. In this paper, we find, similar to prior results, that when the plasma reaches a new equilibrium after ECH is applied, gyro-kinetic simulations indicate that the plasma has transitioned from the ion temperature gradient (ITG) to a trapped electron mode (TEM) regime around mid-radius. However, this transition from ITG to TEM only occures in the core after 100 ms. The pump-out on the other hand, starts immediately and is strongest around rho similar to 0.8. Linear gyrokinetic simulations with TGLF show that there is an increase in turbulence drive simultaneously with the density pump-out and the doppler backscattering (DBS) measures an instant increase in density fluctuations at the same radial location. On the other hand, around mid-radius the DBS measures no increase in density fluctuations. All these calculations along with experimental measurements show that the density pump-out is not the result of a change in turbulence type (i.e. not caused by a change from ITG to TEM), but the result of a change in turbulence drive (an increase in linear growth rates), which is later followed by the ITG to TEM transition. This highlights the need for studying not just the equilibrium conditions after a transition, but also the time-dependent changes.

The potential of the hybrid scenario (first developed as an advanced inductive scenario for high fluence) as a regime for high-beta, steady-state plasmas is demonstrated on the DIII-D tokamak. These experiments show that the beneficial characteristics of hybrids, namely safety factor >= 1 with low central magnetic shear, high stability limits and excellent confinement, are maintained when strong central current drive (electron cyclotron and neutral beam) is applied to increase the calculated non-inductive fraction to approximate to 100% (approximate to 50% bootstrap current). The best discharges achieve normalized beta of 3.4, IPB98(y,2) confinement factor of 1.4, surface loop voltage of 0.01 V, and nearly equal electron and ion temperatures at low collisionality. A 0D physics model shows that steady-state hybrid operation with Q(fus) similar to 5 is feasible in FDF and ITER. The advantage of the hybrid scenario as an advanced tokamak regime is that the external current drive can be deposited near the plasma axis where the efficiency is high; additionally, good alignment between the current drive and plasma current profiles is not necessary as the poloidal magnetic flux pumping self-organizes the current density profile in hybrids with an m/n = 3/2 tearing mode.

In this paper we show that resonant magnetic perturbations (RMPs) affect the L- to H-mode power threshold. We find that during the L-mode phase, RMPs cause the particle pinch to reverse from traditionally inward to outward. As a result, the density at the plasma edge increases, while the density in the plasma core is reduced. Linear stability calculations indicate that the plasma transitions from an ion temperature gradient (ITG) to trapped electron mode (TEM) regime at the plasma edge. If the applied RMP current is below the threshold for penetration and island formation, we find that the changes in the edge radial electric field are minimal, while the carbon toroidal rotation brakes over the whole minor radius. Once the RMP field penetrates and the screening plasma response dissappears, the spin-up of the toroidal rotation at the plasma edge results in a positive radial electric field inside the separatrix.

A monostatic antenna array arrangement has been designed for the microwave front-end of the ITER low-field-side reflectometer (LFSR) system. This paper presents details of the antenna coupling coefficient analyses performed using GENRAY, a 3-D ray tracing code, to evaluate the plasma height accommodation capability of such an antenna array design. Utilizing modeled data for the plasma equilibrium and profiles for the ITER baseline and half-field scenarios, a design study was performed for measurement locations varying from the plasma edge to inside the top of the pedestal. A front-end antenna configuration is recommended for the ITER LFSR system based on the results of this coupling analysis.

In this paper we show how changes in toroidal rotation, by controlling the injected torque, affect particle transport and confinement. The toroidal rotation is altered using the co- and counter neutral beam injection (NBI) in low collisionality H-mode plasmas on DIII-D (Luxon 2002 Nucl. Fusion 42 614) with dominant electron cyclotron heating (ECH). We find that there is no correlation between the toroidal rotation shear and the inverse density gradient, which is observed on AUG when T-e/T-i is varied using ECH (Angioni et al 2011 Phys. Rev. Lett. 107 215003). In DIII-D, we find that in a discharge with balanced torque injection, the E x B shear is smaller than the linear gyrokinetic growth rate for small k(theta)rho(s) for rho = 0.6-0.85. This results in lower particle confinement. In the co- and counter-injected discharges the E x B shear is larger or close to the linear growth rate at the plasma edge and both configurations have higher particle confinement. In order to measure particle transport, we use a small periodic perturbative gas puff. This gas puff perturbs the density profiles and allows us to extract the perturbed diffusion and inward pinch coefficients. We observe a strong increase in the inward particle pinch in the counter-torque injected plasma. Finally, the calculated quasi-linear particle flux, nor the linear growth rates using TGLF (Staebler et al 2005 Phys. Plasmas 12 102508) agree with experimental observations.

DIII-D has made significant progress in developing the techniques required to operate ITER, and in understanding their impact on performance when integrated into operational scenarios at ITER-relevant parameters. Long duration plasmas, stable to m/n = 2/1 tearing modes, with an ITER-similar shape and I-p/aB(T), have been demonstrated in DIII-D, that evolve to stationary conditions. The operating region most likely to reach stable conditions has normalized pressure, beta(N) approximate to 1.9-2.1 (compared to the ITER baseline design of 1.6-1.8), and a Greenwald normalized density fraction, f(GW) 0.42-0.70 (the ITER design is f(GW) approximate to 0.8). The evolution of the current profile, using internal inductance (l(i)) as an indicator, is found to produce a smaller fraction of stable pulses when li is increased above approximate to 1.1 at the beginning of beta(N) flattop. Stable discharges with co-neutral beam injection are generally accompanied with a benign n = 2 magnetohydrodynamic mode. However if this mode exceeds approximate to 10 G, the onset of a m/n = 2/1 tearing mode occurs with a loss of confinement. In addition, stable operation with low applied external torque, at or below the extrapolated value expected for ITER has also been demonstrated. With electron cyclotron injection, the operating region of stable discharges has been further extended at ITER equivalent levels of torque and to edge-localized mode (ELM) free discharges at higher torque but with the addition of an n = 3 magnetic perturbation from the DIII-D internal coil set. The characterization of the ITER baseline scenario evolution for long pulse duration, extension to more ITER-relevant values of torque and electron heating, and suppression of ELMs have significantly advanced the physics basis of this scenario, although significant effort remains in the simultaneous integration of all these requirements.

In this paper we show that changing from an ion temperature gradient (ITG) to a trapped electron mode (TEM) dominant turbulence regime (based on linear gyrokinetic simulations) results experimentally in a strong density pump-out (defined as a reduction in line-averaged density) in low collisionality, low power H-mode plasmas. We vary the turbulence drive by changing the heating from predominantly ion heated using neutral beam injection to electron heated using electron cyclotron heating, which changes the T-e/T-i ratio and the temperature gradients. Perturbed gas puff experiments show an increase in transport outside rho = 0.6, through a strong increase in the perturbed diffusion coefficient and a decrease in the inward pinch. Linear gyrokinetic simulations with TGLF show an increase in the particle flux outside the mid-radius. In conjunction an increase in intermediate-scale length density fluctuations is observed, which indicates an increase in turbulence intensity at typical TEM wavelengths. However, although the experimental changes in particle transport agree with a change from ITG to TEM turbulence regimes, we do not observe a reduction in the core rotation at mid-radius, nor a rotation reversal.

Experiments in the DIII-D tokamak have demonstrated potential new paths to fusion steady state based on peaked current profiles with qmin≈1 that have high ideal stability limits, excellent confinement and benefit from efficient on-axis current drive. In the hybrid scenario, steady-state conditions (Vsurf=0) using central ECCD and NBCD are achieved in 1.0 MA discharges with a beta value (βN=3.6) that is 80%-90% of the ideal n=1 with-wall limit. Interestingly, the hybrid mechanism that anomalously broadens the current profile to maintain qmin>1 and prevents sawteeth continues to function despite the intense central current drive. In the “high li” scenario, the combination of broad pressure profile and a peaked current profile tailored to maximize li allows βN≈4.8 and H98y2≈1.8 to be achieved transiently in a discharge that is overdriven (Vsurf<0). The achieved βN is near the no-wall limit with the idealwall limit higher at βN≈5-6.