Search

This paper reports the significant advancement of our ability to model and to understand how RF waves interact with the SOL plasma, by developing for the full torus 3D scrape-off layer (SOL) plasma simulation together with the antenna and core plasma in the NSTX-U device in HHFW frequency regime for a cold plasma model. The present paper extends the previous modeling (Bertelli et al 2020 AIP Conf. Proc. 2254 030001) to a full 3D device geometry including realistic antenna geometry in order to capture a more realistic experimental situation including the fully 3D wave field effects and the antenna plasma interaction in the SOL plasma, and, at the same time, resulting core wave propagation. The central tool of the present work is the Petra-M code, which is a newly developed state-of-the-art generic electromagnetic simulation tool for modeling RF wave propagation based on MFEM (http://mfem.org), an open source scalable C++ finite element method library.

Non-inductive start-up and ramp-up is an important topic for spherical tokamak reactor design as the central solenoid implementation is highly restrictive particularly for the low-aspect-ratio tokamak configuration. In the high field spherical tokamak (ST), ST-40 with B T0 ⩽ 3 T, a preparation is underway for high power ECH and ECCD current start-up/ramp-up experiments utilizing two MW-class 140/105 GHz gyrotrons. Here, we explored various ECH/ECCD scenarios for a low-field-side (LFS) launch-angle steerable waveguide launcher placed near the mid-plane region. Due to the large toroidal field variation of ST configuration, multiple cyclotron harmonic resonance layers could exist within the plasma. In this start-up and ramp-up regime, both fundamental and second harmonic ECH resonances must be considered. We find that even with the presence of X-II resonance layer in the plasma, an efficient X-I ECH and ECCD regime can be accessed for the low electron temperature T e0 as low as 200 eV which is a typical starting temperature of ECH heated plasmas in an open-field-line configuration. The presence of X-II resonance could become significant at higher T e0 as X-II absorption increases with T e0 which could reduce the current ramp-up efficiency as the power reaching X-I is reduced. Finally for the pure X-I regime where the 2Ωe resonance is moved outside the plasma with B T0 ∼ 3.4 T, we find that it is possible to reach the full current of I p ∼ 1 MA fully non-inductively with the ECH power of ∼1 MW at n e0 ∼ 1.0 × 1019 m−3 using 105 GHz frequency gyrotron. By reducing the outer limiter position R L ∼ 78 cm to 70 cm, the pure X-I regime is recovered at the rated ST-40 magnetic field of B T0 ∼ 3.0 T. This X-I regime is accessible with a relatively broad range of launched n II or the launching angles. A survey of X-mode X-II ECH and ECCD at higher density regimes is also shown for completeness.

International audience; Three identical new WEST Ion Cyclotron Resonance Heating (ICRH) antennas have been designed, assembled then commissioned on plasma from 2013 to 2019. The WEST ICRH system is both load-resilient and compatible with long-pulse operations. The three antennas have been successfully operated together on plasma in 2019 and 2020. The load resilience capability has been demonstrated and the antenna feedback controls for phase and matching have been developed. The breakdown detection systems have been validated and successfully protected the antennas. The use of ICRH in combination with Lower Hybrid has triggered the first high confinement mode transitions identified on WEST.

Fully non-inductive plasma start-up was successfully achieved by using a well-controlled microwave source on the spherical tokamak, QUEST. Non-inductive plasmas were maintained for approximately 3–5 min, during which time power balance estimates could be achieved by monitoring wall and cooling-water temperatures. Approximately 70%–90% of the injected power could be accounted for by calorimetric measurements and approximately half of the injected power was found to be deposited on the vessel wall, which is slightly dependent on the magnetic configuration. The power distribution to water-cooled limiters, which are expected to be exposed to local heat loads, depends significantly on the magnetic configuration, however some of the deposited power is due to energetic electrons, which have large poloidal orbits and are likely to be deposited on the plasma facing components.

A novel, rapid time-response, disruption mitigation system referred to as the electromagnetic particle injector (EPI) is described. This method can accurately deliver the radiative payload to the plasma center on a

Initial results from the recently implemented transient coaxial helicity injection (CHI) system on QUEST are reported. QUEST uses a new CHI electrode configuration in which the CHI insulator is not part of the vacuum boundary, making this configuration easier to implement in fusion reactors. Experimental results show that transient CHI startup in this alternate electrode configuration is indeed possible. Reliable gas breakdown was achieved, and toroidal currents up to 45 kA were generated.

Coaxial Helicity Injection (CHI) has been successfully used in the National Spherical Torus Experiment (NSTX) for a demonstration of closed flux current generation without the use of the central solenoid. The favorable properties of the Spherical Torus (ST) arise from its very small aspect ratio. However, small aspect ratio devices have very restricted space for a substantial central solenoid. Thus methods for initiating the plasma current without relying on induction from a central solenoid are essential for the viability of the ST concept. CHI is a promising candidate for solenoid-free plasma startup in a ST. The method has now produced closed flux current up to 160 kA verifying the high current capability of this method in a large ST built with conventional tokamak components.

Since FY 2004, a four-year Japanese national project has been ongoing to develop an HTS-SMES as a sub-program of the Development of Superconducting Power Network Control Technology. The purpose of this project is to confirm the performance of a high energy density SMES coil with Bi-2212 cables. This coil system is composed of an NbTi outer coil and a Bi-2212 inner coil, consisting of 16 double pancakes. A Bi2212 cable of 140 m length is used for the double-pancake windings, and reinforcing materials are co-wound with the cable. A winding inner diameter of 360 mm was set according to the bending tolerance of the cable. The magnet is to be cooled at 4.2 K by liquid helium in order to obtain high overall current density. In the first year of this project, a conceptual design and various experiments for HTS magnet were carried out. This paper describes the test results of the HTS experimental coil that was developed to evaluate the electrical and thermal characteristics of an HTS. It is important to minimize deterioration and damage to the cable, because such defects cause fatal impact to the coil. Consequentially, high accuracy measurement of V-I characteristics up to10−8 V/cm range was performed in this experiment.

This paper describes a new design concept of SMES coil for bridging instantaneous dips and some experimental results. The technical key point of this work is the design of a NbTi coil composed of a monolith NbTi/Cu wire for DC application. In order to cover the disadvantages of the monolith wire, some solutions have been proposed and tested using a fullscale test coil. Experimental results show the validity of the design concept of the SMES coil.

A spherical torus (ST) fusion energy development path which is complementary to the proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact component test facility (CTF) and high performance advanced regimes leading to more attractive Demo and power plant scale reactors. To provide the physical basis for the CTF an intermediate step needs to be taken, which we refer to as the 'next-step spherical torus' (NSST) device and which we examine in some detail herein. NSST is a 'performance extension' stage ST with a plasma current of 5?10?MA, R = 1.5?m, BT ? 2.6?T and the possibility of varying physical parameters. The mission of NSST is to (1) provide a sufficient physical basis for the design of a CTF; (2) explore advanced operating scenarios with high bootstrap current fraction and high performance which can be utilized by CTF, Demo, and power plants; and (3) contribute to the general science of high ? toroidal plasmas. The NSST is designed to utilize a TFTR-like site to minimize the cost and time required for design and construction.

Global magnetohydrodynamic (MHD) stability limits in the National Spherical Torus Experiment (NSTX) have increased significantly recently due to a combination of device and operational improvements. First, more routine H-mode operation with broadened pressure profiles allows access to higher normalized β and lower internal inductance. Second, the correction of a poloidal field coil induced error-field has largely eliminated locked tearing modes during normal operation and increased the maximum achievable β. As a result of these improvements, peak β values have reached (not simultaneously) βT = 35%, βN = 6.4, βN = 4.5, βN/li = 10, and βP = 1.4. High βP operation with reduced tearing activity has allowed a doubling of discharge pulse-length to just over 1 s with sustained periods of βN≈6 above the ideal no-wall limit and near the with-wall limit. Details of the β-limit scalings and β-limiting instabilities in various operating regimes are described.

The National Spherical Torus Experiment (NSTX) is a low aspect-ratio fusion research facility whose research goal is to make a determination of the attractiveness of the spherical torus concept in the areas of high-β stability, confinement, current drive, and divertor physics. Remarkable progress was made in extending the operational regime of the device in FY 2002. In brief, βt of 34% and βN of 6.5 were achieved. H-mode became the main operational regime, and energy confinement exceeded conventional aspect-ratio tokamak scalings. Heating was demonstrated with the radiofrequency antenna, and signatures of current drive were observed. Current initiation with coaxial helicity injection produced discharges of 400 kA, and first measurements of divertor heat flux profiles in H-mode were made.

Rapidly developing diagnostic, operational, and analysis capability is enabling the first detailed local physics studies to begin in high β plasmas of the National Spherical Torus Experiment (NSTX). These studies are motivated in part by the observation of energy confinement times in neutral-beam-heated discharges that are favourable with respect to predictions from the ITER-89P scaling expression. For plasmas heated with neutral beam injection (NBI), analysis based on profile measurements suggests that electron heat conduction is the dominant thermal loss channel. Cases where early analysis indicates that ion thermal conduction may be exceptionally low is motivating studies of possible sources of ion heating not presently accounted for by classical collisional processes. Gyrokinetic microstability studies indicate that long wavelength turbulence with kθρi~0.1-1 may be stable or suppressed by E×B shear in these plasmas, while modes with kθρi~50 may be robust. High harmonic fast wave (HHFW) heating efficiently heats electrons on NSTX, and studies have begun using it to assess transport in the electron channel. Regarding edge transport, H-mode transitions occur with either NBI or HHFW heating. The power required for L- to H-mode transitions far exceeds that expected from empirical ELM-free H-mode scaling laws derived from moderate aspect ratio devices. Finally, initial fluctuation measurements made with two techniques are permitting the first characterizations of edge turbulence.

Author(s): Menard, JE; Allain, JP; Battaglia, DJ; Bedoya, F; Bell, RE; Belova, E; Berkery, JW; Boyer, MD; Crocker, N; Diallo, A; Ebrahimi, F; Ferraro, N; Fredrickson, E; Frerichs, H; Gerhardt, S; Gorelenkov, N; Guttenfelder, W; Heidbrink, W; Kaita, R; Kaye, SM; Kriete, DM; Kubota, S; Leblanc, BP; Liu, D; Lunsford, R; Mueller, D; Myers, CE; Ono, M; Park, JK; Podesta, M; Raman, R; Reinke, M; Ren, Y; Sabbagh, SA; Schmitz, O; Scotti, F; Sechrest, Y; Skinner, CH; Smith, DR; Soukhanovskii, V; Stoltzfus-Dueck, T; Yuh, H; Wang, Z; Waters, I; Ahn, JW; Andre, R; Barchfeld, R; Beiersdorfer, P; Bertelli, N; Bhattacharjee, A; Brennan, D; Buttery, R; Capece, A; Canal, G; Canik, J; Chang, CS; Darrow, D; Delgado-Aparicio, L; Domier, C; Ethier, S; Evans, T; Ferron, J; Finkenthal, M; Fonck, R; Gan, K; Gates, D; Goumiri, I; Gray, T; Hosea, J; Humphreys, D; Jarboe, T; Jardin, S; Jaworski, MA; Koel, B; Kolemen, E; Ku, S; La Haye, RJ; Levinton, F; Luhmann, N; Maingi, R; Maqueda, R; McKee, G; Meier, E; Myra, J; Perkins, R | Abstract: The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and the NSTX Upgrade (NSTX-U) Project was completed in the summer of 2015. NSTX-U first plasma was subsequently achieved, diagnostic and control systems have been commissioned, the H-mode accessed, magnetic error fields identified and mitigated, and the first physics research campaign carried out. During ten run weeks of operation, NSTX-U surpassed NSTX record pulse-durations and toroidal fields (TF), and high-performance ∼1 MA H-mode plasmas comparable to the best of NSTX have been sustained near and slightly above the n = 1 no-wall stability limit and with H-mode confinement multiplier H98y,2 above 1. Transport and turbulence studies in L-mode plasmas have identified the coexistence of at least two ion-gyro-scale turbulent micro-instabilities near the same radial location but propagating in opposite (i.e. ion and electron diamagnetic) directions. These modes have the characteristics of ion-temperature gradient and micro-tearing modes, respectively, and the role of these modes in contributing to thermal transport is under active investigation. The new second more tangential neutral beam injection was observed to significantly modify the stability of two types of Alfven eigenmodes. Improvements in offline disruption forecasting were made in the areas of identification of rotating MHD modes and other macroscopic instabilities using the disruption event characterization and forecasting code. Lastly, the materials analysis and particle probe was utilized on NSTX-U for the first time and enabled assessments of the correlation between boronized wall conditions and plasma performance. These and other highlights from the first run campaign of NSTX-U are described.

In this contribution we analyze modulation electron cyclotron resonance heating (MECH) experiment and discuss higher harmonic frequency dependence of transport coefficients. We use the bidirectional heat pulse propagation method, in which both inward propagating heat pulse and outward propagating heat pulse are analyzed at a radial range, in order to distinguish frequency dependence of transport coefficients due to hysteresis from that due to other reasons, such as radially dependent transport coefficients, a finite damping term, or boundary effects. The method is applied to MECH experiments performed in various helical and tokamak devices, i.e. Large Helical Device (LHD), TJ-II, Korea Superconducting Tokamak Advanced Research (KSTAR), and Doublet III-D (DIII-D) with different plasma conditions. The frequency dependence of transport coefficients are clearly observed, showing a possibility of existence of transport hysteresis in flux-gradient relation.

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.

Research in NSTX has been conducted to establish spherical torus plasmas to be used for high ?, auxiliary heated experiments. This device has a major radius R0 = 0.86?m and a midplane halfwidth of 0.7?m. It has been operated with toroidal magnetic field B0 ? 0.3?T and Ip ? 1.0?MA. The evolution of the plasma equilibrium is analysed between discharges with an automated version of the EFIT code. Limiter, double null and lower single null diverted configurations have been sustained for several energy confinement times. The plasma stored energy reached 92?kJ (?t = 17.8%) with neutral beam heating. A plasma elongation in the range 1.6 ? ? ? 2.0 and a triangularity in the range 0.25 ? ? ? 0.45 have been sustained, with values of ? = 2.6 and ? = 0.6 being reached transiently. The reconstructed magnetic signals are fitted to the corresponding measured values with low errors. Aspects of the plasma boundary, pressure and safety factor profiles are supported by measurements from non-magnetic diagnostics. Plasma densities have reached 0.8 and 1.2 times the Greenwald limit in deuterium and helium plasmas, respectively, with no clear limit encountered. Instabilities including sawteeth and reconnection events, characterized by Mirnov oscillations, and a perturbation of the Ip, ? and li evolutions, have been observed. A low q limit was observed and is imposed by a low toroidal mode number kink instability.

Coaxial helicity injection (CHI) on the National Spherical Torus Experiment (NSTX) has produced 240 kA of toroidal current without the use of the central solenoid. Values of the current multiplication ratio (CHI produced toroidal current/injector current) up to 10 were obtained, in agreement with predictions. The discharges, which lasted for up to 200 ms, limited only by the programmed waveform, are more than an order of magnitude longer in duration than any CHI discharges previously produced in a spheromak or a spherical torus.

Coaxial helicity injection has been investigated on the National Spherical Torus Experiment (NSTX). Initial experiments produced 130 kA of toroidal current without the use of the central solenoid. The corresponding injector current was 20 kA. Discharges with pulse lengths up to 130 ms have been produced.

The National Spherical Torus Experiment (NSTX) is being built at the Princeton Plasma Physics Laboratory to test the fusion physics principles for the Spherical Torus (ST) concept at the MA level. The NSTX nominal plasma parameters are R {sub 0} = 85 cm, a = 67 cm, R/a greater than or equal to 1.26, B {sub T} = 3 kG, I {sub p} = 1 MA, q {sub 95} = 14, elongation {kappa} less than or equal to 2.2, triangularity {delta} less than or equal to 0.5, and plasma pulse length of up to 5 sec. The plasma heating/current drive (CD) tools are High Harmonic Fast Wave (HHFW) (6 MW, 5 sec), Neutral Beam Injection (NBI) (5 MW, 80 keV, 5 sec), and Coaxial Helicity Injection (CHI). Theoretical calculations predict that NSTX should provide exciting possibilities for exploring a number of important new physics regimes including very high plasma beta, naturally high plasma elongation, high bootstrap current fraction, absolute magnetic well, and high pressure driven sheared flow. In addition, the NSTX program plans to explore fully noninductive plasma start-up, as well as a dispersive scrape-off layer for heat and particle flux handling.

In fast wave to ion Bernstein wave mode conversion experiments in DT supershot plasmas, localized efficient ion heating rather than electron heating was observed, which was due to Doppler broadened tritium cyclotron resonance overlap into the mode conversion region. The ion temperature heat pulse associated with RF power modulation in this regime could provide a diagnostic tool for measuring the local ion thermal conductivity in various confinement regimes. In direct launch ion Bernstein wave heating experiments, core power coupling was limited by the excitation of parasitic edge modes. However, a sheared poloidal flow was observed that is consistent in both magnitude and direction with theoretical models based on RF driven Reynolds stress. With the modest power coupled to the core (~360 kW), the magnitude of the shear in the observed flow was estimated to be a factor of 3-4 too low to trigger transport barrier formation through localized shear suppression of turbulence.

If the ion Bernstein wave (IBW) heating power in an H mode discharge of the PBX-M experiment exceeds a threshold power of about 200 kW, a core transport barrier is created in the central region of the plasma. At lower neutral beam injection (NBI) powers, the core barrier is accompanied by an edge L mode. The high edge localized mode (ELM) repetition frequency (1 kHz) prevents the creation of a strong barrier, so the edge first has to make an H-to-L transition before a strong core transport barrier can be created. At higher NBI powers, the ELM repetition frequency is lowered to less than 200 Hz, which allows the immediate creation of a strong core barrier. Edge localized mode loss, which propagates radially first on a fast (non-diffusive) and then on a slow (diffusive) time-scale all the way to the plasma core, is strongly reduced in the core barrier region. Correlated with the reduced ELM loss, the fluctuations in the core barrier region are also strongly reduced, both during the ELM and during the quiet periods between the ELMs. There is strong evidence that the IBW induced poloidal flow shear is responsible for the stabilization of core turbulence and the creation of the core transport barrier. The large perpendicular E ? B flow shear component of the measured toroidal velocity in co-injection neutral beam heated discharges seems to be largely cancelled by the ion diamagnetic drift shear produced by large ion pressure gradients in the core barrier region. The value of IBW induced poloidal flow has not been experimentally determined, but its numerical value is found to be a factor of 4 larger than either the toroidal velocity or the ion diamagnetic drift shear components, leaving only IBW induced flow shear as the most probable cause for the turbulence stabilization. The core turbulence suppression and the creation of the core transport barrier is also consistent with expectations from a comparison between the E ? B flow shear rate and a rough estimate of the linear ion temperature gradient (ITG) growth rate. The presence of the core barrier region also strongly modifies the other MHD events: crashes on the q = 1.5, 2 surfaces and the disruption.

Two new types of high frequency (HF) MHD mode activity are observed when a core transport barrier is formed during ion Bernstein wave (IBW) heating in PBX-M. These modes, localized in the core transport barrier, seem to be mainly responsible for the soft saturation in the strength of the core transport barrier. The two types of HF mode activity are determined by the presence or absence of a strong 1/1 mode. The n = m modes are observed with the 1/1 mode, while the n = m - 1 modes are seen when the 1/1 mode is absent. The m numbers of both of these modes vary between 4 and 13. They all seem to be coupled and rotate as a rigid body. The n = m - 1 modes seemed to be narrowly localized on their corresponding rational surface, i.e. q = m/(m - 1). The modes are ballooning-like, and their amplitudes increase with increasing β. Calculations with the CASTOR code show that the n = m - 1 modes are most probably pressure driven tearing modes, but the possibility of kinetic ballooning modes (KBMs) cannot be fully excluded. The n = m modes appear on or around the q = 1 surface, and are localized at the X point of the 1/1 mode island. The frequency offset of the n = m modes, which is too low for a TAE, is shown to be consistent with β-induced toroidal Alfven eigenmodes (BAEs).

The physics of current profile modification with lower hybrid waves has been investigated in the Princeton Beta Experiment-Modification tokamak by comparing experimental internal magnetic field profiles with ray tracing code predictions. When the n| spectrum of the launched waves was varied, local changes in the current profile were observed according to equilibria, reconstructed from motional Stark effect polarimetry measurements. Changes in the central safety factor (q) were also determined to be a function of the applied radiofrequency (RF) power. Calculations of the non-inductive current with the Tokamak Simulation Code/Lower Hybrid Simulation Code were unable to duplicate the measured current profiles. Instead, a current broadening mechanism like a finite fast electron diffusion coefficient had to be hypothesized to fit the data. This study was possible only with recent improvements in diagnostics and numerical analysis tools, and it constitutes an important contribution to our understanding of non-inductive current drive for future fusion reactors

A theoretical study has been performed to examine the possibility of generating a radial electric field at the periphery of a tokamak plasma. A localized magnetic well can be produced by a ripple coil system and electron cyclotron resonance heating (ECRH) is used to increase the ripple trapped fraction of the electrons. The magnetic field ripple is arranged in such a way that the ripple trapped electrons then drift vertically, due to Del B, towards the plasma centre. A semi-analytic kinetic analysis has been performed to estimate the ECRH-induced temperature anisotropy of the electrons, which is directly related to the ripple trapping fraction; and a numerical Monte Carlo guiding centre simulation has been performed to study the generation of the radial electric field by the Del B injection of the ripple trapped electrons and the E*B detrapping of them. The present study suggests that about 23 kW of ECRH power in the CDX-U device and 10 MW in ITER may be able to create a large enough radial electric field for the H mode transition

Experimental results obtained in PBX-M suggest that IBW can be used to induce locally a transport barrier and improve core confinement. Using similar neoclassical-like particle and ion energy diffusivity values as inferred inside the IBW-induced core barrier, the CH-mode performance is assessed for TFTR high performance D - T plasmas. By controlling the barrier location, Q-factors in the range can be obtained with a total stored energy of MJ.

In this paper we examine the evolution of the gradients in toroidal velocity, , and ion temperature, , during the CH-mode in PBX-M. This enhanced confinement mode develops from the H-mode following the application of IBW power and is characterized by strong density peaking and the formation of significant gradients in both toroidal velocity and ion temperature. The radial profiles of both the toroidal velocity and the ion temperature are obtained from the PBX-M charge exchange recombination spectroscopy (CHERS) diagnostic. A significant IBW power threshold is observed in the formation of the gradients. A power step of only 25 kW - from 200 to 225 kW - marks the onset. Details of the evolution of the gradients are given and, in particular, the variation with IBW power is described.

An addition of ion Bernstein wave power of only a tenth of the neutral beam power into an H-mode causes the formation of a core transport barrier. At the location of the barrier, reduced ELM losses in the soft x-ray profile are observed. During the IBW-modified ELM, one can observe enhanced fluctuations in the soft x-ray fluctuation profile over the whole of the observed frequency range. In the region of the core transport barrier, the enhancement of fluctuations seems to be strongly reduced. The ELM loss propagates to the core with a velocity greater than that which would be consistent with a normal energy diffusion process.

A reflectometer, based on the amplitude modulation technique, has been developed and operated on a PBX-M for the measurement of electron density profile. The system, operating with the extraordinary mode in the range of 32-50 GHz, is able to measure the density profile in the plasma edge from the scrape-off layer up to typically r/a=0.7. The determination of the time delay for each frequency is achieved by measuring the phase delay of a 200 MHz amplitude modulation envelope superimposed on the millimetre wave probing signal. The system has a final bandwidth of 40 kHz and is able to obtain the edge profile during a 1 ms sweep of the microwave source. High quality profiles are obtained in systematic good agreement with Thomson scattering measurements. The profile reconstruction from the raw data is direct, with only a need for minimal data processing. Profiles have been measured for ohmic, RF and NBI heated discharges. Features of the profile changes in the kH transition are shown. One of the goals of the instrument has been the measurement of the slight modifications to the edge density profile produced by the injection of ion Bernstein waves. These changes have been clearly observed and are in agreement with theoretical expectations.

A stationary, quiescent H-mode (QH-mode) regime with a wide pedestal and improved confinement at low rotation has been discovered on DIII-D with reactor relevant edge parameters and no ELMs. As the injected neutral beam torque is ramped down and the edge E × B rotation shear reduces, the transition from standard QH to the wide pedestal QH-mode occurs. At the transition, the coherent edge harmonic oscillations (EHO) that usually regulate the standard QH edge cease and broadband edge MHD modes appear along with a rapid increase in the pedestal pressure height (by ⩽60%) and width (by ⩽50%). We posit that the enhanced edge turbulence-driven transport, enabled by the lower edge E × B flow shear due to lower torque reduces the pedestal gradient and, combined with the high edge instability limit provided by the balanced double-null plasma shape, permits the development of a broader and thus higher pedestal that is turbulence-transport-limited. Even with the significantly enhanced pedestal pressure, the edge operating point is below the peeling ballooning mode stability boundary and thus without ELMs. Improved transport in the outer core region (0.8 ⩽ ρ ⩽0.9) owing to increased E × B flow shear in that region and the enhanced pedestal boost the overall confinement by up to 45%. These findings advance the physics basis for developing stationary ELM-free high-confinement operation at low rotation for future burning plasma where similar collisionality and rotation levels are expected.

The fourth International Symposium on Liquid Metal Application for Fusion Devices (ISLA-2015) was held on 28–30 September 2015 at Granada, Spain, with growing participation and interest from the community working on general aspects of liquid metal research for fusion energy development. The ISLA symposia remain the largest, and arguably, the most important meetings dedicated to liquid metal application for the magnetic fusion research. Overall, 43 presentations plus 7 posters were given, representing 28 institutions from 12 countries. The latest experimental results from 9 magnetic fusion devices were given in 17 presentations from NSTX and LTX (PPPL, USA), FTU (ENEA, Italy), T-11M (Trinity, RF), T-10 (Kurchatov Institute, RF), TJ-II (CIEMAT, Spain), EAST (ASIPP, China), HT-7 (ASIPP, China), DIII-D (GA, USA), ISTTOK (IPFN, Portugal) and KTM (NNC RK, Kazakhstan). Sessions were devoted to the following: (I) liquid metals (LM) in magnetic confinement experiments (facility overviews), (II) LM in magnetic confinement experiments (topical issues), (III) laboratory experiments, (IV) LM tests in linear plasma devices, (V) LM theory/modeling (VI) LM technology and (VII) a special session on lithium-safety and lithium handling. There were contributions from fusion technology communities including IFMIF and TBM, which provided productive exchanges with physics-oriented magnetic confinement liquid metal research groups. This international workshop will continue on a biennial basis (alternating with the Plasma–Surface Interactions (PSI) Conference), with the next workshop scheduled for Moscow, Russian Federation, in 2017.

Ion Bernstein wave heating (IBWH) is used in the PBX-M tokamak to investigate active control of the pressure profile. A series of experiments was carried out to study impurity behaviour during IBWH under various discharge conditions. In all cases, impurity concentration increased and impurity transport was affected by application of IBWH. Core particle confinement was found to rise during IBWH, which resulted in an increase in the inward convective flow of impurities. Real time boronization of the vacuum vessel was successful in decreasing the overall impurity concentration

Parametric instabilities during the ion Bernstein wave (IBW) plasma heating experiment on PBX-M are investigated both theoretically and experimentally. The theoretical work shows that the RF power threshold of the instabilities is very low when the plasma density near the antenna satisfies the condition 2ωpi >1 MW, and is determined by convective losses due to plasma inhomogeneity when the plasma density near the antenna is sufficiently high that 0.5 < ωpi/ω0 < 1. According to the theory, the parametric instability activity should increase as the plasma is moved away from the antenna, creating a low density electron plasma wave 'gap' region. To test this hypothesis on PBX-M, the plasma position was deliberately varied while monitoring this activity. Under normal IBW operating conditions, very little parametric instability activity was observed ≤50 dB below the pump wave (ω = ωRF). However, when the plasma edge was moved away from the antenna by about 2 cm, the parametric instability activity increased greatly, exceeding 20 dB of the pump wave. This result shows that the observed parametric instability activity can be explained in terms of the plasma inhomogeneity convective model. It was also demonstrated that, by controlling the plasma position with respect to the antenna, parametric instability activity can be controlled at a low level

High frequency pressure driven modes have been observed in high poloidal beta discharges in the Princeton Beta Experiment Modification (PBX-M). These modes are excited in a non-axisymmetric equilibrium characterized by a large, low frequency mt = 1/nt = 1 island, and they are capable of expelling fast ions. The modes reside on or very close to the q = 1 surface and have mode numbers with either mh = nh or (less probably) mh/nh = mh/(mh-1), with mh varying between 3 and 10. Occasionally these modes are simultaneously localized in the vicinity of the ml = 2/nl = 1 island. The high frequency modes near the q = 1 surface also exhibit a ballooning character, being significantly stronger on the large major radius side of the plasma. When a large mt = 1/nt = 1 island is present, the mode is poloidally localized in the immediate vicinity of the X point of the island. The modes occur exclusively in high beta beam heated discharges and are likely to be driven by the beam ions. They can thus be a manifestation of either a toroidicity induced shear Alfven eigenmode (TAE) at q = (2mh+1)/2nh, a kinetic ballooning mode, or some other type of pressure driven (high β) mode. Most of the data are consistent with the theoretical predictions for the TAE gap mode. Since the high frequency modes in PBX-M, however, are found exclusively on or in the immediate neighbourhood of magnetic surfaces with low rational numbers (q = 1, 2,...), other possibilities are not excluded

An experiment in a new regime of ion Bernstein wave (IBW) heating was carried out using 130 MHz high power transmitters in the JIPP T-IIU tokamak. The heating regime utilized the IBW branch between the 3rd and 4th harmonics of the hydrogen ion cyclotron frequencies. This harmonic number is the highest one used in IBW experiments conducted previously. The net radiofrequency (RF) power injected into the plasma is around 400 kW and is limited by the transmitter output power. Core heating of ions and electrons was confirmed in the experiment and density profile peaking was found to be a special feature of IBW heating. Peaking of the density profile was also found when IBWs were injected into neutral beam heated discharges. An analysis, using a transport code with these experimental data, indicates that particle and energy confinement should be improved in the plasma core region upon application of IBW heating. It is also found that the ion energy distribution function observed during IBW heating has a smaller high energy tail than those observed in conventional fast magnetosonic wave ICRF heating regimes. The ion energy distribution function obtained during IBW heating is in reasonable agreement with that calculated using the quasi-linear RF diffusion/Fokker-Planck model

Trends in negative pressure over a few thousand cavitation events have been investigated for a Berthelot system comprising water/stainless steel/tube/sealing Ni plug in order to study the effects and extents of surface pre-treatments for the sealing metal, polishing, etching, and electroplating. The polishing and etching pre-treatments separated two kinds of cavitation nuclei on the metal surface; gases trapped within pores (crevices) (i) in the metal surface, and (ii) extending deep into the metal bulk. The latter, although much tougher, influenced negative pressure for a thousand cavitation events at most. The rise in negative pressure after a total of 1800 temperature cycles levelled off at around -115 bar (-11.5 MPa) irrespective of the surface finishes of the plugs. Electroplated plugs gave poor increasing trends owing to the co-deposited hydrogen. Impurity gases in the metal did limit the rise in negative pressure. These results, together with those reported before, have validated the authors extended gas-trapping crevice model, negative pressure is eventually limited by a supply of gas from sources in the metal bulk to surface crevices.