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Ion cyclotron radio frequency range (ICRF) power plays an important role in heating and current drive in fusion devices. However, experiments show that in the ICRF regime there is a formation of a radio frequency (RF) sheath at the material and antenna boundaries that influences sputtering and power dissipation. Given the size of the sheath relative to the scale of the device, it can be approximated as a boundary condition (BC). Electromagnetic field solvers in the ICRF regime typically treat material boundaries as perfectly conducting, thus ignoring the effect of the RF sheath. Here we describe progress on implementing a model for the RF sheath based on a finite impedance sheath BC formulated by J. Myra and D. A. D'Ippolito, Physics of Plasmas 22 (2015) which provides a representation of the RF rectified sheath including capacitive and resistive effects. This research will discuss the results from the development of a parallelized cold-plasma wave equation solver Stix that implements this non-linear sheath impedance BC through the method of finite elements in pseudo-1D and pseudo-2D using the MFEM library. The verification and comparison of the sheath BC from Stix with results from H. Kohno and J. Myra, Computer Physics Communications 220 129-142 (2017) will also be discussed., The following article has been submitted to/accepted by AIP Conference Proceedings. After it is published, it will be found at https://pubs.aip.org/aip/acp

Based on the community input at the National Academy of Sciences (NAS) Madison and Austin workshops in July and December 2017, respectively, this whitepaper was prepared and submitted to the NAS in the category of US fusion theory and computation. This whitepaper was submitted to NAS as one of five community-approved whitepapers. The revised version was also submitted for the Knoxville American Physical Society Division of Plasma Physics Community Planning Process (APS-DPP-CPP) workshop in September 2019., Fusion theory and computation whitepaper, 10 pages

Proceeding of: 35th International Electric Propulsion Conference (IEPC), October 8-12, 2017, Atlanta, Georgia, USA The physical mechanisms involved in the generation, propagation and absorption of microwaves in electron-cyclotron-resonance plasma thrusters, and their relevance in the operation of these devices, are discussed. The features of the electromagnetic waves and electron motion near the resonance region are analyzed with a one-dimensional model. The characteristics of the two-dimensional wave-plasma problem are examined, and a trade-off of different numerical models is presented as a first step toward the development of an ECR wave-plasma interaction simulation code. The research leading to these results has received funding from the European Union H2020 Program under grant agreement number 730028 (Project MINOTOR). Mario Merino's research visit at the Plasma Science and Fusion Center of the Massachusetts Institute of Technology was funded by the Spanish R&D National Plan under grant number ESP2016-75887-P.

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)

Lower hybrid current drive (LHCD) is an attractive tool for off-axis current profile control in magnetically confined tokamak plasmas and burning plasmas (ITER), because of its high current drive efficiency. The LHCD system on Alcator C-Mod operates at 4.6 GHz, with ~ 1 MW of coupled power, and can produce a wide range of launched parallel refractive index (n[subscript ∣∣]) spectra. A 32 chord, perpendicularly viewing hard x-ray camera has been used to measure the spatial and energy distribution of fast electrons generated by lower hybrid (LH) waves. Square-wave modulation of LH power on a time scale much faster than the current relaxation time does not significantly alter the poloidal magnetic field inside the plasma and thus allows for realistic modeling and consistent plasma conditions for different n[subscript ∣∣] spectra. Inverted hard x-ray profiles show clear changes in LH-driven fast electron location with differing n[subscript ∣∣]. Boxcar binning of hard x-rays during LH power modulation allows for ~ 1 ms time resolution which is sufficient to resolve the build-up, steady-state, and slowing-down phases of fast electrons. Ray-tracing/Fokker-Planck modeling in combination with a synthetic hard x-raydiagnostic shows quantitative agreement with the x-ray data for high n[subscript ∣∣] cases. The time histories of hollow x-ray profiles have been used to measure off-axis fast electron transport in the outer half of the plasma, which is found to be small on a slowing down time scale., United States. Dept. of Energy (Award DE-FC02-99ER54512), United States. Dept. of Energy (Award DE-AC02-76CH03073)

Low-power measurements of the W7-X prototype ICRH antenna and variable capacitor units have been performed at FZJ Julich. The antenna scattering matrices, with and without the carbon limiter tiles, have been de-embedded from network analyzer measurements using custom-built RF adapters. The results have been compared to the matrices calculated with CST Microwave Studio 2017 using a model based on the original CAD files. A very good agreement is seen, since a) the difference in magnitude is less than 0.02 in the whole frequency range, and similar to 0.015 or less in the range of interest (between 20 and 40 MHz) and b) the difference in phase is less than 5 degrees in the whole frequency range, and similar to 2 degrees or less between 20 and 40 MHz. The S-parameters of the tuning capacitors have also been measured, to determine the relationship between the internal DC capacitance C-DC of each unit, and the actual "equivalent" capacitance C-eq seen by the antenna at the ports. It is seen that the latter, 36-475 pF, is substantially different from the nominal DC range, which is 15-200 pF. Nonetheless this is not expected to pose a problem for antenna operation. The S-parameters of other key components of the ICRF system, such as the matching unit and line sections containing directional couplers, were also measured and, where applicable, compared to models or used to extract calibration data.

We present a comprehensive overview of the use of ICRF (ion cyclotron range of frequency) scheme from its beginnings, with specific attention paid to the progress that has been achieved on the understanding of the interaction between the plasma edge and the ICRF heating. The developments and the progress made in overcoming the challenges faced are emphasized. We also describe, based on personal experiences, some of the dead-end roads, the sound deductions that sometimes were prematurely discarded or the intuitive conclusions that were later confirmed by more experimental data. We end with a vision for the future.

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)

Mode conversion of fast waves in the ion cyclotron range of frequencies (ICRF) is known to result in current drive and flow drive under optimised conditions, which may be utilized to control plasma profiles and improve fusion plasma performance. To describe these processes accurately in a realistic toroidal geometry, numerical simulations are essential. Quantitative comparison of these simulations and the actual experimental measurements is important to validate their predictions and to evaluate their limitations. The phase contrast imaging (PCI) diagnostic has been used to directly detect the ICRF waves in the Alcator C-Mod tokamak. The measurements have been compared with full-wave simulations through a synthetic diagnostic technique. Recently, the frequency response of the PCI detector array on Alcator C-Mod was recalibrated, which greatly improved the comparison between the measurements and the simulations. In this study, mode converted waves for D-{superscript 3]He and D-H plasmas with various ion species compositions were re-analyzed with the new calibration. For the minority heating cases, self-consistent electric fields and a minority ion distribution function were simulated by iterating a full-wave code and a Fokker-Planck code. The simulated mode converted wave intensity was in quite reasonable agreement with the measurements close to the antenna, but discrepancies remain for comparison at larger distances., United States. Department of Energy (Grant DE-FG02- 94ER54235)

Parametric decay instabilities (PDI) appear to be an ubiquitous feature of lower hybrid current drive (LHCD) experiments at high density. In density ramp experiments in Alcator C-Mod and other machines the onset of PDI activity has been well correlated with a decrease in current drive efficiency and production of fast electron bremsstrahlung. However whether PDI is the primary cause of the 'density limit', and if so by exactly what mechanism (beyond the obvious one of pump depletion) has not been clearly established. In order to further understand the connection, the frequency spectrum of PDI activity occurring during Alcator C-Mod LHCD experiments has been explored in detail by means of a number of RF probes distributed around the periphery of the C-Mod tokamak including a probe imbedded in the inner wall. The results show that (i) the excited spectra consists mainly of a few discrete ion cyclotron (IC) quasi-modes, which have higher growth than the ion sound branch; (ii) PDI activity can begin either at the inner or outer wall, depending on magnetic configuration; (iii) the frequencies of the IC quasi-modes correspond to the magnetic field strength close to the low-field side (LFS) or high-field side separatrix; and (iv) although PDI activity may initiate near the inner separatrix, the loss in fast electron bremsstrahlung is best correlated with the appearance of IC quasi-modes characteristic of the magnetic field strength near the LFS separatrix. These data, supported by growth rate calculations, point to the importance of the LFS scrape-off layer (SOL) density in determining PDI onset and degradation in current drive efficiency. By minimizing the SOL density it is possible to extend the core density regime over which PDI can be avoided, thus potentially maximizing the effectiveness of LHCD at high density. Increased current drive efficiency at high density has been achieved in FTU and EAST through lithium coating and special fuelling methods, and in recent C-Mod experiments by operating at higher plasma current. Another approach would be to locate the launcher in the inner wall with double null operation. This would reduce the SOL density by an order of magnitude or more and greatly mitigate the effects of PDI as well as other parasitic losses., United States. Department of Energy (DE-FC02-99ER54512)

Launching radio frequency (RF) waves from the high field side (HFS) of a tokamak offers significant advantages over low field side (LFS) launch with respect to both wave physics and plasma material interactions (PMI). For lower hybrid (LH) waves, the higher magnetic field opens the window between wave accessibility (n∥≡ck∥/ω>1−ω2pi/ω2+ω2pe/ω2ce−−−−−−−−−−−−−−−−√+ωpe/∣∣ωce∣∣) and the condition for strong electron Landau damping (n∥∼30/Te−−−−−√ with Te in keV), allowing LH waves from the HFS to penetrate into the core of a burning plasma, while waves launched from the LFS are restricted to the periphery of the plasma. The lower n∥ of waves absorbed at higher Te yields a higher current drive efficiency as well. In the ion cyclotron range of frequencies (ICRF), HFS launch allows for direct access to the mode conversion layer where mode converted waves absorb strongly on thermal electrons and ions, thus avoiding the generation of energetic minority ion tails. The absence of turbulent heat and particle fluxes on the HFS, particularly in double null configuration, makes it the ideal location to minimize PMI damage to the antenna structure. The quiescent SOL also eliminates the need to couple LH waves across a long distance to the separatrix, as the antenna can be located close to plasma without risking damage to the structure. Improved impurity screening on the HFS will help eliminate the long-standing issues of high Z impurity accumulation with ICRF. Looking toward a fusion reactor, the HFS is the only possible location for a plasma-facing RF antenna that will survive long-term. By integrating the antenna into the blanket module it is possible to improve the tritium breeding ratio compared with an antenna occupying an equatorial port plug. Blanket modules will require remote handling of numerous cooling pipes and electrical connections, and the addition of transmission lines will not substantially increase the level of complexity. The obvious engineering challenges associated with locating antenna structures on the HFS can be overcome if HFS antennas are incorporated in the overall experimental design from the start. The Advanced Divertor and radio frequency eXperiment(ADX) will include LH and ICRF antennas located on the HFS. Compact antenna designs based on proven technologies (e.g. multi-junction and “4-way splitter” antennas) fit within the available space on the HFS of ADX. Field aligned ICRF antennas are also located on the HFS. The ADX vacuum vessel design includes dedicated space for transmission lines, pressure windows, and vacuum feedthrus for accessing the HFS wall.

Access to advanced operating regimes in the EAST tokamak will require a combination of electron-cyclotron resonance heating (ECRH), neutral beam injection (NBI) and ion cyclotron range frequency heating (ICRF), with the addition of lower-hybrid current drive (LHCD) for current profile control. Prior experiments at the EAST tokamak facility have shown relatively weak response of the plasma temperature to application of ICRF heating, with typical coupled power about 2 MW out of 12 MW source. The launched spectrum, at n[subscript φ] = 34 for 0-π -0-π phasing and 27 MHz, is largely inaccessible at line-averaged densities of approximately 2 × 10[superscript 19] m[superscript −3]. However, with variable antenna phasing and frequency, this system has considerable latitude to explore different heating schemes. To develop an ICRF actuator control model, we have used the full-wave code TORIC to explore the physics of ICRF wave propagation in EAST. The results presented from this study use a spectrum analysis using a superposition of nφ spanning −50 to +50. The low density regime typical of EAST plasmas results in a perpendicular wavelength comparable to the minor radius which results in global cavity resonance effects and eigenmode formation when the single-pass absorption is low. This behavior indicates that improved performance can be attained by lowering the peak of the k[subscript ||] spectrum by using π/3 phasing of the 4-strap antenna. Based on prior studies conducted at Alcator C-Mod, this phasing is also expected to have the advantage of nearly divergence-free box currents, which should result in reduced levels of impurity production. Significant enhancements of the loading resistance may be achieved by using low k|| phasing and a combination of magnetic field and frequency to vary the location of the resonance and mode conversion regions. TORIC calculations indicate that the significant power may be channeled to the electrons and deuterium majority. We expect that implementation of these recommendations in EAST will yield substantial improvements in the net absorbed power that will greatly assist in the attempt to access advanced tokamak operating regimes., United States. Department of Energy (contract DE-FC02-01ER54648)

The studies of recent JET experiments in H/D approximate to 0.85/0.15 plasma (2.9T/2MA) in which neutron rate was enhanced by applying 2.5MW of ICRH using D-(D-NBI)-H three-ion scheme are reported. An extensive analysis of this novel heating scenario has been carried out by means of integrated TRANSP/TORIC modelling, and a comprehensive validation of the computed Fast Ion Distribution Function (FI DF) with a range of fast ion diagnostics available at JET is presented. The predicted acceleration of D Neutral Beam Injection (NBI) ions beyond their injection energies and the associated changes in FI DF by RF waves are found to be in good agreement with measured neutron yield and TOFOR neutron spectrometer measurements, as well as with multi-channel neutron camera observations and neutral particle analyser diagnostic. An outlook of the possible applications of the developed technique for future DTE2 studies on JET has been highlighted. Controlled acceleration of TNBI ions in D-rich and DNBI ions in T-rich plasmas to optimal energies can be applied to maximise BT fusion rates and contribute to the success of future DT experiments at JET and ITER as illustrated in this study.

A newly developed generic electro-magnetic (EM) simulation tool for modeling RF wave propagation in SOL plasmas is presented. The primary motivation of this development is to extend the domain partitioning approach for incorporating arbitrarily shaped SOL plasmas and antenna to the TORIC core ICRF solver, which was previously demonstrated in the 2D geometry [S. Shiraiwa, et. al., "HISTORIC: extending core ICRF wave simulation to include realistic SOL plasmas", Nucl. Fusion in press], to larger and more complicated simulations by including a 3D realistic antenna and integrating RF rectified sheath potential model. Such an extension requires a scalable high fidelity 3D edge plasma wave simulation. We used the MFEM [http://mfem.org] , open source scalable C++ finite element method library, and developed a Python wrapper for MFEM (PyMFEM), and then a radio frequency (RF) wave physics module in Python. This approach allows for building a physics layer rapidly, while separating the physics implementation being apart from the numerical FEM implementation. An interactive modeling interface was built on pScope [S Shiraiwa, et. al. Fusion Eng. Des. 112, 835] to work with an RF simulation model in a complicated geometry., United States. Department of Energy. Office of Fusion Energy Sciences (Award DE-FC02-99ER54512 ), United States. Department of Energy (Contract DE-FC02-01ER54648)

The goal of the lower hybrid current drive (LHCD) program on Alcator C-Mod is to develop and optimize ITER-relevant steady-state plasmas by controlling the current density profile. Using a 4×16 waveguide array, over 1 MW of LH power at 4.6 GHz has been successfully coupled to the plasmas. However, current drive efficiency precipitously drops as the line averaged density (n̄ e ) increases above 10[superscript 20]m[superscript −3]. Previous numerical work shows that the observed loss of current drive efficiency in high density plasmas stems from the interactions of LH waves with edge/scrape-off layer (SOL) plasmas [Wallace et al., Physics of Plasmas 19, 062505 (2012)]. Recent observations of parametric decay instability (PDI) suggest that non-linear effects should be also taken into account to fully characterize the parasitic loss mechanisms [Baek et al., Plasma Phys. Control Fusion 55, 052001 (2013)]. In particular, magnetic configuration dependent ion cyclotron PDIs are observed using the probes near n̄[subscript e]≈1.2×10[superscript 20]m[superscript −3] . In upper single null plasmas, ion cyclotron PDI is excited near the low field side separatrix with no apparent indications of pump depletion. The observed ion cyclotron PDI becomes weaker in inner wall limited plasmas, which exhibit enhanced current drive effects. In lower single null plasmas, the dominant ion cyclotron PDI is excited near the high field side (HFS) separatrix. In this case, the onset of PDI is correlated with the decrease in pump power, indicating that pump wave power propagates to the HFS and is absorbed locally near the HFS separatrix. Comparing the observed spectra with the homogeneous growth rate calculation indicates that the observed ion cyclotron instability is excited near the plasma periphery. The incident pump power density is high enough to overcome the collisional homogeneous threshold. For C-Mod plasma parameters, the growth rate of ion sound quasi-modes is found to be typically smaller by an order of magnitude than that of ion cyclotron quasi-modes. When considering the convective threshold near the plasma edge, convective growth due to parallel coupling rather than perpendicular coupling is likely to be responsible for the observed strength of the sidebands. To demonstrate the improved LHCD efficiency in high density plasmas, an additional launcher has been designed. In conjunction with the existing launcher, this new launcher will allow access to an ITER-like high single pass absorption regime, replicating the JLH (r) expected in ITER. The predictions from the time domain discharge scenarios, in which the two launchers are used, will be also presented., United States. Dept. of Energy (Award No. DE-FC02-99ER54512), United States. Dept. of Energy (Award No. DE-AC02-76CH03073)

Ion cyclotron resonance heating is the primary auxiliary heating on the Alcator C-Mod tokamak and is commonly used on other devices, and is planned for use on ITER. The RF-power density on C-Mod is above 5 MW m−3 providing for a unique opportunity to study wave–particle effects in the high RF power per particle regime. Minority heating produces a highly energetic tail in the minority distribution function which is measured using a compact neutral particle analyser. In this paper, we present the measurements of the fast-ion spectrum between 200 and 2 MeV, compiled over an entire experimental campaign. We also estimate the effective tail temperatures for the fast-ion distribution. We find that the fast-ion distribution is less energetic and less dense with increasing electron density; is more energetic with increasing plasma current; and is more dense but has no measurable change in energy with increasing RF power. Some possible explanations for these findings are discussed., United States. Dept. of Energy (Award DE-FC02-99ER54512)

Assessing the performance of lower hybrid current drive (LHCD) at high density is critical for developing non-inductive current drive systems on future steady-state experiments. Excellent LHCD efficiency has been observed during fully non-inductive operation (η = 2.0 − 2.5 × 10[superscript 19] AW[superscript –1] m[superscript –2] at [¯ over n][subscript e] = 0.5 × 10[superscript 20] m[superscript –3]) on Alcator C-Mod [I. H. Hutchinson et al., Phys. Plasmas1, 1511 (1994)] under conditions (n[subscript e], magnetic field and topology, and LHCD frequency) relevant to ITER [S. Shiraiwa et al., Nucl. Fusion 51, 103024 (2011)]. To extend these results to advanced tokamak regimes with higher bootstrap current fractions on C-Mod, it is necessary to increase [¯ over n][subscript e] to 1.0 − 1.5 × 10[superscript 20] m[superscript −3]. However, the number of current-carrying, non-thermal electrons generated by LHCD drops sharply in diverted configurations at densities that are well below the density limit previously observed on limited tokamaks. In these cases, changes in scrape off layer (SOL)ionization and density profiles are observed during LHCD, indicating that significant power is transferred from the LH waves to the SOL.Fokker-Planck simulations of these discharges utilizing ray tracing and full wave propagation codes indicate that LH waves in the high density, multi-pass absorption regime linger in the plasma edge, and SOL region, where absorption near or outside the LCFS results in the loss of current drive efficiency. Modeling predicts that non-thermal emission increases with stronger single-pass absorption. Experimental data show that increasing T[subscript e] in high density LH discharges results in higher non-thermal electron emission, as predicted by the models., United States. Dept. of Energy (Award DE-FC02-99ER54512), United States. Dept. of Energy (Award DE-AC02-09CH11466)

The Lower Hybrid Current Drive (LHCD) system on Alcator C-Mod can produce spectra with a wide range of peak parallel refractive index (n||). An experiment in which LH power is square-wave modulated on a time scale much faster than the current relaxation time does not significantly alter the poloidal magnetic field inside the plasma and thus allows for realistic modeling and consistent plasma conditions for different ny spectra. Boxcar binning of hard x-rays during LH power modulation allows for time resolution sufficient to resolve the build-up, steady-state, and slowing-down of fast electrons. A transport model built in Matlab has been used to determine a fast electron pinch velocity for a high-n|| case of 1-2 m/s., United States. Dept. of Energy (US DOE award DE-FC02- 99ER54512), United States. Dept. of Energy (US DOE award DE-AC02-76CH03073)

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)

Radio frequency waves in the ion cyclotron range of frequencies (ICRF) are widely used to heat tokamak plasmas. In ICRF heating schemes involving multiple ion species, the launched fast waves convert to ion cyclotron waves or ion Bernstein waves at the two-ion hybrid resonances. Mode converted waves are of interest as actuators to optimise plasma performance through current drive and flow drive. In order to describe these processes accurately in a realistic tokamak geometry, numerical simulations are essential, and it is important that these codes be validated against experiment. In this study, the mode converted waves were measured using a phase contrast imaging technique in D-H and D-[superscript 3]He plasmas. The measured mode converted wave intensity in the D-[superscript 3]He mode conversion regime was found to be a factor of ∼ 50 weaker than the full-wave predictions. The discrepancy was reduced in the hydrogen minority heating regime, where mode conversion is weaker., United States. Dept. of Energy (DE-FG02-94-ER54235), United States. Dept. of Energy (DE-FC02-99-ER54512), United States. Dept. of Energy (DE-FC02-01ER54648)

This paper reports on new experiments at Alcator C-Mod that were performed in order to investigate the long-standing, unresolved discrepancy between Thomson scattering (TS) and electron cyclotron emission (ECE) measurements of electron temperature in high temperature tokamak plasmas. Ion cyclotron range of frequency (ICRF) heating is used to produce high temperature conditions where the type of TS-ECE discrepancy observed in the past at JET and TFTR should become observable. At Alcator C-Mod, plasmas with Te(0) up to 8 keV are obtained using ion cyclotron resonance heating (ICRH), ICRF mode conversion heating and a combination of the two heating methods in order to explore the hypothesis that the presence of ICRH-generated fast ions may be related to the discrepancy. In all high temperature cases, the TS and ECE measurements of electron temperature agree to within experimental uncertainties. We find no evidence for the type of discrepancy reported at JET and TFTR. These results show that the TS-ECE discrepancy does not depend on high temperatures alone and also that the presence of ICRH-generated fast ions is insufficient to cause the TS-ECE discrepancy., United States. Dept. of Energy (DE-FC02-99ER54512)

Experimental observations of lower hybrid current drive (LHCD) at high density on the Alcator C-Mod tokamak are presented in this paper. Bremsstrahlung emission from relativistic fast electrons in the core plasma drops suddenly above line-averaged densities of 10[superscript 20] m[superscript −3] (ω/ω[subscript LH] ~ 3) in single null discharges with large (≥8 mm) inner gaps, well below the density limit previously observed on limited tokamaks (ω/ω[subscript LH] ~ 2). Modelling and experimental evidence suggest that the absence of LHCD driven fast electrons at high density may be due to parasitic collisional absorption in the scrape-off layer (SOL). Experiments show that the population of fast electrons produced by LHCD at high density ([bar over n][subscript e] > 10[superscript 20] m[superscript -3]) can be increased by operating with an inner gap of less than ~5 mm with the strongest non-thermal emission in inner wall limited plasmas. A change in plasma topology from single to double null produces a modest increase in non-thermal emission at high density. Increasing the electron temperature in the periphery of the plasma (0.8 > r/a > 1.0) also results in a modest increase in non-thermal electron emission above the density limit. Ray tracing/Fokker–Planck simulations of these discharges predict the observed sensitivity to plasma position when the effects of collisional absorption in the SOL are included in the model., United States. Dept. of Energy (Award DE-FC02-99ER54512), United States. Dept. of Energy (Award DE-AC02-76CH03073)

openaire: EC/H2020/633053/EU//EUROfusion The European Integrated Modelling effort (EU-IM) provides the European Transport Simulator (ETS) [1] which was designed to simulate arbitrary tokamak plasma discharges. Two new 1D Fokker-Planck solvers have recently been implemented within ETS: StixRedist [3] and FoPla [4]. To ensure the CPU time remains acceptable, the latter was parallelized with a generic and easy to implement method. In this paper, it will be shown how these modules were integrated in the ETS workflow in particular a first approach adopted to reach a consistency between wave and Fokker-Planck equation resolution. Also, the Verification and Validation efforts will be discussed. JET shots were analyzed and the ETS predictions were cross-checked against earlier validated codes external to the EU-IM effort, TRANSP [5] in particular, as well as against experimental neutron yield data. A good agreement was obtained, both when comparing the predictions with other codes for cases within their reach (minority or beam populations) and with experimental neutron yield data. Simulations illustrating the exploitation of the nonlinear collision operator when solving a set of coupled Fokker-Planck equations for cases when majority species play a key role will be also shown.

This manuscript overviews the latest developments and results from the ASDEX Upgrade ICRF team. The ICRF control system has been upgraded to allow the operation of active antennas with a controlled arbitrary phase difference between them. This upgrade, in combination with an extensive RF probe coverage inside the ASDEX Upgrade torus, makes it possible to study a global wave phenomenon that may result from the two antenna pairs' fields. The first results show that multiple active antennas form a superposition of the launched waves, with limited plasma shielding. A combination of high- and low-field side RF probes makes it possible to study the efficiency of various heating schemes such as the hydrogen minority or the 3-ion species scheme with a He-3 minority. The effect of the hydrogen minority concentration on the ICRF wave absorption and transmission through the core is best shown during conditioning discharges following a torus opening: as the hydrogen fraction is reduced from 80% to below 20%, the amplitude of the RF waves that reaches the high-field side probes is reduced by nearly two orders of magnitude indicating a strong absorption increase. The experimental results are supported by full wave FELICE code showing a similar absorption improvement as a function of the hydrogen fraction. The low-field side probes, on the other hand, do not show a sensitivity to the hydrogen fraction, which is expected as the low-field side wave amplitude is dependent on antenna coupling, not wave absorption. For the case of the 3-ion heating scheme, the high-field side probes detect a minimum in the wave amplitude as the wave absorption layer is radially scanned past the optimal position in the core. The minimum is interpreted as the optimal core absorption condition and matches the observed maximum in the core soft x-ray data. The optimal radial position of the wave absorption layer is also indirectly indicated via the appearance of strong ion cyclotron emission (ICE). The frequency of ICE is matched with the cyclotron frequency of RF-accelerated He-3 ions and the emission origin is placed similar to midway between the magnetic axis and the separatrix. In order to self-consistently assess the global ICRF wave behavior, it is necessary to account for the three dimensional (3D) nature of the toroidal tokamak plasma and the launching antenna structures. The ASDEX Upgrade ICRF team studies multiple aspects of how 3D effects influence launched ICRF wave fields and antenna performance. These include 3D plasma density perturbations in the plasma edge and the scrape off layer, either intrinsically generated by turbulence or externally imposed by resonant magnetic perturbation coils. The experimental study is supported with RAPLICASOL: a computational tool that solves for launched ICRF wave fields in the presence of realistic 3D plasma density profiles and antenna geometry. Additionally, the work of the ASDEX Upgrade ICRF team includes fundamental plasma physics and RF wave studies on the linear test device IShTAR: the areas of interest are RF sheath rectification and arc formation.