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Self-consistent modeling using the stability, transport, equilibrium, and pedestal (STEP) workflow in the OMFIT integrated modeling framework (predicting pedestal with EPED, core profiles with TGYRO, current profile with ONETWO, and EFIT for equilibrium) suggests ITER and future devices such as China Fusion Engineering Test Reactor (CFETR) Zhuang et al (2019 Nucl. Fusion 59 112010) will benefit from high-density operation (Greenwald limit fraction $f_\mathrm{gw} \approx$ 0.7−1.3). Regimes with an operational density near the Greenwald limit will likely need peaked density profiles so that the pedestal density remains below the Greenwald limit. Peaked density profiles can be achieved with the help of pellet injection. A flexible Pellet Ablation Module (PAM), which predicts the density source based on a comprehensive analytical pellet ablation model, has been developed for predicting pellet fueling for transport studies, and has been incorporated into the STEP workflow for predictive modeling. This workflow is applied to DIII-D and finds good agreement with experiments. On ITER the effect of pellet fueling is examined in an advanced inductive scenario, where a fusion gain of up to Q = 9 is predicted with strong central pellet fueling. On CFETR, with a mid-radius density source, an average of $1.5 \times 10^{22}$ electrons s ^−1 are required to achieve the density and temperature profiles necessary for the 1000 MW advanced scenario with a tritium burn-up fraction of ${\sim}3\%$ .

Nonlinear 3D MHD simulations of shattered-pellet injection (SPI) in JET show prototypical SPI-driven disruptions using the M3D-C1 and NIMROD extended-MHD codes. Initially, radiation-driven thermal quenches are accelerated by MHD activity as the pellet crosses rational surfaces, leading to a radiation spike, global stochasticization of the magnetic field, and a complete thermal quench. Eventually, current quenches, preceded by a current spike are seen as the Ohmic heating becomes equal to the radiative cooling. The results are qualitatively similar for both a single monolithic pellet, pencil-beam model, and a realistic shatter to represent the SPI plume. A scan in viscosity from 500 to 2000 m ^2 s ^−1 for MHD simulations finds that reducing viscosity increases MHD activity and decreases thermal quench time slightly. A realistic cloud of fragments modeling shows that mixed-D–Ne pellet travels deeper into the plasma core before the thermal quench. At the slow pellet speeds, the pellet is found to be moving slowly enough inward that even the 5% neon in the mixed pellet is enough to effectively radiate the thermal energy available. Radiation toroidal peaking is predicted to be at levels consistent with experimental observations and reduced as the pellet travels deeper into the plasma. These simulations lay the ground work for more-sophisticated validative and predictive modeling of SPI in JET using both M3D-C1 and NIMROD.

On DIII-D, the high scenario has an internal transport barrier (ITB), , , and very high normalized confinement . Recently, plasmas starting with these conditions have been dynamically driven to and , where we find the ITB and high performance persist for five energy confinement times. These conditions are projected to meet the ITER steady-state goal of Q  =  5. The ITB is maintained at lower with a strong reverse shear, consistent with predictions that negative central shear can lower the threshold for the ITB. There are two observed confinement states in the high scenario: H-mode confinement state with a high edge pedestal, and an enhanced confinement state with a low pedestal and an ITB. It has been observed in a scan of external resonant magnetic perturbation amplitude that when there are no large type-I ELMs, there is no transition to enhanced confinement. This is consistent with the proposed mechanism for ITB formation being a type-I ELM. Quasilinear gyro-Landau fluid predictive modeling of ITER suggests that only a modest reverse shear is required to achieve the ITB formation necessary for Q  =  5 when electromagnetic physics including the kinetic ballooning mode (KBM) is incorporated. [ABSTRACT FROM AUTHOR]

A drift orbit model for relativistic test electrons has been incorporated into the MARS-F code (Liu et al 2000 Phys. Plasmas7 3681), in order to study the runaway electron (RE) behavior in the presence of magneto-hydrodynamic perturbations computed by MARS-F. By implementing the model directly into the MARS-F curve-linear magnetic coordinates, maximal accuracy in representing the full field perturbation is preserved. The updated code is utilized to study the high current RE beam loss in a post-disruption DIII-D plasma, revealing that a fast growing, n  =  1 (n is the toroidal mode number) resistive kink instability, at  ∼100 Gauss level, can induce significant fraction of RE loss, largely by perturbing drift orbits of REs. A  ∼1000 Gauss perturbation fully terminates the RE beam, as found in both experiment and modeling. The 3D field induced loss increases with the perturbation amplitude but decreases with the particle energy. The loss fraction is generally not sensitive to the initial particle pitch angle. The particle velocity change, due to electric field acceleration/deceleration, small pitch angle scattering, synchrotron radiation and Bremsstrahlung, further perturbs the RE trajectory but plays a minor role in prompt RE loss within microseconds time scale. Therefore, the dominant dependencies are simply the RE energy and instability strength. For comparison, a resonant magnetic perturbation field, generated by 4 kAt n  =  3 even parity I-coil currents in DIII-D and with the plasma response field included, is found to induce almost no loss for the same RE beam. [ABSTRACT FROM AUTHOR]

The runaway electron (RE) distributions driven by a large toroidal electric field induced by the drop in the temperature profile due to disruption or pellets are comprehensively simulated by the 3D Fokker–Planck (FP) solver CQL3D (Harvey and McCoy 1992 Proc. of IAEA TCM), recently coupled to the Ampere–Faraday (AF) equations. The evolution of the toroidal current in a plasma occurs on a resistive time scale, τres  =  4πa2/(c2η), which is typically of the order of seconds in present tokamaks. Here, a and η are respectively plasma radius or radial extent of a current density perturbation, and Ohmic resistivity. From the Faraday EM equation, the toroidal electric field is proportional to the time derivative of the poloidal magnetic field, which, from the Ampere equation, is proportional to the toroidal current. Thus, the toroidal electric field rapidly increases due to an abrupt temperature drop decrease in conductivity, to prevent change in the toroidal current faster than τres. This is a example of Lenz’s law. For example, in simulations with KPRAD (Whyte et al 2003 J. Nucl. Mater. 313–6 1239) of neon pellet injection into a DIII-D shot, Te drops from 2 keV to 10 eV in 0.1 ms and Zeff increases 1–4, giving that the electric field increases 3500×  to 0.8 V cm−1. As described in Harvey et al (2000 PoP 7 4590), this places much of the tail electron distribution beyond the Dreicer runaway velocity, giving so-called ‘hot-tail runaways’ which for a time are the dominant source of runaways, more so than the knockon source. In this prior calculation, performed for a single flux surface, the toroidal current density is held constant, on the basis that τres is large. Most of the initial current can be converted to runaway current, which is then dangerous, particularly for ITER. A more comprehensive A–F model recently implemented in CQL3D, taking into account the time-development of the full-plasma-width toroidal electric field on time-scales of order τres applies an iterative technique for the toroidal field previously developed for a different application (Kupfer et al 1996 PoP 3 3644), maintaining the implicit-in-time evolution of CQL3D. The degree of runaway current formation is reduced in AF augmented CQL3D, but the basic mechanism of ‘hot-tail runaways’ remains a dominant contribution to the REs at early times after the Te drop in these simulations. On the other hand, a NIMROD (Sovinec et al 2004 J. Comput. Phys. 195 355) simulation of shattered-pellet shutdown of DIII-D plasma (Kim 2018 APS/DPP Meeting), gives a slower thermal quench; when the plasma profiles and electric field are coupled one-way to CQL3D, the ‘hot-tail’ REs are much less, and growth of RE is dominated by the knockon process. [ABSTRACT FROM AUTHOR]

A new small angle slot (SAS) divertor concept has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Initial tests on DIII-D find a strong interplay between such anticipated ‘SAS’ effects and cross-field drifts, favouring operation with the ion B  ×  ∇B drift away from the X-point, as currently employed for advanced tokamaks. This offers the following key improvements relative to DIII-D’s open lower divertor or partially-closed upper divertor: (i) SAS allows for transition to low temperature moderately detached divertor conditions with Te  ≲  10 eV at very low main plasma densities, lower than are usually attainable at all in DIII-D high confinement (H-mode) plasmas as used in these tests; (ii) Pedestal performance and core confinement are significantly improved with SAS. The final confinement collapse associated with the onset of X-point MARFE (multifaceted asymmetric radiation from the edge) following deep detachment occurs at significantly higher pedestal densities, thus widening the window of H-mode operation compatible with a dissipative divertor. For operation with the ion B  ×  ∇B drift toward the X-point, the divertor plasma transitions to a bifurcative detached state at much higher densities, similar to other divertor configurations in DIII-D. These results highlight the strong interplay between divertor closure and drifts, and point to an interesting divertor optimization path to explore that offers potential for future fusion reactors. [ABSTRACT FROM AUTHOR]

Large, spontaneous m/n  =  1/1 helical cores are shown to be expected in tokamaks such as ITER with extended regions of low- or reversed- magnetic shear profiles and q near 1 in the core. The threshold for this spontaneous symmetry breaking is determined using VMEC scans, beginning with reconstructed 3D equilibria from DIII-D and Alcator C-Mod based on observed internal 3D deformations. The helical core is a saturated internal kink mode (Wesson 1986 Plasma Phys. Control. Fusion28 243); its onset threshold is shown to be proportional to around q  =  1. Below the threshold, applied 3D fields can drive a helical core to finite size, as in DIII-D. The helical core size thereby depends on the magnitude of the applied perturbation. Above it, a small, random 3D kick causes a bifurcation from axisymmetry and excites a spontaneous helical core, which is independent of the kick size. Systematic scans of the q-profile show that the onset threshold is very sensitive to the q-shear in the core. Helical cores occur frequently in Alcator C-Mod during ramp-up when slow current penetration results in a reversed shear q-profile, which is favorable for helical core formation. Finally, a comparison of the helical core onset threshold for discharges from DIII-D, Alcator C-Mod and ITER confirms that while DIII-D is marginally stable, Alcator C-Mod and especially ITER are highly susceptible to helical core formation without being driven by an externally applied 3D magnetic field. [ABSTRACT FROM AUTHOR]

The Small Angle Slot (SAS) was recently installed on DIII-D as an advanced divertor, promising easier plasma detachment and lower temperatures across the whole target. A twofold study of the SAS magnetic geometry and topology is presented in this paper. On one hand, a two-dimensional uncertainty quantification analysis is carried out through a Monte Carlo approach in order to understand the level of accuracy of two-dimensional equilibrium computations in reconstructing the strike point and angle onto the divertor. Under typical experimental conditions, the uncertainties are found to be roughly 6.8 mm and 0.56 deg, respectively. On the other hand, a three-dimensional ‘vacuum’ analysis is carried out to understand the effects of typical external perturbation fields on the scrape-off layer topology. When the non-axisymmetric I-coils are switched on, poloidally-localized lobes are found to appear, grow, and hit the SAS target, although barely, even for 5 kA; at the same time, the strike point modulation is found to be roughly 1.8 mm and thus negligible for most purposes. Such results complement previous two-dimensional analyses in characterizing typical SAS equilibria and provide useful background information for planning and interpreting SAS experiments. [ABSTRACT FROM AUTHOR]

Investigation on the equilibrium operation regime, its ideal magnetohydrodynamics (MHD) stability and edge localized modes (ELM) characteristics is performed for the China Fusion Engineering Test Reactor (CFETR). The CFETR operation regime study starts with a baseline scenario (R  =  5.7 m, BT  =  5 T) derived from multi-code integrated modeling, with key parameters varied to build a systematic database. These parameters, under profile and pedestal constraints, provide the foundation for the engineering design. The long wavelength low-n global ideal MHD stability of the CFETR baseline scenario, including the wall stabilization effect, is evaluated by GATO. It is found that the low-n core modes are stable with a wall at r/a  =  1.2. An investigation of intermediate wavelength ideal MHD modes (peeling ballooning modes) is also carried out by multi-code benchmarking, including GATO, ELITE, BOUT++ and NIMROD. A good agreement is achieved in predicting edge-localized instabilities. Nonlinear behavior of ELMs for the baseline scenario is simulated using BOUT++. A mix of grassy and type I ELMs is identified. When the size and magnetic field of CFETR are increased (R  =  6.6 m, BT  =  6 T), collisionality correspondingly increases and the instability is expected to shift to grassy ELMs. [ABSTRACT FROM AUTHOR]

Transport modeling of a proposed ITER steady-state scenario based on DIII-D high poloidal-beta () discharges finds that ITB formation can occur with either sufficient rotation or a negative central shear q-profile. The high scenario is characterized by a large bootstrap current fraction (80%) which reduces the demands on the external current drive, and a large radius internal transport barrier which is associated with excellent normalized confinement. Modeling predictions of the electron transport in the high scenario improve as approaches levels similar to typical existing models of ITER steady-state and the ion transport is turbulence dominated. Typical temperature and density profiles from the non-inductive high scenario on DIII-D are scaled according to 0D modeling predictions of the requirements for achieving a steady-state fusion gain in ITER with ‘day one’ heating and current drive capabilities. Then, TGLF turbulence modeling is carried out under systematic variations of the toroidal rotation and the core q-profile. A high bootstrap fraction, high scenario is found to be near an ITB formation threshold, and either strong negative central magnetic shear or rotation in a high bootstrap fraction are found to successfully provide the turbulence suppression required to achieve . [ABSTRACT FROM AUTHOR]

An efficient technique for building plasma equilibria with preset or constrained q profiles by solving the nonlinear Grad–Shafranov equation iteratively is proposed. The technique uses a set of dynamic virtual poloidal field probes in the plasma core region. Numerical examples are shown for monotonic q profiles and negative central magnetic shear equilibria which are obtained with this technique. As a proof of principle, the technique is applied on the EAST tokamak, to constrain the plasma equilibrium reconstruction with the experimental q  =  1 surface information from the ECE diagnostic in addition to the external magnetic measurements. [ABSTRACT FROM AUTHOR]

Recent experiments on EAST have achieved the first long pulse H-mode (61 s) with zero loop voltage and an ITER-like tungsten divertor, and have demonstrated access to broad plasma current profiles by increasing the density in fully-noninductive lower hybrid current-driven discharges. These long pulse discharges reach wall thermal and particle balance, exhibit stationary good confinement (H98y2 ~ 1.1) with low core electron transport, and are only possible with optimal active cooling of the tungsten armors. In separate experiments, the electron density was systematically varied in order to study its effect on the deposition profile of the external lower hybrid current drive (LHCD), while keeping the plasma in fully-noninductive conditions and with divertor strike points on the tungsten divertor. A broadening of the current profile is found, as indicated by lower values of the internal inductance at higher density. A broad current profile is attractive because, among other reasons, it enables internal transport barriers at large minor radius, leading to improved confinement as shown in companion DIII-D experiments. These experiments strengthen the physics basis for achieving high performance, steady state discharges in future burning plasmas. [ABSTRACT FROM AUTHOR]

Using the SOLPS5.0 code, an assessment has been undertaken of the relative importance of two key aspects of divertor baffle geometry: (i) divertor closure, and (ii) field-target angle. The figures of merit (FOMs) have been taken to be achieving specified (low) values of electron temperature at the target, Tet, and parallel power flux density at the target, q‖t, at the lowest possible ‘upstream’ density, ne,sep,OMP, for a given power input. As expected, it was found that increased closure for both orthogonal targets and small glancing-angle targets make for a better divertor, based on these FOMs. It was also found that the combination of these aspects is significantly more effective than either aspect separately. The latter appears to be related to a strong correlation between Tet and deuterium gas density at the target, which is the subject of a companion paper (Stangeby P.C. and Sang C.F. 2017 Nucl. Fusion57 056007). [ABSTRACT FROM AUTHOR]

SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks. [ABSTRACT FROM AUTHOR]

The first equilibrium reconstruction of EAST current-density profile based on internal Faraday rotation measurements provided by the POlarimeter-INTerferometer (POINT) diagnostic is demonstrated using the EFIT equilibrium reconstruction code. EFIT incorporates 11 simultaneous line-integrated density and Faraday effect measurements from POINT to self-consistently reconstruct the equilibrium toroidal current density profile using a Faraday rotation reconstruction algorithm. It is shown that the POINT measurements can be applied to improve the accuracy of core plasma current density and q profile on EAST. Comparisons of magnetic surfaces and the q profile reconstructed using external magnetic data against those using magnetic and POINT data are presented. Equilibrium reconstructions using POINT data are found to be consistent with sawtooth phenomena. The sensitivity of equilibrium reconstruction to POINT measurements indicates Faraday rotation provides important constraints for determining the current profile. [ABSTRACT FROM AUTHOR]

Energy transport analyses of the DIII-D high- EAST-demonstration discharges have been performed using the TGYRO transport package with the TGLF turbulent and NEO neoclassical transport models under the OMFIT integrated modeling framework. Ion energy transport is shown to be dominated by neoclassical transport and ion temperature profiles predicted by TGYRO agree closely with the experimental measured profiles for these high- discharges. Ion energy transport is largely insensitive to reductions in the flow shear stabilization. The Shafranov shift is shown to play a role in the suppression of the ion turbulent energy transport below the neoclassical level. Electron turbulent energy transport is under-predicted by TGLF and a significant shortfall in the electron energy transport over the whole core plasma is found with TGLF predictions for these high- discharges. TGYRO can successfully predict the experimental ion and electron temperature profiles by artificially increasing the saturated turbulence level for ETG driven modes used in TGLF. [ABSTRACT FROM AUTHOR]

A technique for tokamak equilibrium reconstructions is used for multiple DIII-D discharges, including L-mode and H-mode cases when weakly 3D fields are applied. The technique couples diagnostics to the non-linear, ideal MHD equilibrium solver VMEC, using the V3FIT code, to find the most likely 3D equilibrium based on a suite of measurements. It is demonstrated that V3FIT can be used to find non-linear 3D equilibria that are consistent with experimental measurements of the plasma response to very weak 3D perturbations, as well as with 2D profile measurements. Observations at DIII-D show that plasma rotation larger than 20 krad s−1 changes the relative phase between the applied 3D fields and the measured plasma response. Discharges with low averaged rotation (10 krad s−1) and peaked rotation profiles (40 krad s−1) are reconstructed. Similarities and differences to forward modeled VMEC equilibria, which do not include rotational effects, are shown. Toroidal phase shifts of up to are found between the measured and forward modeled plasma responses at the highest values of rotation. The plasma response phases of reconstructed equilibra on the other hand match the measured ones. This is the first time V3FIT has been used to reconstruct weakly 3D tokamak equilibria. [ABSTRACT FROM AUTHOR]

In a published Physical Review Letter (Reiman 2007 Phys. Rev. Lett. 99 135007), it was shown that axisymmetric (or vertical) stability can be improved by placing a set of parallelogram coils above and below the plasma oriented at an angle to the constant toroidal planes. The physics of this stabilization can be understood as providing an effective additional positive stability index. The original work was based on a simplified model of a straight tokamak and is not straightforwardly applicable to a finite aspect ratio, strongly shaped plasma such as in DIII-D. Numerical calculations were performed in a real DIII-D -like configuration to provide a proof of principal that 3-D fields can, in fact raise the elongation limits as predicted. A four field period trapezioid-shaped coil set was developed in toroidal geometry and 3D equilibria were computed using trapezium coil currents of , , and . The ideal magnetohydrodynamics growth rates were computed as a function of the conformal wall position for the symmetry-preserving family. The results show an insignificant relative improvement in the stabilizing wall location for the two lower coil current cases, of the order of 10−3 and less. In contrast, the marginal wall position is increased by 7% as the coil current is increased to , confirming the main prediction from the original study in a real geometry case. In DIII-D the shift in marginal wall position of 7% would correspond to being able to move the existing wall outward by 5 to 10 cm. While the predicted effect on the axisymmetric stability is real, it appears to require higher coil currents than could be provided in an upgrade to existing facilities. Additional optimization over the pitch of the coils, the number of field periods and the coil positions, as well as plasma parameters, such as the internal inductivity , , and would mitigate this but seem unlikely to change the conclusion. [ABSTRACT FROM AUTHOR]

Recent EAST/DIII-D joint experiments on the high poloidal beta tokamak regime in DIII-D have demonstrated fully noninductive operation with an internal transport barrier (ITB) at large minor radius, at normalized fusion performance increased by  ⩾30% relative to earlier work (Politzer et al 2005 Nucl. Fusion45 417). The advancement was enabled by improved understanding of the ‘relaxation oscillations’, previously attributed to repetitive ITB collapses, and of the fast ion behavior in this regime. It was found that the ‘relaxation oscillations’ are coupled core-edge modes amenable to wall-stabilization, and that fast ion losses which previously dictated a large plasma-wall separation to avoid wall over-heating, can be reduced to classical levels with sufficient plasma density. By using optimized waveforms of the plasma-wall separation and plasma density, fully noninductive plasmas have been sustained for long durations with excellent energy confinement quality, bootstrap fraction  ⩾80%, , , and . These results bolster the applicability of the high poloidal beta tokamak regime toward the realization of a steady-state fusion reactor. [ABSTRACT FROM AUTHOR]

One modeling framework for integrated tasks (OMFIT) is a comprehensive integrated modeling framework which has been developed to enable physics codes to interact in complicated workflows, and support scientists at all stages of the modeling cycle. The OMFIT development follows a unique bottom-up approach, where the framework design and capabilities organically evolve to support progressive integration of the components that are required to accomplish physics goals of increasing complexity. OMFIT provides a workflow for easily generating full kinetic equilibrium reconstructions that are constrained by magnetic and motional Stark effect measurements, and kinetic profile information that includes fast-ion pressure modeled by a transport code. It was found that magnetic measurements can be used to quantify the amount of anomalous fast-ion diffusion that is present in DIII-D discharges, and provide an estimate that is consistent with what would be needed for transport simulations to match the measured neutron rates. OMFIT was used to streamline edge-stability analyses, and evaluate the effect of resonant magnetic perturbation (RMP) on the pedestal stability, which have been found to be consistent with the experimental observations. The development of a five-dimensional numerical fluid model for estimating the effects of the interaction between magnetohydrodynamic (MHD) and microturbulence, and its systematic verification against analytic models was also supported by the framework. OMFIT was used for optimizing an innovative high-harmonic fast wave system proposed for DIII-D. For a parallel refractive index , the conditions for strong electron-Landau damping were found to be independent of launched and poloidal angle. OMFIT has been the platform of choice for developing a neural-network based approach to efficiently perform a non-linear multivariate regression of local transport fluxes as a function of local dimensionless parameters. Transport predictions for thousands of DIII-D discharges showed excellent agreement with the power balance calculations across the whole plasma radius and over a broad range of operating regimes. Concerning predictive transport simulations, the framework made possible the design and automation of a workflow that enables self-consistent predictions of kinetic profiles and the plasma equilibrium. It is found that the feedback between the transport fluxes and plasma equilibrium can significantly affect the kinetic profiles predictions. Such a rich set of results provide tangible evidence of how bottom-up approaches can potentially provide a fast track to integrated modeling solutions that are functional, cost-effective, and in sync with the research effort of the community. [ABSTRACT FROM AUTHOR]

Measurements show fast ion losses correlated with applied three-dimensional (3D) fields in a variety of plasmas ranging from L-mode to resonant magnetic perturbation (RMP) edge localized mode (ELM) suppressed H-mode discharges. In DIII-D L-mode discharges with a slowly rotating magnetic perturbation, scintillator detector loss signals synchronized with the applied fields are observed to decay within one poloidal transit time after beam turn-off indicating they arise predominantly from prompt loss orbits. Full orbit following using M3D-C1 calculations of the perturbed fields and kinetic profiles reproduce many features of the measured losses and points to the importance of the applied 3D field phase with respect to the beam injection location in determining the overall impact on prompt beam ion loss. Modeling of these results includes a self-consistent calculation of the 3D perturbed beam ion birth profiles and scrape-off-layer ionization, a factor found to be essential to reproducing the experimental measurements. Extension of the simulations to full slowing down timescales, including fueling and the effects of drag and pitch angle scattering, show the applied RMPs in ELM suppressed H-mode plasmas can induce a significant loss of energetic particles from the core. With the applied fields, up to 8.4% of the injected beam power is predicted to be lost, compared to 2.7% with axisymmetric fields only. These fast ions, originating from minor radii , are predicted to be primarily passing particles lost to the divertor region, consistent with wide field-of-view infrared periscope measurements of wall heating in RMP ELM suppressed plasmas. Edge fast ion (FIDA) measurements also confirm a large change in edge fast ion profile due to the fields, where the effect was isolated by using short 50 ms RMP-off periods during which ELM suppression was maintained yet the fast ion profile was allowed to recover. The role of resonances between fast ion drift motion and the applied 3D fields in the context of selectively targeting regions of fast ion phase space is also discussed. [ABSTRACT FROM AUTHOR]