The use of spin-polarized fusion fuels would provide a significant boost towards the ignition of a burning plasma. The cross section for D + T → α + n, would be increased by 1.5 if the fuels were injected with parallel polarization. Furthermore, our simulations demonstrate additional non-linear power gains in large-scale machines such as ITER, due to increased alpha heating. Such benefits require the survival of spin polarizations for periods comparable to the particle confinement time. During the 1980s, calculations predicted that polarizations could survive a plasma environment, although concerns persisted regarding the cumulative impacts of wall recycling. In that era, technical challenges prevented direct tests and left the large scale fueling of a power reactor beyond reach. Over the last decades, this situation has changed dramatically. Detailed simulations of ITER have predicted negligible wall recycling in a high-power reactor, and recent advances in laser-driven sources project the capability of producing large quantities of ∼100% polarized D and T. The remaining crucial step is an in-situ demonstration of polarization survival in a plasma. For this, we outline a measurement strategy using the isospin-mirror reaction, D + 3 He → α + p. Polarized 3 He avoids the complexities of handling tritium, while encompassing the same spin-physics. We evaluate two methods of delivering deuterium, using dynamically polarized Lithium-Deuteride (with vector polarization P V D of 70%) or frozen-spin Hydrogen-Deuteride (with P V D of 40%), together with a method of injecting optically-pumped 3 He (with 65% polarization). Pellets of these materials all have long polarization decay times (∼6 min for LiD at 2 K, ∼2 months for HD at 2 K, and ∼3 d for 3 He at 77 K), all far greater than a plasma shot in a research tokamak such as DIII-D (∼20 s). Both species can be propelled from a single cryogenic injection gun. We review plasma requirements and strategies for detecting polarization survival. Polarization alters both fusion yields and the angular distribution of fusion products, and each of these provides a potential signal. In this paper we simulate a selection of shots with similar characteristics in a future high-T ion H plasma, and find ratios of yields from shots with fuel spins parallel and antiparallel reaching 1.3 (HD + 3 He) to 1.6 (LiD + 3 He) over a wide range of poloidal angles. (A companion paper finds sensitivity to fusion product angular distributions as reflected in the pitch angles of protons and alphas reaching the plasma facing wall.)

Injection of low-Z granules into high performance discharges on DIII-D has been shown to promptly trigger Edge Localized Modes (ELMs) providing high-Z impurity control without significant plasma degradation. The ability to provide ELM triggering over a range of injection and discharge parameters suggests that the mechanical introduction of granules can be considered as an additional method of impurity control in ITER. Utilizing a spherically symmetric vapor shielding model for granule ablation, benchmarked with impurity granule injections on DIII-D, we simulate the injection of beryllium granules into ITER baseline discharges. By comparing the granule induced ELM triggering size required for deuterium and non-fuel pellets on DIII-D and cross-correlating with a previously simulated JOREK calcuation of D pellet size required for ELM triggering in ITER, we estimate that a beryllium pellet of 1.5 mm diameter should provide reliable ELM triggering on ITER. This size pellet, delivered at 200 m/s should penetrate 3.5 cm past the separatrix, solidly within the H-mode steep gradient region, a location found to be advantageous for ELM triggering with minimal pellet size.

Analysis of the evolving current density (J), pedestal and rotation profiles in a database of 200 ITER Baseline Scenario (IBS) demonstration discharges in the DIII-D tokamak identifies the current profile as a key determinant of the disruptive instability limiting both high and low torque operation of these q95 = 3 plasmas. The m = 2/n = 1 tearing modes, occurring after several pressure-relaxation times, are linked to the shape of the current profile in the outer region of the plasma. The q = 2 surface is located at ρ ~ 0.78, near a minimum in J. A steeper ‘well’ around the q = 2 surface is correlated with the instabilities, and this finding is consistent for modes that occur both early (<1 τR) and late (>2 τR) on the βN flattop, at high and low input torque, for discharges with and without direct electron heating. Low rotation or differential rotation in the core plasmas is shown to not be the direct cause of the onset of these instabilities. The current profile trends identified in this work indicate that the modes limiting the Q = 10 operation in the ITER demonstration plasmas are likely due to the slow evolution of the current profile, and reveal which shape of the current profile is more prone to the 2/1 tearing instability.

Experiments have been conducted on DIII-D investigating high repetition rate injection of non-fuel pellets as a tool for pacing Edge Localized Modes (ELMs) and mitigating their transient divertor heat loads. Effective ELM pacing was obtained with injection of Li granules in different H-mode scenarios, at frequencies 3–5 times larger than the natural ELM frequency, with subsequent reduction of strike-point heat flux (Bortolon et al., Nucl. Fus., 56, 056008, 2016). However, in scenarios with high pedestal density (∼6 ×1019m−3), the magnitude of granule triggered ELMs shows a broad distribution, in terms of stored energy loss and peak heat flux, challenging the effectiveness of ELM mitigation. Furthermore, transient heat-flux deposition correlated with granule injections was observed far from the strike-points. Field line tracing suggest this phenomenon to be consistent with particle loss into the mid-plane far scrape-off layer, at toroidal location of the granule injection.

Joint DIII-D/EAST experiments in the radio-frequency (RF) heated H-mode scheme with comparison to that of neutral beam (NB) heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-offlayer (SOL) width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broader SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. The joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH) H-mode plasmas. (C) 2017 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND license.

An overview of the preliminary design of the ITER plasma control system (PCS) is described here, which focusses on the needs for 1st plasma and early plasma operation in hydrogen/helium (H/He) up to a plasma current of 15 MA with moderate auxiliary heating power in low confinement mode (L-mode). Candidate control schemes for basic magnetic control, including divertor operation and kinetic control of the electron density with gas puffing and pellet injection, were developed. Commissioning of the auxiliary heating systems is included as well as support functions for stray field topology and real-time plasma boundary reconstruction. Initial exception handling schemes for faults of essential plant systems and for disruption protection were developed. The PCS architecture was also developed to be capable of handling basic control for early commissioning and the advanced control functions that will be needed for future high performance operation. A plasma control simulator is also being developed to test and validate control schemes. To handle the complexity of the ITER PCS, a systems engineering approach has been adopted with the development of a plasma control database to keep track of all control requirements.

Low-wavenumber density fluctuations exhibit unique characteristics during the current ramp-up phase of ITER-like discharges that can partially explain the challenges of correctly modeling transport behavior and predicting global plasma parameters during this period. A strong interaction takes place between the evolving transport, safety factor (q) and kinetic profiles as well as the appearance and evolution of low-order rational surfaces. Density fluctuations from 0.75 < rho < 0.9 are transiently reduced to exceptionally low levels during early times and from 0.8 < rho < 0.9 at late times in the ramp-up in a manner that is different from behavior observed during steady-state plasma conditions with similar values of q(95). Turbulence is suppressed as low-order-rational q-surfaces enter the plasma; the local electron temperature likewise exhibits transient increases during these periods of reduced fluctuations indicating changes in transport that impact temperature and consequently the evolution of current density and plasma inductance. These observations can explain discrepancies between CORSICA modelling and the higher electron temperature found previously over the outer half radius. Comparison of turbulence properties with time-varying linear growth rates with GYRO and GENE demonstrate qualitative consistency with measured fluctuation levels, but calculations don't exhibit reduced growth rates near low-order rational surfaces, which is inconsistent with experimental observations. This indicates a mechanism that can contribute to reconciling observed turbulence behavior with transport models, allowing for the development of more accurate predictive tools.

At DIII-D, lithium granules were radially injected into the plasma at the outer midplane to trigger and pace edge localized modes (ELMs). Granules ranging in size from 300 to 1000 microns were horizontally launched into H-mode discharges with velocities near 100m/s, and granule to granule injection frequencies less than 500Hz. While the smaller granules were only successful in triggering ELMs approximately 20% of the time, the larger granules regularly demonstrated ELM triggering efficiencies of greater than 80%. A fast visible camera looking along the axis of injection observed the ablation of the lithium granules. The duration of ablation was used as a benchmark for a neutral gas shielding calculation, and approximated the ablation rate and mass deposition location for the various size granules, using measured edge plasma profiles as inputs. This calculation suggests that the low triggering efficiency of the smaller granules is due to the inability of these granules to traverse the steep edge pressure gradient region and reach the top of the pedestal prior to full ablation.

External n = 1 magnetic fields are applied in RFX-mod and DIII-D low safety factor Tokamak plasmas to investigate their interaction with the internal MHD dynamics and in particular with the sawtooth instability. In these experiments the applied magnetic fields cause a reduction of both the sawtooth amplitude and period, leading to an overall stabilizing effect on the oscillations. In RFX-mod sawteeth eventually disappear and are replaced by a stationary m = 1, n = 1 helical equilibrium without an increase in disruptivity. However toroidal rotation is significantly reduced in these plasmas, thus it is likely that the sawtooth mitigation in these experiments is due to the combination of the helically deformed core and the reduced rotation. The former effect is qualitatively well reproduced by nonlinear MHD simulations performed with the PIXIE3D code. The results obtained in these RFX-mod experiments motivated similar ones in DIII-D L-mode diverted Tokamak plasmas at low q(95). These experiments succeeded in reproducing the sawtooth mitigation with the approach developed in RFX-mod. In DIII-D this effect is correlated with a clear increase of the n = 1 plasma response, that indicates an enhancement of the coupling to the marginally stable n = 1 external kink, as simulations with the linear MHD code IPEC suggest. A significant rotation braking in the plasma core is also observed in DIII-D. Numerical calculations of the neoclassical toroidal viscosity (NTV) carried out with PENT identify this torque as a possible contributor for this effect.

A newly installed Lithium Granule Injector (LCil) was used to pace edge localized modes (ELM) in D111-D. ELM pacing efficiency was studied injecting lithium granules of nominal diameter 0.3-0.9mm, speed of 50-120 m s(-1) and average injection rates up to 100 Hz for 0.9 mm granules and up to 700 Hz for 0.3 mm granules. The efficiency of ELM triggering was found to depend strongly on size of the injected granules, with triggering efficiency close to 100% obtained with 0 9mm diameter granules, lower with smaller sizes, and weakly depending on granule velocity. Robust ELM pacing was demonstrated in ITER-1. e plasmas for the entire shot length, at ELM frequencies 3-5 times larger than the 'natural' ELM frequency observed in reference discharges. Within the range of ELM frequencies obtained, the peak ELM heat flux at the outer strike point was reduced with increasing pacing frequency. The peak heat flux reduction at the inner strike point appears to saturate at high pacing frequency. Lithium was found in the plasma core, with a concurrent reduction of metallic impurities and carbon. Overall, high frequency ELM pacing using the lithium granule injection appears to be compatible with both H -mode energy confinement and attractive H -mode pedestal characteristics, hut further assessment is needed to determine whether the projected heat flux reduction required for LEER can be met.

A long-term energy option that is just approaching the horizon after decades of struggle, is fusion. Recent developments allow us to apply techniques from spin physics to advance its viability. The cross section for the primary fusion fuel in a tokamak reactor, D+T=>alpha+n, would be increased by a factor of 1.5 if the fuels were polarized. Simulations predict further non-linear power gains in large-scale machines such as ITER, due to increased alpha heating. These are significant enhancements that could lower the requirements needed to reach ignition and could be used to extend useful reactor life by compensating for neutron degradation. The potential realization rests on the survival of spin polarization for periods comparable to the energy containment time. Interest in polarized fuel options had an initial peak of activity in the 1980s, where calculations predicted that polarizations could in fact survive a plasma. However, concerns were raised regarding the cumulative impacts of fuel recycling from the reactor walls. In addition, the technical challenges of preparing and handling polarized materials prevented direct tests. Over the last several decades, this situation has changed dramatically. Detailed simulations of the ITER plasma have projected negligible wall recycling in a high power reactor. In addition, a combination of advances in three areas - polarized material technologies, polymer pellets developed for Inertial Confinement, and cryogenic injection guns developed for delivering fuel into the core of tokamaks - have matured to the point where a direct it in situ measurement is possible. A Jefferson Lab - DIII-D/General Atomics - University of Virginia collaboration is developing designs for a proof-of-principle polarization survival experiment using the isospin mirror reaction, D+3He=>alpha+p, at the DIII-D tokamak in San Diego.

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

The aim of executing Single Null discharges in RFX-mod operating as a Tokamak led to the design and implementation of a plasma shape feedback control system. A fully model-based approach was followed which allowed dealing with critical issues such as the presence of a conducting shell, the strong coupling of the poloidal field coils and the voltage limits of the power supplies. A Linear Quadratic regulator and a Kalman state estimator were designed and implemented in the real time MARTe framework together with an algorithm for the real-time plasma boundary reconstruction. The problem of a number of sensors along the poloidal direction adequate only for circular discharges was also successfully tackled. The development of the system and its performances in terms of tracking and disturbance rejection capability are presented in the paper. (C) 2016 Consorzio RFX. Published by Elsevier B.V. All rights reserved.

DIII-D studies of high-frequency ELM pacing by pellet injection were extended to ITER scenarios at low neutral beam torque and to the use of non-fuel materials, a technique that could potentially reduce the throughput to the pumping and fuel reprocessing systems. ELM peak heat flux mitigation was obtained with injection of D2 pellets at frequencies up to 90 Hz in low-torque ITER baseline scenario (IBS, q95=3.2, βN=1.7, ne,ped=7.5x10 m, Tinj<0.1 N-m). D2 pellet injection resulted in ELM pacing at frequencies 6-8x the natural ELM frequency (10 Hz). The resulting inner divertor peak heat flux was reduced by more than a factor of 10, with minor modification in the heat-flux footprint width. However, at the highest injection frequencies, changes of the pedestal parameters similar to gas puffing were observed, resulting in a 20% reduction of pedestal pressure. The feasibility of ELM pacing with non-fuel pellets was explored with the new Impurity Granule Injector, capable of injecting sub-millimeter granules of non-fuel materials (e.g. Li, C, B4C), at frequencies up to 200 Hz, with controllable speed (50-150 m/s) and selectable granule size. Experiments with Li granules 0.4-0.9 mm injected in plasmas at q95~4.6, ne,ped~4x10 m, βN=1.5 demonstrated full-shot ELM pacing, with a 3-5x increase of fELM over the natural ELM frequency, and a reduction of ELM peak heat flux ~1/fELM. However, in the lower q95~3.2, low-torque IBS, ELM pacing by Li granule injection did not translate into heat flux mitigation, with a fraction of paced ELMs showing peak heat flux similar to natural ELMs. The combined dataset provides a unique contribution to ELM control research, challenging the understanding of the relation between high frequency ELM pacing and heat flux mitigation.

The march towards electricity production through tokamaks requires the construction of new facilities and the inevitable replacement of the previous generation. There are, however, research topics that are better suited to the existing tokamaks, areas of great potential that are not sufficiently mature for implementation in high power machines, and these provide strong support for a balanced policy that includes the redirection of existing programs. Spin polarized fusion, in which the nuclei of tokamak fuel particles are spin-aligned and favorably change both the fusion cross-section and the distribution of initial velocity vectors of charged fusion products, is described here as an example of a technological and physics topic that is ripe for development in a machine such as the DIII-D tokamak. Such research and development experiments may not be efficient at the ITER-scale, while the plasma performance, diagnostic access, and collaborative personnel available within the United States’ magnetic fusion research program, and at the DIII-D facility in particular, provide a unique opportunity to further fusion progress.

Periods of edge localized mode (ELM)-free H-mode with increased pedestal pressure and width were observed in the DIII-D tokamak when density fluctuations localized to the region near the separatrix were present. Injection of a powder of 45 µm diameter lithium particles increased the duration of the enhanced pedestal phases to up to 350 ms, and also increased the likelihood of a transition to the enhanced phase. Lithium injection at a level sufficient for triggering the extended enhanced phases resulted in significant lithium in the plasma core, but carbon and other higher Z impurities as well as radiated power levels were reduced. Recycling of the working deuterium gas appeared unaffected by this level of lithium injection. The ion scale, kθρs ∼ 0.1–0.2, density fluctuations propagated in the electron drift direction with f ∼ 80 kHz and occurred in bursts every ∼1 ms. The fluctuation bursts correlated with plasma loss resulting in a flattening of the pressure profile in a region near the separatrix. This localized flattening allowed higher overall pedestal pressure at the peeling–ballooning stability limit and higher pressure than expected under the EPED model due to reduction of the pressure gradient below the ‘ballooning critical profile’. Reduction of the ion pressure by lithium dilution may contribute to the long ELM-free periods.

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

Periods of ELM free H-mode (up to 350ms) with increased pedestal pressure and width were observed in the DIII-D tokamak when lithium particles were injected into the plasma. These periods are associated with enhanced density fluctuations (named ‘Bursty Chirping Mode’ in the later discussion) localized to the pedestal region. The dramatically increased fluctuations are shown to drive outward crossfield particle transport, which reduced the pressure gradient just inside the separatrix. This allowed higher overall pedestal pressure (twice that of normal H-modes) because of the improved peeling-ballooning stability limit.