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Major developments on the Mega Amp Spherical Tokamak (MAST) have enabled important advances in support of ITER and the physics basis of a spherical tokamak (ST) based component test facility (CTF), as well as providing new insight into underlying tokamak physics. For example, L–H transition studies benefit from high spatial and temporal resolution measurements of pedestal profile evolution (temperature, density and radial electric field) and in support of pedestal stability studies the edge current density profile has been inferred from motional Stark effect measurements. The influence of the q-profile and E × B flow shear on transport has been studied in MAST and equilibrium flow shear has been included in gyro-kinetic codes, improving comparisons with the experimental data. H-modes exhibit a weaker q and stronger collisionality dependence of heat diffusivity than implied by IPB98(y,2) scaling, which may have important implications for the design of an ST-based CTF. ELM mitigation, an important issue for ITER, has been demonstrated by applying resonant magnetic perturbations (RMPs) using both internal and external coils, but full stabilization of type-I ELMs has not been observed. Modelling shows the importance of including the plasma response to the RMP fields. MAST plasmas with q > 1 and weak central magnetic shear regularly exhibit a long-lived saturated ideal internal mode. Measured plasma braking in the presence of this mode compares well with neo-classical toroidal viscosity theory. In support of basic physics understanding, high resolution Thomson scattering measurements are providing new insight into sawtooth crash dynamics and neo-classical tearing mode critical island widths. Retarding field analyser measurements show elevated ion temperatures in the scrape-off layer of L-mode plasmas and, in the presence of type-I ELMs, ions with energy greater than 500 eV are detected 20 cm outside the separatrix. Disruption mitigation by massive gas injection has reduced divertor heat loads by up to 70%.

Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

Results of three-dimensional (3D) numerical simulations of field-reversed configurations (FRCs) are presented. The emphasis of this work is on the nonlinear evolution of magnetohydrodynamic (MHD) instabilities in kinetic FRCs and the new FRC formation method by counter-helicity spheromak merging. Kinetic simulations show nonlinear saturation of the n = 1 tilt mode, where n is the toroidal mode number. The n = 2 and n = 3 rotational modes are observed to grow during the nonlinear phase of the tilt instability due to the ion spin-up in the toroidal direction. The ion toroidal spin-up is shown to be related to the resistive decay of the internal flux and the resulting loss of particle confinement. Three-dimensional MHD simulations of counter-helicity spheromak merging and FRC formation show good qualitative agreement with the results from the SSX–FRC experiment. The simulations show the formation of an FRC in about 20–30 Alfven times for typical experimental parameters. The growth rate of the n = 1 tilt mode is shown to be significantly reduced compared with the MHD growth rate due to the large plasma viscosity and field-line-tying effects.

Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e. the degree to which the divertor topology is single-null or double-null (DN) and (2) the direction of the of B × ∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the B × ∇B ion particle drift direction. Our data suggests that the presence of B × ∇B and E × B ion particle drifts in the scrape-off layer and divertor(s) play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density. These results have implications for particle control in ITER and other future tokamaks.

This paper describes a new approach for modelling the pedestal energy transport in the presence of a small radial magnetic perturbation. The cases of a ballooning instability leading to Type I edge localized modes (ELMs) and a magnetic perturbation generated by external coils are treated. The model for Type I ELMs is based on the linear ideal MHD code MISHKA coupled with the non-linear energy transport code TELM in a realistic tokamak geometry. The main mechanism of the increased transport through the external transport barrier in this model of ELMs is due to the appearance of a radial velocity and a radial magnetic field perturbation due to the MHD mode. Both lead to additional transport perpendicular to the magnetic surface and hence to a relaxation of the pressure profile in the unstable zone. The typical Type I ELM time-cycle was reproduced numerically including the destabilization of the ballooning modes leading to the fast (250 μ s) collapse of the pedestal pressure followed by the edge pressure profile re-building on a diffusive time scale. A possible mechanism for the control of Type I ELMs using a stochastic plasma boundary created by external coils is modelled in this paper using data on ELM suppression by I-coils from the DIII-D experiment. In the stochastic layer the transverse transport is effectively increased by diffusion of the magnetic field lines. The modelling results demonstrate the possibility of decreasing the edge pressure gradient to a value that is just below the ideal ballooning limit, leading to a high confinement regime without Type I ELMs.

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfven eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

Large sub-millisecond heat pulses due to Type-I edge localized modes (ELMs) have been eliminated reproducibly in DIII-D for periods approaching nine energy confinement times (tau(E)) with small dc currents driven in a simple magnetic perturbation coil. The current required to eliminate all but a few isolated Type-l ELM impulses during a coil pulse is less than 0.4% of plasma current. Based on magnetic field line modelling, the perturbation fields resonate with plasma flux surfaces across most of the pedestal region (0.9 4-6 tau(E)) have been reproduced numerous times, on multiple experimental run days in high and intermediate triangularity plasmas, including cases matching the baseline ITER scenario 2 flux surface shape. In low triangularity, lower single null plasmas, with collisionalities near that expected in ITER, Type-l ELMs are replaced by small amplitude, high frequency Type-II-like ELMs and are often accompanied by one or more ELM-free periods approaching 1-2 tau(E). Large Type-I ELM impulses represent a severe constraint on the survivability of the divertor target plates in future burning plasma devices. Results presented in this paper demonstrate that non-axisymmetric edge magnetic perturbations provide a very attractive development path for active ELM control in future tokamaks such as ITER.

This work investigates the potential roles played by the scrape-off-layer current (SOLC) in MHD activity of tokamak plasmas, including effects on stability. SOLCs are found to be an integral part of the MHD activity, with a propensity to flow in a non-axisymmetric pattern and with magnitude potentially large enough to play a role in the MHD stability. Candidate mechanisms that can drive these SOLCs are identified: (a) non-axisymmetric thermoelectric potential, (b) electromotive force from MHD activity, and (c) flux swing, both toroidal and poloidal, of the plasma column. Other potential magnetic consequences of the SOLC are identified: (a) its error field can introduce complications in feedback control schemes for stabilizing MHD activity, and (b) its non-axisymmetric field can be falsely identified as an axisymmetric field by the tokamak control logic and in equilibrium reconstruction. The radial profile of an SOLC observed during a quiescent discharge period is determined, and found to possess polarity reversals as a function of radial distance.

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7?MW) and high harmonic fast wave (HHFW) heating (up to 6?MW), toroidal fields up to 6?kG, plasma currents up to 1.5?MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10?MW?m?2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400?kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks, leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardware-related fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments (Scoville J.T. and La Haye R.J. 2003 Nucl. Fusion 43 250) and resistive wall mode experiments (Garofalo A.M., La Haye R.J. and Scoville J.T. 2002 Nucl. Fusion 42 1335) done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made in DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported (La Haye R.J. and Scoville J.T. 1991 Rev. Sci. Instrum. 61 2146). Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the Ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all the coils are one-third the values indicated from the stability experiments (Garofalo et al 2002, Scoville and La Haye 2003). These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks.

Details of the low-high (L-H) confinement transition are studied by a new technique which allows for an arbitrarily slow transition between the L- and H-modes on DIII-D. During the transition, the plasma is in an intermediate state (IM-mode) of temperature, density, confinement and edge shear flow. The IM-mode is characterized by periodic bursts of an edge instability, governed by relaxation oscillations, which evolve into type III edge localized modes (ELMs) as the neutral beam heating power is raised. An ELM-free H-mode is achieved when the edge pressure gradient is large enough to support shear flow sufficient to quell the IM-mode edge instability.

Densities of up to 40% above the Greenwald limit are reproducibly achieved in high confinement (HITER89P = 2) ELMing H mode discharges. Simultaneous gas fuelling and divertor pumping were used to obtain these results. Confinement of these discharges, similar to moderate density H mode, is characterized by a stiff temperature profile, and is therefore sensitive to the density profile. A particle transport model is presented that explains the roles of divertor pumping and geometry for access to high densities. The energy loss per ELM at high density is a factor of five lower than the predictions of an earlier scaling, based on data from lower density discharges.

Particle pumping characteristics were compared between pumping from the inner side private flux region (IPP) and pumping from both sides of the private flux region (BPP) in the JT-60U W shaped divertor, and between JT-60U IPP and pumping in the DIII-D lower baffled divertor. The pumping flux for BPP is smaller than that for IPP by about a factor of 2 with weak in-out asymmetry of recycling neutral flux and by a factor of 3.5-6.5 with strong in-out asymmetry. The reduction of the pumping flux for BPP is consistent with Monte Carlo simulations, where backflow at the outer pumping slot is observed due to in-out recycling asymmetry. The pumping flux in DIII-D at Ip = 0.8 MA and BT = 1.6 T is comparable to or smaller than that for JT-60U IPP at Ip = 1.0 MA, BT = 3.8 T and Ip = 1.5 MA, BT = 3.5 T in the same density regime. In the DIII-D divertor with pumping from the private flux region, the pumping flux decreases with increasing in-out asymmetry. The pumping flux normalized by the integrated Dα emission over the whole plasma exhibits a similar dependence on the distance between the pumping slot and the strike point in JT-60U IPP and the DIII-D lower divertor with pumping through the outer divertor plasma region.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

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

Recent progress towards an increased understanding of the physical processes in the divertor and scrape-off layer (SOL) plasmas in DIII-D has been made possible by a combination of new diagnostics, improved computational models and changes in divertor geometry. The work focused primarily on ELMing H mode discharges. The physics of partially detached divertor plasmas, with divertor heat flux reduction by divertor radiation enhancement using D2 puffing, was studied in two dimensions, and a model of the heat and particle transport was developed that includes conduction, convection, ionization, recombination and flows. Plasma and impurity particle flows were measured with Mach probes and spectroscopy and compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation was increased in the divertor and SOL with `puff and pump' techniques using SOL D2 puffing, divertor cryopumping and argon puffing. The important physical processes in plasma-wall interactions were examined with a DiMES (divertor material evaluation system) probe, plasma characterization near the divertor plate and the REDEP code. Experiments comparing single null plasma operation in baffled and open divertors demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H mode with pumping and baffling resulted in a reduction in H mode core densities to ne/nGr ≈ 0.25 (with nGr the Greenwald density). Divertor particle exhaust and heat flux were studied as the plasma shape was varied from a lower single null to a balanced double null, and finally to an upper single null.

A series of controlled experiments has been carried out in DIII-D to induce a bulk ion flow in the SOL and evaluate its effect on the localization of impurities in the divertor. This induced SOL flow was created by simultaneous deuterium puffing and divertor exhaust using a divertor cryopump, and the impurity enrichment was measured directly. The experiments were designed to compare enrichment in discharges with and without induced flow having otherwise similiar divertor parameters. Significant increases in impurity compression and enrichment are observed when flow is induced, and the degree of impurity enrichment in the divertor is found to be dependent on the impurity of interest. Detailed particle measurements made possible by the direct measurement of impurity densities in several reservoirs indicate reasonable particle balance for helium throughout the duration of the discharge. Conversely, while the total input of neon is balanced by the total exhaust by the end of a discharge, particle balance is not observed during the course of the discharge. A significant wall inventory with a short release time (~10 ms) is surmised.

{open_quotes}Slow{close_quotes} liner fusion ({approximately}10 ms compression time) implosions are nondestructive and make repetitive ({approximately} 1 Hz) pulsed liner fusion reactors possible. This paper summarizes a General Atomics physics-based fusion reactor study that showed slow liner feasibility, even with conservative open-line axial magnetic field confinement and Bohm radial transport.

The particle exhaust fate of the DIII-D tokamak divertor cryopump system was investigated as a function of gas-puff-regulated core plasma line-averaged density, divertor electrical bias potential and position of the divertor strike point with respect to the pump entrance aperture. The experiment was done with rapidly ELMing H mode plasmas. Divertor bias approximately doubled the maximum achievable exhaust rate and also reduced its sensitivity to the strike point position. The achievable exhaust rate varied roughly as the third power of the core plasma density

Reduction of the net wall particle inventory has been achieved in the DIII-D tokamak by the use of a divertor cryopump for the particle exhaust. A sequence of 12 discharges was conducted without helium glow discharge cleaning (HeGDC) and without active cryopump exhaust, resulting in a net wall loading of 1250 torr.L (8.8*1022 atoms) by the end of the sequence. The cryopump was activated for the subsequent set of 10 discharges. At the end of this set of discharges, the net wall loading was reduced to its initial value or less, i.e. the wall loading state for the reference discharges with HeGDC. Plasma performance did not suffer without the HeGDC in that the stored energy during the ELMy phase was restored to the reference discharge level during the discharges with the active cryopump. Hence, with continuous particle exhaust (provided, for example, by a pumped divertor), next generation, long pulse devices, such as ITER, will not require interdischarge HeGDC for particle control, which would require turning off superconducting coils between discharges

Results are presented from an experiment in which a large amount of deuterium gas was continuously puffed at the scrape-off layer and exhausted from the DIII-D pump divertor ('puff and pump' technique). The experiments were conducted with a single null divertor configuration during edge localized mode (ELM)ing H mode. Steady state density conditions were attained during the strong deuterium puff. The deuterium puff reduced the central plasma density of argon, puffed continuously as a trace impurity, by as much as a factor of 20, and the central argon radiated power density by as much as a factor of 10

The open divertor in DIII-D has been modified by the installation of a baffled plenum with a narrow entrance aperture. Neutral particle pressures in the range of 2-20 mtorr have been measured in this new plenum during experiments in L and H mode plasmas in which the neutral beam power was varied and the divertor target location was scanned with respect to the entrance aperture. These pressures exceed the predicted minimum required to provide adequate particle exhaust for controlling the density in H mode discharges when pumping experiments are conducted in the future. The pressure measurements presented here, which were carried out in the absence of a pump, were modelled with the aid of two transport codes: an edge plasma code and a neutrals code. It was found in this modelling process that the divertor flux amplification factor required to explain the measured pressures is of the order of 30 to 40, much higher than the factors used in the modelling of data in the absence of the baffle. This higher flux amplification factor results from increased recycling at the divertor owing to the presence of the baffle

A toroidal ring bias electrode in contact with the single-null divertor scrape-off layer in the DIII-D tokamak achieved electric field control of plasma flow in the layer. Bias strongly altered the static pressure of neutralized gas in the new divertor chamber (designed for future pumping) during Ohmic, L-mode and ELMing H-mode plasma operation. The pressure increased or decreased, depending on the signs of the applied electric potential and toroidal magnetic field, in agreement with the E*B drift direction. The variation of the core plasma density was opposite to the variation of the pump chamber gas pressure and was proportionally much smaller. A large bias induced pressure rise appeared even when the scrape-off layer was not directed at the pump entrance aperture-a surprising effect that is interpreted qualitatively in terms of radial potential discontinuities. Bias also affected divertor heat fluxes, and preliminary observations are discussed qualitatively

A prototype tile current array, consisting of resistors to sense current flowing from six divertor tiles into the metallic vacuum vessel, was installed in DIII-D. Data show that typically 1-2 kA circulates poloidally around the single-null divertor scrapeoff layer (SOL). A current of similar magnitude flows vertically in the outer SOL of double-null plasmas and a much smaller current in the inner SOL. Little or no current flows across the short SOL connecting the inner and outer divertor strike points. The SOL current increases with core plasma current and edge localized modes (ELMs); it decreases at the onset of the H-mode and with heavy gas puffing. The SOL current jumps rapidly to 5-10 kA during giant ELM events. Three possible mechanisms to drive the SOL current are examined qualitatively with respect to the data, but the evidence is insufficient to identify the dominant one. During disruptions, poloidal current in excess of 100 kA flows to the wall.

The ideal and resistive MHD free-boundary stability of m = 0, 1, 2 and 3 modes is calculated for the circular force-free reversed field pinch (RFP). Functions ?(r)= ?0?/B2 specifying the equilibrium configurations are chosen to approximate the deviations of experimental RFPs from the fully relaxed state. Families of unstable modes are identified. All are resistive in the usual RFP parameter range, unless the distance to the conducting shell is very large. The usually stable m = 2 and 3 modes, as well as m = 0 and m = 1, are unstable near the plasma edge for practical shell separations, and these modes might promote greater edge relaxation than if plasma and shell were in contact. When a small central peaking of ?(r) is present, the unstable mode spectrum hardly varies as the shell is moved outward. It is suggested that extreme closeness of the shell to the plasma might not be required. Moderate plasma-shell separation might also eliminate sawtoothing at high reversals. Supporting evidence from OHTE and HBTX 1C experiments is cited.

The nonlinear response of a low-beta tokamak plasma to non-axisymmetric fields offers an alternative to direct measurement of the non-axisymmetric part of the vacuum magnetic fields, often termed ‘error fields’. Possible approaches are discussed for determination of error fields and the required current in non-axisymmetric correction coils, with an emphasis on two relatively new methods: measurement of the torque balance on a saturated magnetic island, and measurement of the braking of plasma rotation in the absence of an island. The former is well suited to ohmically heated discharges, while the latter is more appropriate for discharges with a modest amount of neutral beam heating to drive rotation. Both can potentially provide continuous measurements during a discharge, subject to the limitation of a minimum averaging time. The applicability of these methods to ITER is discussed, and an estimate is made of their uncertainties in light of the specifications of ITER’s diagnostic systems. The use of plasma response-based techniques in normal ITER operational scenarios may allow identification of the error field contributions by individual central solenoid coils, but identification of the individual contributions by the outer poloidal field coils or other sources is less likely to be feasible.

Infrared imaging of hot spots induced by localized magnetic perturbations using the test blanket module (TBM) mock-up on DIII-D is in good agreement with beam-ion loss simulations. The hot spots were seen on the carbon protective tiles surrounding the TBM as they reached temperatures over 1000 °C. The localization of the hot spots on the protective tiles is in fair agreement with fast-ion loss simulations using a range of codes: ASCOT, SPIRAL and OFMCs while the codes predicted peak heat loads that are within 30% of the measured ones. The orbit calculations take into account the birth profile of the beam ions as well as the scattering and slowing down of the ions as they interact with the localized TBM field. The close agreement between orbit calculations and measurements validate the analysis of beam-ion loss calculations for ITER where ferritic material inside the tritium breeding TBMs is expected to produce localized hot spots on the first wall. © 2013 IAEA, Vienna.

In-vessel, non-axisymmetric, control coils have proven to be an important option for mitigating and suppressing edge-localized modes (ELMs) in high performance operating regimes on a growing number of tokamaks. Additionally, an in-vessel non-axisymmetric ELM control coil is being considered in the ITER baseline design. In preparing for the initial operation of this coil set, a comprehensive study was carried out to characterize the linear superposition of the 3D vacuum magnetic field, produced by the ELM coil, on a series of equilibria representing nine standard ITER operating scenarios. Here, the spatial phase angle of toroidally distributed currents, specified with a cosine waveform, in the upper and lower rows of the ITER ELM coil (IEC) set is varied in 2? steps while holding the current in the equatorial row of coils constant. The peak current in each of the three toroidal rows of window-frame coils making up the IEC is scanned between 5?kAt and 90?kAt in 5?kAt steps and the width of the edge region covered by overlapping vacuum field magnetic islands is calculated. This width is compared to a vacuum field ELM suppression correlation criterion found in DIII-D. A minimum coil current satisfying the DIII-D criterion, along with an associated set of phase angles, is identified for each ITER operating scenario. These currents range from 20?kAt to 75?kAt depending on the operating scenario being used and the toroidal mode number (n) of the cosine waveform. Comparisons between the scaling of the divertor footprint area in cases with n?=?3 perturbation fields versus those with n?=?4 show significant advantages when using n?=?3. In addition, it is found that the DIII-D correlation criterion can be satisfied in the event that various combinations of individual IEC window-frame coils need to be turned off due to malfunctioning components located inside the vacuum vessel. Details of these results for both the full set of 27 window-frame coils and various reduced sets, using either n?=?3 and n?=?4 perturbation fields, are discussed.

Operating ITER in the reference inductive scenario at the design values of Ip = 15 MA and QDT = 10 requires the achievement of good H-mode confinement that relies on the presence of an edge transport barrier whose pedestal pressure height is key to plasma performance. Strong gradients occur at the edge in such conditions that can drive magnetohydrodynamic instabilities resulting in edge localized modes (ELMs), which produce a rapid energy loss from the pedestal region to the plasma facing components (PFC). Without appropriate control, the heat loads on PFCs during ELMs in ITER are expected to become significant for operation in H-mode at Ip = 6–9 MA; operation at higher plasma currents would result in a very reduced life time of the PFCs. Currently, several options are being considered for the achievement of the required level of ELM control in ITER; this includes operation in plasma regimes which naturally have no or very small ELMs, decreasing the ELM energy loss by increasing their frequency by a factor of up to 30 and avoidance of ELMs by actively controlling the edge with magnetic perturbations. Small/no ELM regimes obtained by influencing the edge stability (by plasma shaping, rotational shear control, etc) have shown in present experiments a significant reduction of the ELM heat fluxes compared to type-I ELMs. However, so far they have only been observed under a limited range of pedestal conditions depending on each specific device and their extrapolation to ITER remains uncertain. ELM control by increasing their frequency relies on the controlled triggering of the edge instability leading to the ELM. This has been presently demonstrated with the injection of pellets and with plasma vertical movements; pellets having provided the results more promising for application in ITER conditions. ELM avoidance/suppression takes advantage of the fact that relatively small changes in the pedestal plasma and magnetic field parameters seem to have a large stabilizing effect on large ELMs. Application of edge magnetic field perturbation with non-axisymmetric fields is found to affect transport at the plasma edge and thus prevent the uncontrolled rise of the plasma pressure gradients and the occurrence of type-I ELMs. This paper compiles a brief overview of various ELM control approaches, summarizes their present achievements and briefly discusses the open issues regarding their application in ITER.

The energetic ion loss has been assessed using the F3D-OFMC code for a 15 MA inductive scenario with Q = 10 and the latest information on the first wall geometry, the implementation of ferritic inserts (FI) and the ELM mitigation/control coils. Alpha particles and NB ions generated by the neutral beam injectors with the injection energy of 1 MeV are well confined and the heat load on the first wall is negligibly small and allowable for the magnetic background by the toroidal field coils and FI. However, an increase in the loss of these energetic ions is observed when the magnetic field by the ELM coils is applied. The increase in the loss fraction is larger for NB ions than for alpha particles under the ELM coil field. The origin of the expelled NB ions is dominantly trapped ions generated in the peripheral region due to a high-density plasma of the 15 MA scenario.

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The non-linear reduced four-field RMHD model in cylindrical geometry was extended to include plasma rotation, neoclassical poloidal viscosity and two fluid diamagnetic effects. Interaction of the static resonant magnetic perturbations (RMPs) with the rotating plasmas in tokamaks was studied. The self-consistent evolution of equilibrium electric field due to RMP penetration is taken into account in the model. It is demonstrated that in the pedestal region with steep pressure gradients, mean flows perpendicular to the magnetic field, which includes and electron diamagnetic components plays an essential role in RMP screening by plasma. Generally, the screening effect increases for lower resistivity, stronger rotation and smaller RMP amplitude. Strong screening of central islands was observed limiting RMP penetration to the narrow region near the separatrix. However, at certain plasma parameters and due to the non-linear evolution of the radial electric field produced by RMPs, the rotation can be compensated by electron diamagnetic rotation locally. In this case, RMPs can penetrate and form magnetic islands. Typical plasma parameters and RMPs spectra on DIII-D, JET and ITER were used in modelling examples presented in the paper.

Fast beam-ion losses were studied in DIII-D in the presence of a scaled mock-up of two test blanket modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot, predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot, which is predicted to be different among the various codes. © 2011 IAEA, Vienna.