The mega amp spherical tokamak (MAST) was a low aspect ratio device (R/a = 0.85/0.65 similar to 1.3) with similar poloidal cross-section to other medium-size tokamaks. The physics programme concentrates on addressing key physics issues for the operation of ITER, design of DEMO and future spherical tokamaks by utilising high resolution diagnostic measurements closely coupled with theory and modelling to significantly advance our understanding. An empirical scaling of the energy confinement time that favours higher power, lower collisionality devices is consistent with gyrokinetic modelling of electron scale turbulence. Measurements of ion scale turbulence with beam emission spectroscopy and gyrokinetic modelling in up-down symmetric plasmas find that the symmetry of the turbulence is broken by flow shear. Near the non-linear stability threshold, flow shear tilts the density fluctuation correlation function and skews the fluctuation amplitude distribution. Results from fast particle physics studies include the observation that sawteeth are found to redistribute passing and trapped fast particles injected from neutral beam injectors in equal measure, suggesting that resonances between the m = 1 perturbation and the fast ion orbits may be playing a dominant role in the fast ion transport. Measured D-D fusion products from a neutron camera and a charged fusion product detector are 40% lower than predictions from TRANSP/NUBEAM, highlighting possible deficiencies in the guiding centre approximation. Modelling of fast ion losses in the presence of resonant magnetic perturbations (RMPs) can reproduce trends observed in experiments when the plasma response and charge-exchange losses are accounted for. Measurements with a neutral particle analyser during merging-compression start-up indicate the acceleration of ions and electrons. Transport at the plasma edge has been improved through reciprocating probe measurements that have characterised a geodesic acoustic mode at the edge of an ohmic L-mode plasma and particle-in-cell modelling has improved the interpretation of plasma potential estimates from ball-pen probes. The application of RMPs leads to a reduction in particle confinement in L-mode and H-mode and an increase in the core ionization source. The ejection of secondary filaments following type-I ELMs correlates with interactions with surfaces near the X-point. Simulations of the interaction between pairs of filaments in the scrape-off layer suggest this results in modest changes to their velocity, and in most cases can be treated as moving independently. A stochastic model of scrape-off layer profile formation based on the superposition of non-interacting filaments is in good agreement with measured time-average profiles. Transport in the divertor has been improved through fast camera imaging, indicating the presence of a quiescent region devoid of filament near the X-point, extending from the separatrix to psi(n) similar to 1.02. Simulations of turbulent transport in the divertor show that the angle between the divertor leg on the curvature vector strongly influences transport into the private flux region via the interchange mechanism. Coherence imaging measurements show counter-streaming flows of impurities due to gas puffing increasing the pressure on field lines where the gas is ionised. MAST Upgrade is based on the original MAST device, with substantially improved capabilities to operate with a Super-X divertor to test extended divertor leg concepts. SOLPS-ITER modelling predicts the detachment threshold will be reduced by more than a factor of 2, in terms of upstream density, in the Super-X compared with a conventional configuration and that the radiation front movement is passively stabilised before it reaches the X-point. 1D fluid modelling reveals the key role of momentum and power loss mechanisms in governing detachment onset and evolution. Analytic modelling indicates that long legs placed at large major radius, or equivalently low B at the target compared with the X-point arc more amenable to external control. With MAST Upgrade experiments expected in 2019, a thorough characterisation of the sources of the intrinsic error field has been carried out and a mitigation strategy developed.

The ELM wetted area is a key factor in the peak power load during an ELM, as it sets the region over which the ELM energy is deposited. The deposited heat flux at the target is seen to have striations in the profiles that are generated by the arrival of filaments ejected from the confined plasma. The effect of the filaments arriving at the target on the ELM wetted area, and the relation to the midplane mode number is investigated in this paper using infrared (IR) thermography and high speed visible imaging (>10 kHz). Type I ELMs are analysed, as these have the largest heat fluxes and are observed to have toroidal mode numbers of between 5 and 15. The IR profiles during the ELMs show clear filamentary structures that evolve during the ELM cycle. An increasing number of striations at the target is seen to correspond to an increase in the wetted area. Analysis shows that the ratio of the ELM wetted area to the inter-ELM wetted area, a key parameter for ITER, for the type I ELMs is between 3 and 6 for lower single null plasmas and varies with the ELM midplane mode number, as determined by visible measurements. Monte-Carlo modelling of the ELMs is used to understand the variation seen in the wetted area and the effect of an increased mode number; the modelling replicates the trends seen in the experimental data and supports the observation of increased toroidal mode number generating larger target ELM wetted areas. ITER is thought to be peeling unstable which would imply a lower ELM mode number compared to MAST which is peeling-ballooning unstable. The results of this analysis suggest that the lower n peeling unstable ELMs expected for ITER will have smaller wetted areas than peeling-ballooning unstable ELMs. A smaller wetted area will increase the level of ELM control required, therefore a key prediction required for ITER is the expected ELM mode number.

The transport of particles via intermittent filamentary structures in the private flux region (PFR) of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggest that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the PFR of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1–2 cm in diameter, but appear more elongated near the divertor target. The most probable toroidal quasi-mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a speed of 0.5–1.0 km/s. Probe measurements at the inner divertor target suggest that the fluctuations in the particle flux to the inner target are strongest in the private flux region, and that the amplitude and distribution of these fluctuations are insensitive to the electron density of the core plasma, auxiliary heating and whether the plasma is single-null or double-null. It is found that the e-folding width of the time-average particle flux in the PFR decreases with increasing plasma current, but the fluctuations appear to be unaffected. At the outer divertor target, the fluctuations in particle and power fluxes are strongest in the SOL.

Measurements of intensity fluctuation of light emission within the divertor volume of MAST provide strong evidence for the existence of filamentary structures within the private flux region (PFR). These filaments are observed in L-mode and H-mode confinement regimes. Correlation analysis of the camera data supports the hypothesis that the filaments observed in the line integrated camera data are genuinely within the PFR, as fluctuations at a given location in the PFR in the image are correlated with fluctuations elsewhere in the PFR, and these two regions are connected by field lines. The filaments appear to move from a position in the PFR of the inner divertor leg, moving towards the inner divertor target, whilst ejecting secondary blobs of plasma deeper into the PFR away from the separatrix. (C) 2014 Elsevier B.V. All rights reserved.

The transport of particles via intermittent filamentary structures in the private flux region of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggests that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the private flux region (PFR) of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1-2cm in diameter. The most probable toroidal mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a speed of 0.5-1.0km/s. Probe measurements at the inner divertor target suggest that the fluctuations in the particle flux to the inner target are strongest in the private flux region, and that the amplitude and distribution of these fluctuations are insensitive to the electron density of the core plasma, auxiliary heating and whether the plasma is single-null or double-null. It is found that the e-folding width of the time-average particle flux in the PFR decreases with increasing plasma current, but the fluctuations are unchanged. At the outer divertor target, the fluctuations in particle and power fluxes are strongest in the SOL.

The distribution of particles and power to plasma-facing components is of key importance in the design of next-generation fusion devices. Power and particle decay lengths have been measured in a number of MAST L-mode and H-mode discharges in order to determine their parametric dependencies, by fitting power and particle flux profiles measured by divertor Langmuir probes, to a convolution of an exponential decay and a Gaussian function. In all discharges analysed, it is found that exponential decay lengths mapped to the midplane are mostly dependent on separatrix electron density ne,sep0.65±0.15 L-mode, ne,sep0.76±0.19 H-mode) and plasma current Ip-0.36±0.11 L-mode, Ip-1.05±0.18 H-mode) (or parallel connection length). The widths of the convolved Gaussian functions have been used to derive an approximate diffusion coefficient, which is found to vary from 1 m2/s to 7 m2/s, and is systematically lower in H-mode compared with L-mode.

Results are presented from a series of dedicated experiments carried out on JET in tritium, DT, deuterium and hydrogen plasmas to determine the dependence of the H mode power threshold on the plasma isotopic mass. The Pthr ∝ Aeff-1 scaling is established over the whole isotopic range. This result makes it possible for a fusion reactor with a 50:50 DT mixture to access the H mode regime with about 20% less power than that needed in a DD mixture. Results on the first systematic measurements of the power necessary for the transition of the plasma to the type I ELM regime, which occurs after the transition to H mode, are also in agreement with the Aeff-1 scaling. For a subset of discharges, measurements of Te and Ti at the top of the profile pedestal have been obtained, indicating a weak influence of the isotopic mass on the critical edge temperature thought to be necessary for the H mode transition.

In current large tokamaks, non-intrinsic seeded impurities have been used to produce divertor power loads which would be considered acceptable when extrapolated to ITER. Many devices have achieved the goals of high fractional radiated powers, small frequent ELMs and detachment which are characteristic of radiative H mode regimes. The influence of divertor geometry on these characteristics is described. It has been a matter of concern that the Zeff associated with the seeded impurities may exceed that allowable in ITER and also that the degradation in energy confinement may be unacceptable. Confidence can only be built in the prediction of these parameters in ITER if reliable scalings are available for impurity content and energy confinement which have a sound physics basis. Work is described at JET in this area whilst using multimachine data to characterize the size scaling and provide a context for the JET data. Predicted levels for the impurity content of seeded ITER plasmas appear to be of marginal acceptability. Discharges run in the JET Mark I, Mark IIA and Mark IIAP divertors are compared and indicate that increased divertor closure has brought relatively minor benefits in highly radiative discharges. The acceptability of the energy confinement of radiation for ITER remains unclear. Dimensionless parameter scaling experiments have been conducted in which β, q25, fractional radiated power and Zeff are held constant for a range of ρ*. The price paid for high edge radiation and small ELMs appears to be a 25% loss in total stored energy as a result of edge pedestal degradation. However, the underlying energy confinement scaling may still be consistent with gyro-Bohm scaling, which would give an adequate margin for ITER. This conclusion is, however, sensitive to the scaling of confinement with collisionality, which is difficult to determine due to the coupling between ρ* and ν* which is a consequence of radiation dominated regimes.

Results of experiments investigating the performance of the JET Mark IIA divertor are reported and compared with the performance of its Mark I predecessor. The principal effect of reducing the divertor width (increasing closure) was to increase pumping for both deuterium and impurities while reducing upstream neutral pressure. Neither the orientation of the divertor target relative to the divertor plasma nor the width of the divertor had a major influence on core plasma performance in ELMy H modes. Changing the core triangularity and thus the edge magnetic shear modifies the ELM frequency in ELMy H mode plasmas, thereby changing the peak divertor power loading. The integrated performance of the core and divertor plasmas is reviewed with a view to extrapolation to the requirements of ITER. The confinement of JET ELMy H modes with hot, medium density edges is good (H97 ≈ 1) and follows a gyro-Bohm scaling. The impurity content of these discharges is low and within the ITER requirements. When an attempt is made to raise the density with deuterium gas fuelling, the ELM frequency increases and the confinement, especially in the edge, decreases. Good confinement can be achieved in JET either by producing a large edge pedestal, typically in discharges with NB heating or by centrally peaked heating with ICRH schemes. Large amplitude type I ELMs, which are present in all discharges with a large edge pedestal, would result in unacceptable divertor plate erosion when scaled to ITER. Since the power deposition profile due to α heating in ITER is calculated to be intermediate between the JET NB and RF heating profiles, it is likely that operation in ITER with small ELMs in order to reduce first wall loading will result in degraded confinement compared with present day scaling laws.

On the basis of experiments carried out in JET in D:T mixtures varying from 100:0 to 5:95 and those carried out in hydrogen plasmas, the isotopic mass dependence of ELM parameters and the edge pedestal pressure in NBI and ICRF heated H mode plasmas is presented. The ELM frequency is found to decrease with the atomic mass number in both ICRH and NBI discharges. However, the frequency in the case of ICRH is about 8-10 times higher than that in the case of NBI. Assuming that ELMs occur at a critical edge pressure gradient, limited by the ballooning instability, the scaling of the maximum edge pressure is most consistent with the assumption that the width of the transport barrier scales as the ion poloidal Larmor radius governed by the average energy of fast ions at the edge. The critical edge pressure in NBI heated discharges increases with the isotopic mass, which is consistent with the higher deduced width of the edge transport barrier in tritium than those in deuterium and hydrogen. The critical edge pressure in ICRH discharges is smaller, presumably, due to the smaller fast ion contribution to the edge region. As a consequence of the edge pressure scaling with isotopic mass, the edge operational space in an n(e)-T-e diagram increases with operation in tritium. If the evidence that the edge pedestal width is governed by the average energy of fast ions in the edge prevails, the pedestal in ITER would be controlled by the slowing down energy spectrum of alpha particles in the edge.

Because of its large size, single null divertor and flexible magnetic geometry, JET is capable of producing the most reactor relevant plasmas of any present generation tokamak. In recent DT experiments, the fusion performance of these plasmas was tested for the first time. Over 4 MW of fusion power was produced in a high power, steady state pulse of 5 s, limited by the duration of the heating power. The fusion Q(E), defined simply as the fusion energy produced divided by the input energy over this 5 s interval, was 0.18. These DT ELMY H mode discharges performed up to expectations based on DD preparation pulses and thus establish a firm basis for extrapolating to a next step machine. Operation at low q(95) is possible in JET with no degradation in the confinement enhancement factor and provides an improved margin to ignition when extrapolated to ITER. Considerable uncertainties remain, nonetheless. In particular, access to high density, relative to the Greenwald limit, and operation in close proximity to the H mode threshold may both result in a degradation of the confinement in the next step machine.

An overview of JET experimental results in DT plasmas directly relevant to ITER modes of operation is presented. Experiments in D:T mixtures varying from 100:0 to 10:90 and those carried out in hydrogen plasmas show that the H mode threshold power has an approximately inverse isotope mass dependence. Matching some of the key dimensionless parameters to the ITER values, the ITER similarity experiments with ITER shape and safety factor q show that the global energy confinement time is practically independent of isotopic mass (~A0.03±0.08), where A is the atomic mass of the hydrogenic species. Subtracting the edge pedestal energy (which scales as ~A0.57±0.2) from the total stored energy leads to a ~A-0.17±0.1 dependence of confinement in the plasma core, very similar to that expected from the gyro-Bohm transport (~ A-0.2) model. The observed scaling of the edge pedestal energy is consistent with a model in which the edge pressure gradient saturates at the ballooning limit over a region of width that scales as the ion poloidal Larmor radius governed by the average energy of the fast ions in the edge. The steady state total stored energy for a given input power in both ICRH and NBI discharges is the same despite the lower edge pedestal in the ICRH case, which is compensated for by more peaked power deposition profiles in ICRH. The ELM frequency is smaller with NBI; it decreases with isotopic mass in both NBI and ICRH discharges. A steady state, type I ELMy H mode discharge with ITER shape and q at 3.8 T/3.8 MA with an input power of 22 MW produced a Q ≈ 0.18 for 3.5 s and extrapolates well to ignition with ITER parameters. Here, Q is the ratio of fusion output power to input power. The thermal ELMy H mode confinement in both deuterium and tritium gas fuelled plasmas decreases significantly when the plasma density exceeds 0.75 of the Greenwald (nGW) limit, and the maximum density achieved is 0.85nGW. In L mode, the density limit decreases with increasing isotope mass roughly in accordance with code predictions. ITER reference ICRH scenarios have been evaluated. Second harmonic heating of tritium at the densities available in JET produces strong tails and heats electrons predominantly as expected. The 3He minority in 50:50 D:T and tritium dominated plasmas showed strong bulk ion heating leading to ion temperatures up to 13 keV with ICRH alone. Deuterium minority ion cyclotron heating in tritium plasmas at a power level of 6 MW produced steady state record values of Q ≈ 0.22 for more than 2.5 s.

There are various conditions in JET under which the energy confinement of a plasma is reduced from H mode towards L mode levels. Examples are reported here for which the loss of confinement is associated with a reduction of the electron pressure pedestal. This loss is brought about by strong gas puffing that increases the frequency of edge localized modes (ELMs). The effect of ELMs on confinement is modelled, and used to explain the observed degradation of confinement. The model is generalized for application to other machines.

Giant edge localized modes (ELMs) play an important role in the transport of plasma at the edge of tokamaks with a divertor. The loss of plasma, by mechanisms that are not well understood, causes very high power loading on parts of the wall structure. The flows of particles and energy have been determined on a rapid time-scale for the first time in JET. The soft X ray measurements have a complex structure and show a rapid onset of the instability. They prove that some plasma is lost outside the divertor and that recycling occurs to well within the separatrix. The implications of the results for the ELM instability mechanism and the design of tokamaks are discussed.

A non-rotating helical magnetic field perturbation grows in the JET plasma when the electron density falls below a minimum value. The growth of this perturbation usually leads to a disruption. The helical phase of the perturbation is reproducible for a given configuration of the applied poloidal and toroidal magnetic fields. Small asymmetries in the applied magnetic fields are shown to cause the growth of the perturbation. Various sources of such field errors in JET have been identified. The existence of a minimum density for a stable plasma can place a severe constraint on the operation of JET. The magnitude of the minimum density depends upon various plasma parameters. Neutral beam injection can be used to reduce the minimum density

Magnetic feedback control has been used in the DITE tokamak to substantially reduce saturated m=2, n=1 instabilities in both Ohmic discharges and discharges with lower hybrid current drive (LHCD). Feedback has been used for the first time to significantly increase the disruptive density limit in a tokamak. LHCD on DITE stabilizes the sawtooth instability but generates a large m=1,2, n=1 instability. Both components of this mode have been controlled for the first time with feedback. Open loop experiments with LHCD show that mode locking occurs and allow detailed study of this phenomenon.