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The ITER baseline scenario, with 500 MW of DT fusion power and Q = 10, will rely on a Type I ELMy H-mode and will be achieved with a tungsten (W) divertor. W atoms sputtered from divertor targets during mitigated ELMs are expected to be the dominant source in ITER. W impurity concentration in the plasma core can dramatically degrade its performance and lead to potentially damaging disruptions. Understanding the physics of the target W source due to sputtering during ELMs and inter-ELMs is important and can be helped by experimental measurements with improved precision. It has been established that the ELMy target ion impact energy has a simple linear dependence with the pedestal electron temperature measured by Electron Cyclotron Emission (ECE). It has also been shown that Langmuir Probes (LP) ion flux measurements are reliable during ELMs due to the surprisingly low electron temperature. Therefore, in this paper, LP and ECE measurements in JET-ITER-Like-Wall (ILW) unseeded Type I ELMy H-mode experiments have been used to estimate the W sputtering flux from divertor targets in ELM and inter-ELM conditions. Comparison with similar estimates using W I spectroscopy measurements shows a reasonable agreement for the ELM and inter-ELM W source. The main advantage of the method involving LP measurements is the very high time resolution of the diagnostic (∼10 μs) allowing very precise description of the W sputtering source during ELMs. EURATOM 633053 MSMT INGO LG14002 Fundação para a Ciência e Tecnologia UID/FIS/50010/2013

The effects of applied three-dimensional magnetic error fields on magnetic surfaces are measured in both the Joint European Torus (JET) and the Mega Ampere Spherical Tokamak (MAST). Static perturbation fields with various amplitudes have been applied in both devices using non-axisymmetric coils producing a dominantly n = 1 magnetic field to systematically perturb the plasma flux surfaces. The displacement of the plasma boundary is found to increase linearly with the strength of the applied field, and agrees well with predictive modelling which superimposes vacuum non-axisymmetric fields on a 2D equilibrium.

Results are presented from the JET Trace Tritium Experimental (TTE) campaign using minority tritium (T) plasmas (n(T)/n(D) 2 MA) and monotonic q-profiles. In CH discharges the gamma-ray emission decay times are much lower than classical (tau(Ts) + tau(alpha s)), indicating alpha confinement degradation, due to the orbit losses and particle orbit drift predicted by a 3-D Fokker-Planck numerical code and modelled using TRANSP.

Results from recent experiments to study the effects of divertor geometry and increased plasma shaping on the L–H transition threshold on JET are reported. Equivalent septum configurations run with the new septum replacement plate (SRP) in the MkII Gas Box divertor have shown that the presence of the septum lowers the L–H transition power threshold, Pth, by 20%. For X-point to virtual septum top distances of less than 6 cm, the SRP plasmas also demonstrate a significant decrease of the L–H Pth and pedestal electron temperature, Te with reduced X-point height. Although, the SRP plasma's Pth remains above that with the septum, there is no difference in the pedestal Te at the L–H transition. The influence of plasma shaping on the L–H transition has also been investigated for the first time on JET in a series of density scans at Ip/Bt of 2.5 MA/2.7 T. While keeping the lower triangularity, δlower, and divertor geometry constant, the upper triangularity, δupper, has been increased from 0.23 to 0.34 with no effect on Pth or the pedestal Te or Ti. In a separate edge ne scan, two configurations with different δupper/δlower values of 0.23/0.23 and 0.43/0.33 have been compared. A large difference in the transition threshold is observed at values of edge ne above 1.8 × 1019 m−3, with the higher δ plasmas characterized by Pth of up to 25% lower at the highest densities scanned. This is thought to be the result of lowered X-point and outer strike point heights with increased δ.

Ion cyclotron resonance frequencies (ICRF) mode conversion has been developed for localized on-axis and off-axis bulk electron heating on the JET tokamak. The fast magnetosonic waves launched from the low-field side ICRF antennas are mode-converted to short-wavelength waves on the high-field side of the He-3 ion cyclotron resonance layer in D and He-4 plasmas and subsequently damped on the bulk electrons. The resulting electron power deposition, measured using ICRF power modulation, is narrow with a typical full-width at half-maximum of approximate to30 cm (i.e. about 30% of the minor radius) and the total deposited power to electrons comprises at least up to 80% of the applied ICRF power. The ICRF mode conversion power deposition has been kept constant using He-3 bleed throughout the ICRF phase with a typical duration of 4-6 s, i.e. 15-40 energy confinement times. Using waves propagating in the counter-current direction minimizes competing ion damping in the presence of co-injected deuterium beam ions.

JET carbon screening experiments were performed using methane gas injection. L-Mode experiments scanned parameters influencing the JET scrape-off-layer (SOL) and/or intrinsic impurity level. Scaling relations are derived to describe methane injected into L-Mode plasmas from the JET horizontal mid-plane. L-Mode screening was 3–20 times better for plasmas connected to the divertor than for similar limited plasmas. The screening was worse for methane injection from the mid-plane and best for injection from the divertor. The screening was 1.5–2 times worse for H-Mode than L-Mode. Both ELM-averaged and inter-ELM H-Mode screening was documented. The screening results were used to understand the intrinsic impurity levels. Zeff reduced at higher densities partly due to better carbon screening at the higher density, and partly due to decreased carbon influxes. Diverted L-Mode intrinsic carbon levels arose from both main chamber and divertor sources, while H-mode carbon primarily originated from the divertor. DIVIMP and EDGE2D were used to model the observed screening. The modelling indicated that carbon removal to the divertor required lower temperatures for Coulomb collisions to couple the impurity ions to the SOL deuterium flows. The carbon removal occurred primarily in the outer SOL regions.

Melting is one of the major risks associated with tungsten (W) plasma-facing components (PFCs) in tokamaks like JET or ITER. These components are designed such that leading edges and hence excessiv ...

The effects of increased geometrical closure on the behaviour of the recycling and intrinsic impurities are investigated in JET Mark I, Mark IIA and Mark IIGB pumped diverters. Increasing the divertor closure leads to a significant improvement in exhaust for both deuterium and recycling impurities. However, the impurity enrichment in the exhaust gases remains unchanged due to a simultaneous increase in deuterium and impurity compression in the divertor. A comparison is made for helium, neon and argon under different plasma conditions. In addition, the operation of the Mark II and Mark IIGB diverters has shown that Z(eff) is reduced with the improved divertor closure in the L mode discharges, although no obvious changes in the Z(eff) values have been observed in the ELMy H modes. The divertor target surface temperature has a strong influence on intrinsic carbon production. The carbon source in the Mark II and Mark IIGB diverters is significantly higher than that in the Mark I divertor, which is attributed to enhanced chemical sputtering at the increased divertor tile temperature of the Mark II and Mark IIGB diverters (related to the divertor cooling system), as opposed to the increased closure. The consequences of this elevated yield for plasmas under different operation conditions are discussed, and further evidence, obtained from a specific wall/divertor temperature reduction experiment, is presented. The effect of the divertor screening on the chemically produced impurities is investigated using the EDGE2D/NIMBUS/DIVIMP codes for the different recycling regimes and comparisons are made with experimental observations from the Mark I, Mark IIA and Mark IIAP diverters taking into account the change in chemical sputtering yield due to the different tile temperatures of these diverters.

The effects of changing beam and plasma species on the edge transport barrier are investigated for ELM-free hot ion H mode discharges from the recent DT experiments on JET. The measured pressure at the top of the pedestal is higher for mixed deuterium and tritium and pure tritium plasmas over and above the level measured in pure deuterium plasmas at the same heating power. The pedestal pressure increases with beam tritium concentration for mixed deuterium-tritium beam injection into deuterium plasmas where the measured edge tritium concentration remains low. Alpha heating plays a significant role in the core of such plasmas, and the possible impact on the edge is discussed together with possible direct isotopic effects. Heuristic models for the transport barrier width are proposed, and used to explore a wider range of edge measurements including full power DD and DT pulses. This analysis supports the plasma current and mass dependence for a barrier width set by the orbit loss of either thermal or fast ions, though it does not unambiguously distinguish between them. The fast ion hypothesis could well account for some of the JET observations, though more theoretical work and direct experimental measurement would be required to confirm this. An ad hoc model for the power loss through the separatrix, Ploss ∝ nedge2 Zeff,edgeIp-1, is proposed based on neoclassical theory, a ballooning limit to the edge gradient and a barrier width set by the poloidal ion gyroradius. Such a model is compared with experimental data from JET. In particular, the model ascribes the systematic difference in loss power between the Mark I and Mark II divertors to the change in the measured Zeff. This change in Zeff is consistent with the observed change in impurity production, which is described in some detail, together with a possible explanation provided by the temperature dependence of chemical sputtering.

The scaling of the energy confinement in H mode plasmas with different hydrogenic isotopes (hydrogen, deuterium, DT and tritium) is investigated in JET. For ELM-free H modes the thermal energy confinement time τth is found to decrease weakly with the isotope mass (τth ~M-0.25±0.22), whilst in ELMy H modes the energy confinement time shows practically no mass dependence (τth ~M0.03±0.1). Detailed local transport analysis of the ELMy H mode plasmas reveals that the confinement in the edge region increases strongly with the isotope mass, whereas the confinement in the core region decreases with mass (τthcore ∝ M-0.16), in approximate agreement with theoretical models of the gyro-Bohm type (τgB ~M-0.2).

A number of experiments with heating of DT plasmas using ICRF waves have been carried out at JET. The results of these experiments have been analysed by comparing experimentally measured quantities with the results of numerical simulations. In particular, four scenarios have been examined: (a) heating of minority (~5-20%) deuterons at the fundamental ion cyclotron frequency, ω = ωcD; (b) second harmonic heating of tritium, ω = 2ωcT; (c) fundamental minority heating of 3He with a few per cent of 3He; (d) second harmonic heating of deuterium, ω = 2ωcD. An important aim of the analysis was to assess whether the present understanding of the ICRF physics is adequate for predicting the performance of ICRF in DT plasmas. In general, good agreement between experimental results and simulations was found which increases the confidence in predictions of the impact of ICRF heating in future reactors. However, when a relatively high deuterium concentration was used in the ω = ωcD scenario, discrepancies were observed. In order to increase confidence in the simulations, the sensitivity of the simulation results to various plasma parameters has been studied.

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.

Reactor relevant ICRH scenarios have been assessed during DT experiments on the JET tokamak using H mode divertor discharges with ITER-like shapes and safety factors. Deuterium minority heating in tritium plasmas was demonstrated for the first time. For 9% deuterium, an ICRH power of 6 MW gave 1.66 MW of fusion power from reactions between suprathermal deuterons and thermal tritons. The Q value of the steady state discharge reached 0.22 for the length of the RF flat-top (2.7 s), corresponding to three plasma energy replacement times. The Doppler broadened neutron spectrum showed a deuteron energy of 125 keV, which was optimum for fusion and close to the critical energy. Thus, strong bulk ion heating was obtained at the same time as high fusion efficiency. Deuterium fractions around 20% produced the strongest ion heating together with a strong reduction of the suprathermal deuteron tail. The ELMs had low amplitude and high frequency and each ELM transported less plasma energy content than the 1% required by ITER. The energy confinement time, on the ITERH97-P scale, was 0.90, which is sufficient for ignition in ITER. 3He minority heating, in approximately 50:50 D:T plasmas with up to 10% 3He, also demonstrated strong bulk ion heating. Central ion temperatures up to 13 keV were achieved, together with central electron temperatures up to 12 keV. The normalized H mode confinement time was 0.95. Second harmonic tritium heating produced energetic tritons above the critical energy. This scheme heats the electrons in JET, unlike in ITER where the lower power density will allow mainly ion heating. The inverted scenario of tritium minority ICRH in a deuterium plasma was demonstrated as a successful heating method producing both suprathermal neutrons and bulk ion heating. Theoretical calculations of the DT reactivity mostly give excellent agreement with the measured reaction rates.

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 ne-Te 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.

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.

Impurity production and plasma impurity concentrations are surveyed over a wide range of operating conditions but with emphasis on ELMy H mode operation using NB heating. The Mark I divertor campaigns with carbon and beryllium targets are compared with the Mark II campaign, which had carbon targets and a more closed divertor geometry. The beryllium target campaign has a significantly lower central impurity concentration than that with carbon targets, although the Zeff value is still dominated by carbon. The divertor geometry appears to have little effect on the central impurity level, but it is found that Zeff and the carbon and nickel concentrations are correlated with the ELM frequency, with maxima at a frequency of around 5-15 Hz. This occurs whether the ELM frequency is varied by changing the plasma triangularity or the input power, or by increasing the fuelling rate. The absolute impurity influxes from the inner and outer targets and from the inner wall have been estimated from the C II and C III line intensities. These fluxes are compared with the central carbon concentration, and an attempt is made to correlate the influxes and the concentrations in the confined plasma, both in their time evolution and in their absolute values.

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.

Heating during the early part of the current rise phase gives a low or negative magnetic shear (= (dq/dr)) in the centre of JET plasmas. Under these conditions the confinement improves with high additional heating power heating during the current ramp-up phase of the discharge. The reduction in the transport manifests itself as a peaking of the profiles with a large gradient region near = 0.55. The best discharges have no transport barrier at the edge of the plasma (L-mode). This allows central power deposition by the neutral beams in JET. A control of the plasma pressure, using feedback of the additional heating power in real-time, minimizes the impact of magnetohydrodynamic instabilities. As a result, these discharges achieve the highest D-D neutron rates in JET; , with , and .

Experiments using high-power heating during the current ramp-up phase of the discharge have obtained the highest D-D neutron rates in JET; neutrons , with , keV and keV. The best discharges ( MA and tesla) have peaked pressure profiles with a transport barrier located at = 0.55. The pressure peaking is limited by MHD modes and requires active input power control to achieve the best performance. Deuterium neutral beam injection into a tritium-rich target plasma has established internal transport barriers at power levels close to the lowest threshold for pure deuterium plasmas.

An analysis of the complete JET D-D threshold database, from 1990 to 1997 inclusive is presented. A JET scaling is derived for the threshold power through the separatrix, estimated by considering also radiation losses from the bulk plasma. Such scalings are applied to obtain a first qualitative estimate of the amount of power above threshold that is needed to obtain steady-state high confinement (H89 > 1.8). The L-H and H-L transitions are also analysed in terms of local edge parameters. The first results on isotope scaling of the H-mode power threshold in D-T and T-T plasmas are also presented, confirming the scaling.

Evidence for volume recombination in the JET divertor has been obtained from the population ratios of the excited states of neutral hydrogen. The ratio of the Balmer series lines Dγ/Dα has been used to identify recombination at both the inner and the outer divertor target. At the outer target the onset of recombination is well correlated with the plasma detachment as observed with Langmuir probes in the divertor target. While Te obtained from the target probe characteristics falls to ~2 eV, Te from the population distribution of the excited states is 0.8±0.1 eV. The peak plasma density viewing across the outer divertor leg just above the target exceeds 2.5 × 1020 m-3. When the plasma becomes partially detached from the outer target, the radial distribution of recombination shows a maximum close to the outer divertor strike point.

The experimental characteristics of divertor detachment in the JET tokamak with the Mark?I pumped divertor are presented for ohmic, L?mode and ELMy H?mode experiments with the main emphasis on discharges with deuterium fuelling only. The range over which divertor detachment is observed for the various regimes, as well as the influence of divertor configuration, direction of the toroidal field, divertor target material and active pumping on detachment, will be described. The observed detachment characteristics, such as the existence of a considerable electron pressure drop along the field lines in the scrape-off layer (SOL), and the compatibility of the decrease in plasma flux to the divertor plate with the observed increase of neutral pressure and D? emission from the divertor region, will be examined in the light of existing results from analytical and numerical models for plasma detachment. Finally, a method to evaluate the degree of detachment and the window of detachment is proposed, and all the observations of the JET Mark?I divertor experiments are summarized in the light of this new quantitative definition of divertor detachment.

All major JET systems have been fully commissioned for D-T and the DTE1 series of experiments has started with the D-T fuel mixture and operating conditions foreseen for ITER. In the area of ITER physics, significant results have been produced in both D-D and D-T. In D-D, the LH threshold power database has been extended, the bounds on edge-electron temperature and density in ELMy H-modes have been defined and the advantages of Types I and III ELMy discharges have been compared. In D-T plasmas, the isotope effect on H-mode threshold power and transport has been determined so that a more accurate assessment can be made of the ignition margin and heating requirements for ITER. Trace tritium experiments have provided first particle transport measurements and an assessment of the ITER reference ion-cyclotron resonance-frequency heating scenarios has been started, In the area of fusion performance, record D-D neutron yields have been obtained by controlling the plasma and current profiles in hot ion ELM-free H-modes and optimized shear modes. In D-T, internal transport barriers have been readily established in optimized shear discharges and Alfven eigenmodes have been observed.

In nearly all hot ion H modes in JET, a magnetohydrodynamic (MHD) event is clearly observed just before the time the stored plasma energy saturates and the neutron yield starts to decline. The results of a systematic analysis of MHD observations for a large number of discharges is reported. The relationship between MHD phenomena and the onset of confinement limitation is discussed, as are aspects of the three main types of performance limiting MHD: (a) low-n modes in the outer regions of the plasma, (b) sawteeth and (c) giant edge localized modes (ELMs). Model simulations indicate how the transport is affected and allow an assessment of how much the neutron yield would be improved if the MHD activity were absent

Dedicated experiments on JET clarified the effect of the toroidal field reversal on divertor asymmetries. In high recycling L-mode discharges the field reversal had a significant impact on density, temperature, radiated power, particle flux and emission distribution between the strike zones. The in/out ratio of and emission was reduced following the change from `normal' to `reversed' field configuration, while the in/out asymmetry was increased. This behaviour is consistent with previous observations and suggests the radial drift as a possible candidate to explain the observed effects. In contrast with the L-mode discharges, a significantly weaker effect of the field reversal on the asymmetries was observed in high density ohmic discharges, which is consistent with the expected reduced drift contribution to plasma transport in low-temperature plasmas. In the L-mode series of discharges with reversed , the previously reported q-dependence of power sharing between the strike zones, with more symmetric power deposition at low was validated. Asymmetry in the power deposition to the target was strongly in favour of the outer strike zone for the series of L-mode discharges analysed, and could not be significantly altered by the inclusion of local strike point radiation. The shifts in the power asymmetry between the targets due to the reversal agrees with previous observations on JET and elsewhere that power distribution between the sides is more symmetric in the `reversed' toroidal field.

An extensive experimental investigation of reactor-relevant features of H-mode plasmas has been carried out using the JET-pumped divertor. Steady-state H-modes have been established with high confinement and with durations of up to 20 s. At the highest densities, these plasmas return to the L-mode when the radiated power fraction is %. Detached H-modes have been produced by gas-puffing deuterium plus nitrogen, and their properties studied at total input powers of up to 32 MW. Helium transport and exhaust have also been investigated, and a systematic comparison of H-modes using carbon fibre composite and beryllium divertor targets has been carried out.

The results of a series of ITER simulation experiments on JET are described. A series of H-mode threshold experiments are shown to reproduce one of the standard power threshold scaling expressions that is being used to predict the power threshold in ITER. Then, from a series of experiments in which the Larmor radius scaling of ELMy H-modes is examined, it is concluded that the scaling of the confinement is gyro-Bohm-like provided the power levels are well above the threshold. Finally, we show that at high there is a dramatic reduction in the confinement at .

L to H transition studies at JET have revealed an n = 0, m = 1 magnetic oscillation starting immediately at the L to H transition (called M-mode for brevity). While the magnetic oscillation is present a weak ELM-less H-mode regime is obtained, with a clear increase of density and a weak electron temperature pedestal. It is an intermediate state between L and H-mode. In ICRH heated plasmas or low density NBI plasmas the magnetic mode and the pedestal can remain steady (with small oscillations) for the duration of the heating phase, of order 10 s or more. The axisymmetric magnetic oscillation has period ~0.5–2 ms, and poloidal mode number m = 1: it looks like a pedestal localised up/down oscillation, although it is clearly a natural oscillation of the plasma, not driven by the position control system. Electron cyclotron emission, interferometry, reflectometry and fast Li beam measurements locate the mode in the pedestal region. D α , fast infrared camera and Langmuir probe measurements show that the mode modulates heat and particle fluxes to the target. The mode frequency appears to scale with the poloidal Alfvén velocity, and not with sound speed (i.e. it is not a geodesic acoustic mode). A heuristic model is proposed for the frequency scaling of the mode. We discuss the relationship between the M-mode and other related observations near the L–H transition.

An account is given of the high performance plasmas established by development of the H-mode regime in JET in the experimental campaigns up to 1992. High performance in this case is measured in terms of the confinement enhancement achieved over the L-mode scaling as measured using the plasma diamagnetism. Three JET H-mode regimes have achieved enhancement factors (H G DIA ) over Goldston L-mode scaling of 2.5 < H G DIA < 4.0. These are the Pellet Enhanced Performance (PEP) H-MODE, the high bootstrap fraction (high β POL ) H-mode and the Hot Ion (HI) H-mode. The phenomenology of these three regimes is reviewed and contrasts and common threads are elucidated

High poloidal beta plasmas with a correspondingly high bootstrap fraction of 70+or-15% have been produced using ion cyclotron resonance heating, in an ELM free H mode confinement regime. This regime is terminated by a large edge instability, causing a reversion to L mode confinement. Possible causes for this instability are discussed and the overall implications of the results for steady state bootstrap dominated operation are examined

Preliminary experiments were carried out in the Joint European Torus (JET) using 120 keV helium (He) neutral beam injection (NBI). Injected power levels up to 5 MW with 3He and 7 MW with 4He, lasting up to 3 s were reached. The 3 s helium NBI produced efficient ion heating and similar global and local energy confinements to those obtained with deuterium (D) NBI, in both limiter and X-point plasma geometries, in L-mode and H-mode plasma regimes. The elimination of beam-plasma and beam-beam fusion reactions by replacing D NBI with He NBI extended the range for measuring ion temperatures with the JET neutron profile monitor and neutron spectrometers.

The JET ICRF antennas, equipped with beryllium (Be) Faraday screens (FS), can be operated in such a way that the RF specific effects on the plasma boundary, by the impurity influx originating at the screens, are reduced to a negligible level. In dipole phasing, k|| = 7 m-1, the influx is for all purposes negligible. In monopole phasing (k|| = 0 m-1) the beryllium influx does not exceed BeFS = 1 ? 1019 atoms.MW-1.s-1 and the corresponding delta Zeff/PRF is < 0.005 MW-1. The observed dependences of BeFS (in monopole phasing) on plasma density antenna voltage, antenna phasing, and the angle between FS elements and the magnetic field in the boundary, B(a) = B?(a) + BT(a), confirm that the release mechanism is sputtering by ions accelerated in the RF enhanced Bohm-Debye sheaths forming at the front face of the FS. When the angle between FS and B(a) is 22?, the fraction of the RF power radiated by the antenna, dissipated at the screen, can reach 40%. At high antenna voltage, arcing across the FS can occur. With dipole phasing the heating efficiency is not degraded, even with the large angle, and all the power coupled by the antenna is absorbed at the resonance position near the plasma centre. The open screen design did not introduce any disadvantages. The experience from JET operation at powers of up to 22 MW shows that, if the necessary conditions are met, i.e. if RF rectification is minimized, antennas are phased as dipoles and material with low sputtering coefficients at energies of 0.5-1 keV is used, then the influx from the FS is, for all practical purposes, eliminated

Spectral observations of ions of beryllium in the JET plasma are described and are interpreted in terms of the interplay of atomic reactions and local plasma conditions. The methods of generalized-collisional-radiative theory are extended and used to provide a quantitative model. Emphasis is placed on the role of metastable ion states. The diagnostic scope of beryllium observations is explored for the influx ions Be0+ and Be1+ in the visible spectral range, for Be2+ in the XUV, and for Be3+ by charge exchange spectroscopy. Useful results for general diagnostic interpretation are presented in tables and graphs. The calculation procedures and data form part of the 'Atomic Data and Analysis Structure' in use at JET.

During ICRF heating in JET highly-localized impurity influxes from the Faraday screens (FS) of RF antennas are observed. The influxes result from the acceleration of deuterium and impurity ions in the RF-enhanced Bohm-Debye sheaths. The RF-enhanced sheaths form when the equilibrium magnetic field line, linking a large RF magnetic flux, connects two surfaces and closes a loop where the RF voltage is induced. The FS influxes depend on two intimately-linked parameters, the antenna voltage and the density at the FS, as well as on the magnetic field angle, antenna phasing, FS geometry and material.

The baseline type-I ELMy H-mode scenario has been re-established in JET with the new tungsten MKII-HD divertor and beryllium on the main wall (hereafter called the ITER-like wall, JET-ILW).The first JET-ILW results show that the confinement is degraded by 20–30% in the baseline scenarios compared to the previous carbon wall JET (JET-C) plasmas. The degradation is mainly driven by the reduction in the pedestal temperature. Stored energies and pedestal temperature comparable to the JET-C have been obtained to date in JET-ILW baseline plasmas only in the high triangularity shape using N2 seeding.This work compares the energy losses during ELMs and the corresponding time scales of the temperature and density collapse in JET-ILW baseline plasmas with and without N2 seeding with similar JET-C baseline plasmas. ELMs in the JET-ILW differ from those with the carbon wall both in terms of time scales and energy losses. The ELM time scale, defined as the time to reach the minimum pedestal temperature soon after the ELM collapse, is ~2 ms in the JET-ILW and lower than 1 ms in the JET-C. The energy losses are in the range ΔWELM/Wped ≈ 7–12% in the JET-ILW and ΔWELM/Wped ≈ 10–20% in JET-C, and fit relatively well with earlier multi-machine empirical scalings of ΔWELM/Wped with collisionality. The time scale of the ELM collapse seems to be related to the pedestal collisionality. Most of the non-seeded JET-ILW ELMs are followed by a further energy drop characterized by a slower time scale ~8–10 ms (hereafter called slow transport events), that can lead to losses in the range ΔWslow/Wped ≈ 15–22%, slightly larger than the losses in JET-C. The N2 seeding in JET-ILW significantly affects the ELMs. The JET-ILW plasmas with N2 seeding are characterized by ELM energy losses and time scales similar to the JET-C and by the absence of the slow transport events.

When the ITER-like wall was installed in JET, one of the 218 Be inner wall guard limiter tiles had been enriched with 10Be as a bulk isotopic marker. During the shutdown in 2012-2013, a set of tiles were sampled nondestructively to collect material for accelerator mass spectroscopy measurements of 10Be concentration. The letter shows how the marker experiment was set up, presents first results and compares them to preliminary predictions of marker redistribution, made with the ASCOT numerical code. Finally an outline is shown of what experimental data are likely to become available later and the possibilities for comparison with modelling using the WallDYN, ERO and ASCOT codes are discussed.

Estimations of the ITER first wall (FW) lifetime, previously made using the three-dimensional Monte-Carlo ERO code (Borodin et al 2011 Phys. Scr. T145 014008), depend strongly on the assumptions of the physical sputtering yield for beryllium (Be). It is of importance to validate the respective model and data at existing devices including the JET ITER-like wall (ILW) as most ITER-relevant experiments. Applying the same sputtering input data in ERO as those used before in the ITER-predictions, the ERO simulations for the Be light intensity (using up to date atomic data from ADAS and measured plasma conditions) reveal a factor of 2 overestimation in the assumed yield even if the low estimate assuming 50% D surface content is used. This result indicates the preference of this assumption for plasma-wetted areas. It points to a possible necessity to correct (reduce) the respective estimates for the Be sputtering yield and, accordingly, re-visit the ITER FW lifetime predictions.

Recent experiments at JET combining reciprocating probe measurements upstream and infrared thermography at the plasma-facing components (PFC) on plasmas in limiter configurations show that the common approach to predicting the power load on the limiter underestimates the heat flux at the contact point by a factor 1.5-3. The current model and scaling laws used for predicting the power load onto the first wall during limiter current ramp-up/down in ITER are uncertain and a better understanding of the heat transport to the PFCs is required. The heat loads on PFCs are usually predicted by projecting the parallel heat flux associated with scrape-off layer (SOL) properties at the outer mid-plane (upstream) along the magnetic field lines to the limiter surface and deducing the surface heat flux through a cosine law, thus ignoring any local effect of the PFC on transport within the SOL. The underestimate of the heat flux is systematic in inner wall limiter configurations, independent of the plasma parameters, whereas in outer limiter configuration this is not observed, probably because of the much shorter SOL power decay length. Models that can explain this enhanced heat flux around the contact point are proposed and discussed although no definitive conclusion can be drawn. © 2013 IAEA, Vienna.

We present in this paper modelling of the JET scrape-off layer (SOL) including direct SOL ionization by launched lower hybrid (LH) waves. For that purpose, the two-dimensional fluid code EDGE2D-NIMBUS was modified in order to account for possible enhanced ionization in the SOL, due to the LH power absorbed by the electrons in a layer extending in the radial direction from the launcher. By taking the direct LH SOL ionization into account, the observed modification of the density in front of the LH antenna during a LH power and gas injection, can be explained. The JET grill private SOL limiters, acting as sinks for the particles, are also included in the model. We compare the efficiency of gas puffing from the top and the outer mid-plane for SOL density enhancement and improvement in LH coupling. The observed reduction of the temporal variation of the LH wave reflection coefficient and of measured saturated currents in the SOL during ELMs, is explained. The density depletion, by ponderomotive forces in front of the grill, is estimated.

The code COREDIV, self-consistent with respect to both the interaction of plasma core-edge and main plasma impurities, is used to simulate nitrogen-seeded discharges in JET. The model is fully described and numerical results are compared with experimental data pertaining to two series of discharges differing in input power, confinement, the level of power radiated and the puffing rate of the main gas. Impurity sources, their transport and densities are considered as functions of the edge and core parameters and consistency of their radiated power is discussed. Special emphasis is given to analysis of the fluxes of carbon and recycled deuterium.

Strike-point sweeping and real-time-controlled (RTC) impurity seeding are both expected to be needed on JET following its upgrade to an all-metal wall with enhanced neutral-beam heating, thereby anticipating exhaust-control requirements plus the materials planned for ITER. Preliminary trials in the previous carbon device have combined these techniques in high-triangularity type I H-mode plasmas, using a VUV spectroscopic signal for feedback control of nitrogen injection. Compared with earlier unswept feedforward counterparts, similar strong mitigation of divertor heat load between ELMs was achieved in swept RTC cases for less than half the integrated nitrogen input and correspondingly less adverse effect upon other properties. Both sweeping and RT control contributed to this improvement. Time-average normalized energy confinement H98yt ~ 1, Greenwald density fraction fGwdt ~ 0.9 and particularly purity denoted by effective ionic charge Zefft ≈ 1.7, all remained closer to good reference levels. Transient effluxes in ELMs were also less affected, however, and would require separate active control.

Modification of the sawtooth period through ion cyclotron resonance frequency (ICRF) heating and current drive has been demonstrated in a number of experiments. The effect has been seen to depend critically on the location of the ICRF absorption region with respect to the q = 1 flux surface. Consequently, for ICRF to be a viable tool for sawtooth control, one must be able to control the ICRF absorption location in real time so as to follow variations in the location of the q = 1 surface. To achieve this, the JET ICRF system has been modified to allow the JET real time central controller to control the frequency of the ICRF generators. An algorithm for real time determination of the sawtooth period has been developed and a closed loop controller, which modifies the frequency of the ICRF generators to bring the measured sawtooth period to the desired reference value, has been implemented. This paper shows the first experimental demonstration of closed loop sawtooth period control by real time variation of the ICRF wave frequency.

In anticipation of revised operational limits accompanying upgrade of the JET tokamak to an all-metal, ITER-like wall, systematic studies have been conducted of moderating divertor heat load in type I ELMy H-mode by enhanced deuterium fuelling and seeding with the extrinsic impurities neon or nitrogen. These were found to differ mainly in the roughly ten times higher level of nitrogen than neon presently required to approach outboard detachment, and in their effects upon ELMs. With either species up to input power of ?17?MW, strongly reduced target loads between ELMs, compatible with the new wall, were sustained for longer than 10 energy confinement times while losing

The ELM filamentary structure in the JET scrape-off layer (SOL) has been investigated using a multi-pin reciprocating Langmuir probe system inserted at the top, low field side of the poloidal cross-section. These measurements clearly demonstrate that both Type I and III edge localized modes (ELMs) are composed of a number of filaments. The filament propagation in the SOL is investigated and the radial velocity estimated using different methods. The large radial velocities observed during Type I ELMs (up to 6 km s−1) imply that filaments have no time to drain significant energy along the field lines, reaching the wall with a reasonable fraction of the pedestal density and temperature. In spite of the time between Type I ELMs being much longer than the ELM duration, most of the particle flux to the main chamber plasma-facing components occurs in the latter period. Furthermore, filaments in Type I ELMs are significantly larger and faster than those observed in Type III ELMs. During Type I ELMs a strong parallel flow (M ~ 0.4) is measured towards the inner divertor, which increases during filaments. These observations are in agreement with the ballooning nature of the ELM losses.

The first experimental evidence showing the connection between blob/hole formation and zonal-flow generation was obtained in the edge plasma of the JET tokamak. Holes as well as blobs are observed to be born in the edge shear layer, where zonal-flows shear off meso-scale coherent structures, leading to disconnection of positive and negative pressure perturbations. The newly formed blobs transport azimuthal momentum up the gradient of the azimuthal flow and drive the zonal-flow shear while moving outwards. During this process energy is transferred from the meso-scale coherent structures to the zonal flows via the turbulent Reynolds stress, resulting in nonlinear saturation of edge turbulence and suppression of meso-scale fluctuations. These findings carry significant implications for the mechanism of structure formation in magnetically confined plasma turbulence.

The determination of a good relative and absolute sensitivity calibration for wideband VUV spectrometers is challenging. On JET, the possible T and Be contamination of the VUV spectrometer precludes its removal to a synchrotron source and, consequently, a range of alternative in situ techniques have been investigated in depth. This has resulted in a reliable calibration for the complete spectral range, the relative calibration at short wavelengths being particularly accurate. At these wavelengths, a novel approach is used, in which the calibration is extended using a number of Na- and Li-like metal doublets. At longer wavelengths, the Li-like doublets of Ar and Ne have been used in conjunction with CII, CIII and CIV line intensity ratios. Unexplained discrepancies between the measured and modelled C results have meant that the exceptional short wavelength accuracy has not be repeated at these longer wavelengths. The absolute sensitivity has been determined from branching ratios to an absolutely calibrated visible spectrometer. The long term stability of the calibration is discussed.

Radio frequency heating of majority ions is of prime importance for understanding the basic role of auxiliary heating in the activated D-T phase of ITER. Majority deuterium ion cyclotron resonance heating (ICRH) experiments at the fundamental cyclotron frequency were performed in JET. In spite of the poor antenna coupling at 25 MHz, this heating scheme proved promising when adopted in combination with D neutral beam injection (NBI). The effect of fundamental ICRH of a D population was clearly demonstrated in these experiments: by adding similar to 25% of heating power the fusion power was increased up to 30-50%, depending on the type of NBI adopted. At this power level, the ion and electron temperatures increased from T-i similar to 4.0 keV and T-e similar to 4.5 keV (NBI-only phase) to T-i similar to 5.5 keV and T-e similar to 5.2 keV (ICRH + NBI phase), respectively. The increase in the neutron yield was stronger when 80 keV rather than 130 keV deuterons were injected in the plasma. It is shown that the neutron rate, the diamagnetic energy and the electron as well as the ion temperature scale roughly linearly with the applied RF power. A synergistic effect of the combined use of ICRF and NBI heating was observed: (i) the number of neutron counts measured by the neutron camera during the combined ICRF + NBI phases of the discharges exceeded the sum of the individual counts of the NBI-only and ICRF-only phases; (ii) a substantial increase in the number of slowing-down beam ions was detected by the time of flight neutron spectrometer when ICRF power was switched on; (iii) a small D subpopulation with energies slightly above the NBI launch energy was detected by the neutral particle analyzer and gamma-ray spectroscopy.

The influence of the physical isolation of inner and outer divertor volumes by a septum plate of the Mk-II gas box divertor, thus increasing divertor closure and neutral compression, on the plasma and divertor performance has been studied at the Joint European Torus (JET). The septum plate was installed in 1999, together with the original Mk-II gas box divertor, and was then replaced by a simple protection plate in 2001. This removal reduced the closure of the divertor by opening a line of sight path for neutrals to travel between the inner to the outer divertor volumes. Comparison of identical discharges with and without the septum thus provides direct evidence of the effect of divertor closure on plasma behaviour. With this aim, following septum removal, several dedicated L-mode and H-mode discharges have been performed, in each case repeating earlier discharges when the septum was still in place. In each case, the fuelling location was varied between the inner/outer divertor and the main chamber, and differences in detachment in the inner and outer divertors were studied. Under L-mode conditions, differences in detachment dynamics were indeed observed between closed (with septum) and open (without septum) divertor configurations, although the differences were only significant in the medium density range. In contrast, the ultimate density limit was not affected, being determined in each case by the formation of a wall multifacedted asymmetric radiation from the edge (MARFE), rather than an X-point MARFE. Under H-mode conditions, the differences were more subtle. Although the ion fluxes to the targets were unaffected, the target electron temperatures were found to be lower in the closed divertor configuration. In this case, the fuelling efficiency was the largest when the gas injected from the inner divertor, with implications on global energy confinement and ELM frequency. Otherwise, no difference in the confinement of the discharges with and without septum was observed.