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Divertor monitoring pulses (DiMPle) have been run in JET from the C35 campaign onwards. They provide an opportunity to study the impurity contamination of the plasma when it is limited by different surfaces within the machine, as well as the longer term behaviour of the impurities. In these discharges the plasma is first limited on the outer wall, then on the inner wall and, subsequently, in the X-point configuration the outer strike point is positioned on the horizontal tile 5 of the machine followed by tile 6 and then the vertical tile 7. The present study details the impurity behaviour in the DiMPle pulses from JET-ILW campaigns C35 to C38, which ran from 2015 to 2019. The impurities can largely be divided into two groups. The first, including most gases, are present immediately after their use in the machine; the second group includes those elements that are retained on plasma facing surfaces within the vessel. Most of these are metals, for which a systematic behaviour is found. Influxes due to metallic dust particles behave more like the elements of the first group. The origin of the impurities where this is known is given as well as details of the systematic behaviour, including differences due to the line-of-sight of the observing spectrometer. A clear difference is seen when the discharge fuel is H and this has implications for tritium and deuterium-tritium operations.

A discrepancy in the divertor radiated powers between EDGE2D-EIRENE simulations and JET-ILW experiments employing a set of NBI-heated L-mode discharges with step-wise density variation is investigated. The analysis shows the importance of the contributions from the deuterium fuel to the divertor radiated power, making these discharges ideal for this study. Molecular radiation losses from D2 molecules have been included for the first time and the present simulations allow the atomic and molecular electron power loss terms in EIRENE to be independently scaled in order to understand their importance in determining the simulated electron temperatures and distribution of radiated power. PSI-22 keywords: Divertor modelling, Divertor diagnostic, EDGE2D, JET, ILW

In preparation for the upcoming JET D-T campaign, great effort has been devoted during the 2015-2016 JET campaigns with the ITER-like wall (ILW) to the extension of the high performance H-mode phase in baseline and hybrid scenarios. Hybrid discharges were the only ones that have been stopped by the real-time vessel protection system due hot-spot formation on the outboard poloidal limiter. Generation of hot-spots was linked to the application of high neutral beams injection and ion cyclotron resonance heating (ICRH) power. In tokamaks with high-Z plasma components, the use of ICRH heating is also accompanied by an increased metallic impurity content. Simultaneous control of hot-spot temperature and the core impurity content was crucial due to the fact that the same plasma-wall interaction mechanism is responsible for both phenomena. Impurity data collected by SXR, EUV and VUV diagnostics were able to provide for the first time comprehensive information concerning tungsten and mid-Z impurities such as nickel, iron, and cooper. To determine absolute mid-Z impurity concentrations a new relative calibration technique, compatible with JET-ILW, has been developed based on cross-calibration with a calibrated spectrometer via the quasicontinuum of W in the 200-400 Å wavelength range. In hybrid discharges, it was found that local D2 gas injection, plasma current, separatrix density, and fast ion losses appeared to impact hot-spot temperature and core impurity levels. Analysis showed a reduced maximum hot-spot temperature and impurity concentration at higher gas rate. Changes in the plasma current had a strong impact on the plasma-wall interaction, both via modifications in the edge density and in the fast ion losses. At constant gas injection rate, both the hot-spot temperature and the core impurity content decreased with the separatrix density. The main mechanism responsible for the formation of the hot-spots was found to be linked to the fast ion losses, but RF sheath effects may also be playing a role in the high limiter temperatures observed in these experiments. © Institute of Plasma Physics and Laser Microfusion.

Collision strengths (Ω) are calculated for all 6328 transitions among the lowest 113 levels belonging to the 2 s 2 2 p 5 , 2 s 2 p 6 , 2 s 2 2 p 4 3 l , 2 s 2 p 5 3 l , and 2 p 6 3 l configurations of fluorine-like krypton, Kr XXVIII, using the Dirac Atomic R-matrix Code. All partial waves with angular momentum J ⩽ 40 are included, sufficient for the convergence of Ω for forbidden transitions. For allowed transitions a top-up is employed to obtain converged values of Ω up to an energy of 400 Ryd. Resonances in the thresholds region are resolved on a narrow energy mesh, and results for effective collision strengths (ϒ) are obtained after averaging the values of Ω over a Maxwellian distribution of electron velocities. Values of ϒ are reported over a wide temperature range below 10 7.1 K , and the accuracy of the results is assessed. In addition, effective collision strengths are listed for the temperature range 7.0 ⩽ log T e ( K ) ⩽ 9.0 , obtained from non-resonant collision strengths generated with the FAC code.

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.

The design value for ITER is based on operation at n/n(GW) = 0.85, beta(n) = 1.8 and H98(y, 2) = 1. These values have been routinely achieved in JET in argon seeded ELMy H-mode discharges in different divertor configurations and with different triangularities. Two main scenarios are emerging from the experiments.First, low triangularity (delta(u) = 0.19) in septum configuration. In this case large D-2 fuelling rates lead to confinement degradation towards L-mode. The seeding of Ar during the D-2 fuelling phase gives rise to a density close to the Greenwald value. After the switch-off of the D-2 gas fuelling ('afterpuff ' phase), the confinement recovers to H-mode quality whereas the density stays near the value reached at the end of the main fuelling phase and Z(eff) stays close to or below 2. Acting on the refuelling of Ar and D-2 in the 'afterpuff 'phase allows us to improve the stationarity of the high performance phase while maintaining up to the end of the heating phase the good confinement, density and radiation level.Second, high triangularity (delta(u) = 0.45) in vertical target configuration. In this case large fuelling rates do not lead to strong confinement degradation and the D-2 fuelling is applied continuously throughout the discharge. A radiated power fraction of up to 70%, H98(y, 2) = 0.9 at beta(n) = 2.1 and n = 1.15n(GW)-together with the formation of a radiating mantle and moderate Z(eff)-are achieved in this scenario. Furthermore, there are indications of significantly reduced heat load on the divertor target plates.

This paper addresses the issues of impurity behaviour during the argon seeding experiments in JET, in which argon and D2 have been simultaneously puffed in ELMy H-mode discharges to reach high density regimes, maintaining good confinement properties throughout the plasma discharge. The analysis is based mainly on a 1-D impurity diffusion model, which evaluates the impurity transport coefficients in three experimental scenarios. The available experimental data include, as a function of the time, the brightnesses of the soft x-rays and of several impurity lines, the effective charge Zeff, the profiles of the radiated power and the fully stripped Ar and C ion densities from charge-exchange recombination. From the simulation of these data it is possible to characterize the considered scenarios in terms of high-Z ion increase in the plasma centre and high radiation from the edge. In particular, it has been found that the configuration at high triangularity with continuous D2 puffing, characterized by an outward impurity pinch velocity, features flat or slightly hollow impurity ion profiles and high radiation from a narrow region at the edge, possibly associated to the presence of a significant neutral density. In this case an edge radiative mantle is established without high radiation from the plasma centre.

Magnetically confined plasmas, such as those produced in the tokamak JET, contain measurable amounts of impurity ions produced during plasma-wall interactions (PWI) from the plasma-facing components and recessed wall areas. The impurities, including high- and mid-Z elements such as tungsten (W) from first wall tiles and nickel (Ni) from Inconel structure material, need to be controlled within tolerable limits, to ensure they do not significantly affect the performance of the plasma. This contribution focuses on documenting W and Ni impurity behavior during Ion Cyclotron Resonance Heating (ICRH) operation with the new ITER-Like Wall (ILW). Ni- and W-concentration were derived from VUV spectroscopy and the impact of applied power level, relative phasing of the antenna straps, plasma separatrix - antenna strap distance, IC resonance position, edge density and different plasma configuration, on the impurity release during ICRH are presented. For the same ICRH power the Ni and W concentration was lower with dipole phasing than in the case of −π/2 phasing. The Ni concentration was found to increase with ICRH power and for the same NBI power level, ICRH-heated plasmas were characterized by two times higher Ni impurity content. Both W and Ni concentrations increased strongly with decreasing edge density which is equivalent to higher edge electron temperatures and more energetic ions responsible for the sputtering. In either case higher levels were found in ICRH than in NBI heated discharges. When the central plasma temperature was similar, ICRH on-axis heating resulted in higher core Ni impurity concentration in comparison to off-axis ICRH in L-mode. It was also found that the main core radiation during ICRH came from W.

Methane screening experiments were performed in JET L-Mode plasmas. Potentially, methane puffing quantifies the SOL processes governing the intrinsic carbon, its penetration to the plasma core, and its removal to the divertor. We measured the perturbations to the core and the SOL plasmas as a function of CD4 and D2 injection rates. The deduced screening was independent of the methane injection rate indicating that the methane itself did not significantly perturb the deduced screening. Identification of appropriate reference plasmas is important. For JET L-Modes, the reference plasma should have enough deuterium puffing to achieve the same density as that which occurred with the methane. The screening results are susceptible to systematic errors in the visible and charge exchange determination of the core carbon density. For JET, these two measurements yield screening values that differ by a `factor-of-three'.

ELMy H-mode pulses have been obtained with different hydrogenic isotopes (H and D) but having the same profiles of the dimensionless parameters ρ*,β,ν* and q, to test whether the confinement scale invariance principle is valid in a tokamak. The fact that the confinement times, the ELM and sawtooth frequencies in the two pulses all scale as expected suggests that the invariance principle is satisfied through the plasma radial extent, in spite of the differing physical processes taking place in the plasma centre, core and edge regions. An application of this `isotope windtunnel technique' to predicting the D-T performance of the next step devices is discussed.

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.

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.

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.

An empirical procedure, “LINT,” for relating emission line intensities of intrinsic impurity ions to their elemental contributions to the total, bolometric, radiation loss and the volume-averaged effective ion charge, Zeff, has been developed and applied to limiter plasmas in the JET tokamak. In this article we discuss extensions to the data base to include x-ray lines and continua intensities, applicable to a wider range of tokamak plasma configurations such as X-point plasmas and quasi-steady-state, edge-cooled ELMy H modes. Examples are shown of the technique applied to reference discharges during which the plasma configuration is changed continuously. The total data set, comprising line and continua irradiances, tomographic bolometry, and x-ray emission and Zeff imposes constraints on the diffusion parameters used in models of impurity ion transport.

Ion cyclotron resonance heating (ICRH) experiments have been carried out in JET D-T plasmas using scenarios applicable to reactors. Deuterium minority heating in tritium plasmas is used for the first time and produces 1.66 MW of D-T fusion power for an ICRH power of 6 MW. The Q value is 0.22, which is a record for steady state discharges. Fundamental He-3 minority ICRH, in both 50:50 D-T and tritium dominated plasmas, generates strong bulk ion heating and ion temperatures up to 13 keV. Second harmonic tritium ICRH is seen to heat mainly the electrons as expected for JET conditions. All three schemes produce H-mode plasmas.

The analysis of the contamination of the plasma by a puffed recycling impurity, such as neon, is intrinsically time-dependent and requires simultaneous time-dependent modelling of both the core region and the SOL and divertor region. To this end, a new code, combining core and edge codes, has been developed: its main features will be described and its application to the analysis of an L-mode discharge in which neon was puffed will be shown.

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

Sophisticated X-ray spectroscopic instruments have been developed and used for confinement studies in tokamaks at JET and Culham Laboratory. The Bragg rotor survey spectrometer, when programmed to operate in monochromatic or limited spectral range (LSR) modes, has been used to derive transport coefficients from the transient spectral signature of injected, non-recycling, test ions. The coefficients appropriate to specific confinement regimes have been investigated with respect to ion charge state and tokamak machine parameters. The LSR mode allows discrimination against background X-:ray light, a particularly serious problem at high electron density, ne ≈ 1020 m−3, in H mode where there is evidence, in COMPASS-D, for modest accumulation of impurities. Impurity ion accumulation is observed most clearly in passive spectroscopic studies of the 1s2 − 1s2p line emission from intrinsic impurity ions such as CI(XVI) using a high resolution Johann configuration crystal spectrometer. Line shapes and line ratios within the limited spectral range 4.45-4.50 A are interpreted in terms of ion and electron temperature and ion fluid velocity. In particular, the ‘q’/‘w’ line intensity ratios are interpreted in terms of the relative abundance of the CI(XV)/CI(XVI) ions which are modelled in terms of ion transport. These continuous, passive measurements of ion confinement are compared with results from the transient injection method. Absolute levels of the X-ray continuum, adjacent to the Cl(XVI) lines, are used to derive Zeff. Both the Bragg and Johann X-ray spectrometers have been calibrated for absolute photon sensitivity. The efficiencies of the diffractors in these instruments can be measured using a newly developed, two-axis goniometer.

This paper presents a spectroscopic study of deuterium and impurity radiation in the MkI divertor of JET in different experimental regimes. Radiated power measurements using absolutely calibrated VUV spectroscopy are consistent with bolometer measurements in a variety of impurity scenarios. These observations, combined with measurements of visible emission, are compared with code predictions of divertor radiation distributions, highlighting uncertainties in the description of the background plasma and of the carbon source, which affect detailed modelling of divertor impurity transport.

The experience at JET and in other tokamaks shows that the achievement of reproducible plasma conditions and the improvement on overall plasm performance depend crucially on vessel conditioning. The new pumped divertor, with its coils internal to the vacuum vessel, introduces new elements in the problem of JET wall conditioning the presence inside the vessel of hidden surfaces not fully exposed to the glow discharge and of non bakable elements, such as the divertor coils case and heat shields. The new cleaning and conditioning techniques, adapted to the changes of the in-vessel configuration, are described in this paper, and the results of the vessel glow cleaning with an improved GDC system presented. For the first time at JET, the glow discharge cleaning has been extensively diagnosed and the glow current distribution, edge potential, density and temperature measured in conjunction with systematic scan of the glow species current and pressure. A statistical survey of a large number of plasma pulses shows that the “base level” impurity content of the plasma is low, and comparable to the 1991 levels.

Pellet enhanced performance (PEP) has been observed in a number of JET discharges at various plasma conditions, in both L and H modes, with the H multiplier (the confinement enhancement factor over the Goldston confinement time) covering the range from 1 to 4, and with plasma currents from 1 MA to 4.1 MA. Most of the PEP plasmas have been created by refuelling with pellets of 4 mm diameter injected at 1.2 km/s. PEPs show an improved central confinement with an effective heat conductivity reduced by factors of approximately 2-5 relative to otherwise comparable discharges. This is possibly related to the inverted shear in the plasma core due to the large local bootstrap current density. The limitations in the PEP performance seem to be set by at least two mechanisms: impurity behaviour, MHD activity or a combination of both. In certain discharges, MHD modes seem to be able to check the often observed impurity accumulation. Too much MHD mode activity, however, easily destroys the enhanced confinement of the PEP discharge. The stability of the ballooning modes has been studied and the PEP plasma core is found to be in the second stability region against ballooning modes or close to marginal stability. In a number of discharges complex high (m,n) modes have been observed with the soft X-ray cameras. The behaviour of the low (m,n) MHD modes can only be understood by considering the detailed evolution of the inverted q profile, which exists in a given discharge

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

The properties of JET plasmas obtained under similar L mode conditions, but using different ion species, are compared. In hydrogenic limiter discharges with NBI heating, particle confinement and sawtooth activity display a clear isotopic dependence. However, the gain in global energy confinement from hydrogen to deuterium is less pronounced than that implied by popular L mode scaling relationships and the improvement appears to be only partly due to a reduction in the local conductive heat flux. No significant change in energy or particle confinement is observed when deuterium is replaced by 3He

Recently calculated electron impact excitation rates for transitions in helium-like CI XVI are used to derive the electron temperature sensitive emission line ratio G = [(x + y + z)/w] and the electron density sensitive ratio R = [z/(x + y)] where w, x, y and z are the resonance 1s2 1S0-1s2p 1P1, intercombination 1s2 1S0-1s2p 3P2,1 and forbidden 1s2 1S0-1s2s 3S1 transitions, respectively. These calculations, which differ from those of previous authors, are compared with experimental R and G ratios obtained from X-ray spectra from the JET (Joint European Torus) plasma, for which the electron temperature and density have been measured by independent methods. The R ratio is approximately in its low density limit (Ro) for this device and is therefore not electron density sensitive. However it does function as an electron temperature diagnostic as does the G ratio. A comparison is also made with the previously reported results from the Alcator-C tokamak, where the experimental results lie in the electron density sensitive region. The current results show good agreement between theory and observation, with discrepancies of typically 7% in Ro and 10% in G for JET, implying that the theoretical results may be applied to the analysis of remote plasma sources in order to derive values for these parameters, where no independent measurements exist. A departure from coronal equilibrium due to diffusion effects is taken into account for the JET plasma conditions, and the resulting non-coronal ionization balance is found to produce changes to the theoretical ratios at temperatures exceeding Tm.

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 D-D fusion reactivity from JET plasmas has been optimized by using combined neutral beam injection and radiofrequency heating. Reaction rates of up to 2.9 x 10(16) reactions per second (equivalent to neutron emission rates of 1.45 x 10(16) neutrons per second) have been achieved. The best results were obtained in the double-null magnetic configuration or with plasmas limited by the inner wall. By far the largest fraction of the neutron production originates from beam induced reactions, although, in some cases, about 25% of the production can be attributed to the acceleration of deuterium ions by the radiofrequency. The instantaneous neutron yield was limited by a strong influx of carbon impurity ions due to excessive, but local, wall heating.

The combination of two regimes of enhanced performance, the H-mode and the pellet enhanced performance (PEP) mode, has been achieved in JET. The strong enhancement of the central plasma parameters, obtained with pellet injection and subsequent auxiliary heating, is found to persist well into the H-mode phase. A characteristic of the PEP regime is that an improvement of the fusion reactivity over non-pellet discharges is obtained under the condition of nearly equal electron and ion temperatures. A maximum neutron production rate of 0.95 ? 10l6 s?1 was obtained in a double-null X-point discharge with 2.5 MW of neutral beam heating and 9 MW of ion cyclotron resonance heating, with central ion and electron temperatures of about 10 keV and a central deuterium density of 8.0 ? 1019 m?3. The corresponding fusion product nD(0)?ETi(0) is between 7.0 and 8.6 ? 1020 m?3?s?keV. The enhanced neutron production is predominantly of thermonuclear (Maxwellian) origin. The compatibility of these regimes is an important issue in the context of tokamak ignition strategies. Several technical developments on JET have played a role in the achievement of this result: (1) the use of low voltage plasma breakdown (0.15 V/m) to permit pellet injection in an X-point configuration before the formation of a q = 1 surface; (2) the elimination of ICRH specific impurities with antenna Faraday screens made of solid beryllium; (3) the use of a novel system of plasma radial position control that stabilizes the coupling resistance of the ion cyclotron heating system.

Observations of the JET tokamak are used to illustrate contrasting spectral emission from a range of plasma environments differing in temperature, density, interaction with boundary surfaces, dynamic state and excitation mechanisms. Some spectral features of diagnostic value are identified. Models based on collisional radiative theory are used to investigate these features. The generality of this approach to modelling the radiating properties of atoms in arbitrary dynamic plasmas is indicated.

H-modes produced by ion-cyclotron resonance heating (ICRH) in a double-null X-point open divertor configuration in the JET Tokamak are studied, where the ICRH antennas, operated in the dipole (0, pi ) phasing, are located on the low-field side of the Tokamak. Beryllium evaporation on the nickel antenna screen and the first wall of the Tokamak played a crucial role in achieving H-modes with ICRH alone. In these discharges, all the characteristics typical of neutral-beam injection heating H-modes are found, and the global energy confinement time approaches two times the Goldston L-mode prediction. For most of their duration, ICRH H-modes are free from ELMs (edge-localized modes). In addition, the monster sawtooth (long sawtooth-free period) feature of ICRH is maintained during H-modes, leading to a central electron temperature (Tc0) of 10 keV, nearly twice the value of the central ion temperature (Ti0). ICRH H-modes often occur as a two-step transition and the antenna plasma coupling resistance (Rc) also decreases in two steps. Theoretical values of Rc agree well with experimental values during the H-phase.

The scrape-off layer of a JET X-point discharge has been modelled with a multi-fluid plasma model linked to a Monte Carlo model for the neutral deuterium and impurity species. A relatively complete set of boundary plasma data has been obtained, is shown to be consistent within the experimental uncertainties and suggests significant shielding of the carbon impurities released at the graphite target plate. The agreement with the main experimental parameters, n e , T e , is good, as also is the match with the observed radiation and impurity distribution provided that a rather large plasma flow is included in the model.

Beryllium films (150 to 300 A) have been evaporated on all JET in-vessel components (carbon limiters, nickel screen, inconel walls). Immediate and lasting improvements of the plasma have been observed in all plasma regimes. We concentrate on the results obtained with ICRH and NBI on the density limit and on H-mode plasmas which are particularly sensitive to impurity release. Specific effects produced by RF waves are reduced to a negligible level when the following 3 aspects are combined - beryllium evaporation, dipole phasing, correct screen angle. The density limit obtained with ICRF is now identical to NBI heated cases and substantially higher than the values obtained in the pre-beryllium phase. Long H-modes (1.3 s) have been obtained with RF alone. The H mode confinement is similar to that of NBI cases. In limiter plasmas Zeff decreased to 1.2–1.4 at high density and for 10 MW input power. The results are related to oxygen gettering, suppression of nickel sputtering from the antenna screens and to the use of dipole phasing.

An analysis of helium exhaust experiments on JET in the MkII-GB divertor configuration is presented. Helium is pumped by applying an argon frost layer on the divertor cryo pump. Measurement of the helium retention time, tau(He)(*),, is performed in two ways: by the introduction of helium in gas puffs and measurement of the subsequent decay time constant of the helium content, tau(He)(d*); and by helium beam injection and measurement of the helium replacement time, tau(He)(r*). In ELMy H-mode, with plasma configuration optimized for pumping, tau(He)(d*) approximate to 7.2 x tau(E)(th) is achieved, where tau(E)(th) is the thermal energy replacement time. For quasi-steady internal transport barrier (ITB) discharges, the achieved tau(He)(r*) approximate to 4.1 x tau(E)(th) is significantly lower. The achieved helium recycling coefficient, confirmed by an independent measurement to be R-eff approximate to 0.91, is the same in both scenarios. None of the discharges are dominated by core confinement. The difference in tau(He)(*)/tau(E)(th) is instead due to the confinement properties of the edge plasma, which is characterized by Type I ELMs for the H-mode discharges studied, and Type III ELMs for the quasi-steady ITB discharges. This difference is quantified by an independent measurement of the ratio of the helium replacement time with a helium edge source to the energy confinement time.

In the initial phase of ITER H plasmas will be used in order to avoid activating the machine. The reference ICRF healing scenarios rely on minority species such as Helium (3He) or deuterium (D). These schemes’ distinctive feature comes from the presence of the fast magnetosonic wave ion‐ion hybrid resonance/cut‐off pair, between the antennas and the minority cyclotron layer. In order to document these unusual heating schemes, ICRF experiments were carried out recently on JET. First, the use of 3He ions in H plasmas was investigated with a sequence of discharges in which 5 MW of ICRF power was coupled to the plasma and the 3He concentration was varied from below 1% up to 10%. The inverted minority heating regime was observed at low concentrations (up to ∼2%). Energetic tails in the 3He distribution were observed with effective temperatures up to 300 keV and central electron temperatures up to 6 keV. At around 2%, a sudden transition was reproducibly observed to the mode conversion regime, in which the ICRF...

Impurity seeding in both the Joint European Torus (JET) and DIII-D tokamaks has produced L-mode discharges with confinement enhancements comparable to H-mode and a near doubling of the core ion temperature when compared,to similar unseeded discharges. Although Z(eff) increases with the neon injection, the total neutron rate is as high, or higher, than reference discharges and the calculated thermal neutron rate increases dramatically in both devices. Modelling with the gyrokinetic simulation code shows a reduction in low k turbulence growth rates with neon injection decreasing to less than the E x B shearing rate, consistent with stabilization of ion temperature gradient modes in both JET and DIII-D. Reductions in ion thermal diffusivity are also observed with impurity seeding. Neoclassical m/n = 3/2 tearing modes limit the duration of best performance in DIII-D with neon injection, while n = 1 and n = 2 magnetohydrodynamic modes limit the performance in JET.

Recent experiments at the Joint European Torus [Rebut , Fusion Eng. Des. 22, 7 (1993)] aim to improve confinement quality in high-confinement-mode (H-mode) plasmas at high densities. Energy confinement time as predicted by the International Thermonuclear Experimental Reactor ITER-H98(y,2) scaling at densities near or in excess of 85% of the Greenwald density limit scaling has been obtained by (i) strong plasma shaping (triangularity 0.35

Recent R-matrix calculations of electron impact excitation rates in Ni XII are used to derive the emission line ratios R1 = I (154.17 Å)/I (152.15 Å), R2 = I (152.95 Å)/I (152.15 Å) and R3 = I (160.55 Å)/I (152.15 Å). This is the first time (to our knowledge) that theoretical emission line ratios have been calculated for this ion. The ratios are found to be insensitive to changes in the adopted electron density (Ne) when Ne >= 5 × 10^11 cm−3, typical of laboratory plasmas. However, they do vary with electron temperature (Te), with for example R1 and R3 changing by factors of 1.3 and 1.8, respectively, between Te = 10^5 and 10^6 K. A comparison of the theoretical line ratios with measurements from the Joint European Torus (JET) tokamak reveals very good agreement between theory and observation for R1, with an average discrepancy of only 7%. Agreement between the calculated and experimental ratios for R2 and R3 is less satisfactory, with average differences of 30 and 33%, respectively. These probably arise from errors in the JET instrument calibration curve. However, the discrepancies are smaller than the uncertainties in the R2 and R3 measurements. Our results, in particular for R1, provide experimental support for the accuracy of the Ni XII line ratio calculations, and hence for the atomic data adopted in their derivation.

The main populating and depopulating mechanisms of the excited energy levels of ions in plasmas with densities, 29 pages, 5 figures. This is an author-created, un-copyedited version of an article accepted for publication in Journal of Physics B. IOP Publishing Ltd is not responsible for any errors or omissions in this version of the manuscript or any version derived from it

The soft x-ray spectral region betwn 0.1 nm and 10 nm includes the peak of the radiated power and contains a wide range of diagnostic information about the temperature and density of electrons and impurity ions. Spectroscopic coverage of this band will be essential for ITER.

Additional heating systems can be a source of impurities in fusion plasmas. Studying the behaviour of such impurities is important to understand and minimize their effects on tokamak plasma performance. In this paper we present a detailed study of the influence of ion cyclotron resonance heating (ICRH) on the impurity content of JET plasmas. Using spectroscopic methods we monitor the Ni impurity release by the two ICRH antenna types in JET: the so-called ITER-like antenna and the A2 antennas. The release of Ni during ICRH is presented as a function of the power density, total ICRH power applied, the relative phasing of the A2 antenna straps, the D2 gas injection level and the plasma–strap distance. We also estimate the contribution of Ni impurity to Zeff and to plasma dilution. L-mode discharges are compared for different relative phasings of the four antenna straps of the A2 antennas: (0 π 0 π), (0 π π 0), (0 0 π π), (00 π/2 π/2), (0 π/2 π 3π/2), (− − π 0) and (0 π − −). We observe that for antenna phasings that allow the best coupling, the Ni impurity concentration is reduced in the central part of the plasma. The impurity content is also monitored in H-mode experiments aiming at improving the ICRH coupling at large antenna strap–separatrix distances, up to 19 cm, by injecting gas from different inlets. Interestingly, with increasing gas injection rate the coupling improves and the Ni concentration decreases, and the actual location of gas injection is found to influence the Ni concentration.

The provision of measurements of metallic impurity densities, Zeff and dilution for a large number of discharges within a campaign facilitates the analysis of impurity trends. Such trends are of increasing importance as additional heating power and pulse length increase. This is particularly important for RF heating and therefore it is in particular relevant to the assessment of the ITER-like ICRF antenna (ILA) on JET. To this end, a method is presented for determining the metal impurity density, ΔZeff and dilution in steady-state JET plasmas using passive VUV emission. The method is based on the combination of absolutely calibrated VUV transition intensity measurements with Universal Transport Code (UTC) simulations. In the analysis the line-integrated measurements of transitions in Li-like Ni, Fe and Cu have been used for test discharges characterized by widely varied plasma profiles. The simulations use a wide class of transport coefficients for diffusion D(r) and convection V(r). For a given pair of D(r) and V(r), the simulated line intensity has been matched to the line intensity measured in the experiment. An approximately linear dependence of the derived metal densities, ΔZeff and dilution normalized to a Li-like line intensity on electron temperature has been obtained which is valid in a localized, mid-radius plasma region. These linear dependences are exploited to derive local metal densities for JET discharges.

A description of the radiation emitted by impurities from within a plasma is crucial if spectral line intensities are to be used in detailed studies, such as the analysis of impurity transport. The simplest and most direct check that can be made on measurements of line intensities is to analyse their ratios with other lines from the same ion. This avoids uncertainties in determining the volume of the emitting plasma and the absolute sensitivity calibration of the spectrometer and, in some cases, the need even for accurate measurements of parameters such as electron density. Consistency is required between the measured line intensity ratios and the theoretical values. The expected consistency has not been found for radiation emitted from the JET scrape-off layer (e.g. Lawson et al 2009a JINST 4 P04013), meaning that the description of the spectral line intensities of impurity emission from the plasma edge is incomplete. In order to gain further understanding of the discrepancies, an analysis has been carried out for emission from the JET divertor plasma and this is reported in this paper.Carbon was the main low Z intrinsic impurity in JET and an analysis of spectral line intensity ratios has been made for the C?IV radiation emitted from the JET divertor. In this case, agreement is found between the measured and theoretical ratios to a very high accuracy, namely to within the experimental uncertainty of ~?10%. This confirms that the description of the line intensities for the present observations is complete. For some elements and ionization stages, an analysis of line intensity ratios can lead to the determination of parameters such as the electron temperature of the emitting plasma region and estimates of the contribution of recombination to the electron energy level populations. This applies to C?IV and, to show the value and possibilities of the spectral measurements, these parameters have been calculated for a database of Ohmic and additionally heated phases of a large number of pulses. The importance of dielectronic, radiative and charge-exchange recombination as well as ionization has been investigated. In addition, the development of Te throughout two example discharges is illustrated. The presented results indicate a number of areas for further investigation.

Collision strengths (Ω) and effective collision strengths (γ) are calculated for 6328 transitions among the lowest 113 levels of Kr XXVIII, which belong to the 2s22p5, 2s2p6, 2s22p43l, 2s2p53l, and 2p63l configurations. For the generation of wavefunctions fully relativistic GRASP code is adopted whereas for the scattering process DARC is employed. Values of Ω are calculated up to an energy of 400 Ryd which is sufficient to calculate results for γ up to a temperature of 107 K. Resonances have been resolved in a fine energy mesh in the thresholds region and their contribution is observed to be significant for many transitions, particularly the forbidden ones.

The European Fusion Development Agreement's mission for JET is the development of ITER scenarios exploiting the specific properties of the device. This task requires significant improvements in the measuring techniques. The most innovative diagnostic upgrades are in the fields of edge measurements, detection of fast magnetohydrodynamics modes and "burning plasma" diagnostics. The importance of plasma-wall interactions, and, in particular, the issue of tritium inventory promoted the development of the quartz microbalance, a detector with improved time resolution to measure material redeposition in the remote areas of the inner divertor. Measurement of Alfven cascades with unprecedented spectral resolution, reaching a toroidal n number of up to 16, was obtained using an O-mode microwave reflectometer as an interferometer. For the diagnosis of the fusion products, a new approach is being developed to measure the He ash based on double charge exchange between thermalized particles and neutrals from JET beams. There have been several upgrades of the neutron diagnostic systems, one of which, the new time of flight neutron spectrometer designed for high counting rates is described. (c) 2006 American Institute of Physics.

Crystal and synthetic multilayer diffractors, deployed either as flat Bragg reflectors, or curved, as in the Johann configuration, have been used to study the spectrum of COMPASS-D and other tokamaks in the wavelength region 1–100 A. In particular, line emission from CIXVI and other He-like ions of the third atomic period elements have been shown to be a rich source of diagnostic information on ion transport and impurity ion accumulation, ion and electron temperatures, and plasma fluid velocities. In this article, we concentrate on the measurement of absolute photon fluxes and the derivation of volume emissivities of the lines and continua in the x-ray region. The sensitivities of these instruments to absolute photon flux are constructed from the individual component efficiencies, including published values of the diffractor reflectivities. Where diffractors have been used for which there is no published information, the reflectivity is measured using a double-axis goniometer or from line branching ratios...

Heating the JET plasma well above temperatures reached in the ohmic phase is the aim of the two additional heating systems planned for JET: ion cyclotron resonance heating (ICRF) and neutral beam injection (NBI). Operations with the latter started in February 1986, initially with hydrogen injection, up to a power level of 7 MW. ICRF power has been delivered to the plasma by three antennae and has reached power levels of 6 MW for 2 s. In most experiments, the frequency of the waves is adjusted to position the minority ion resonance layer close to the centre of the discharge, resulting in a centrally peaked power deposition profile. Results have also shown that the increase of the volume averaged electron temperature was much less dependent on the radial position of the ion resonance layer than were the central temperatures of both ions and electrons which can both approach 5 keV. In cases of 'on axis' heating, large increases in the sawtooth activity is observed with sawtooth periods exceeding 0.3 s. Initial results with NBI have shown a decrease of the global energy confinement time with additional power similar to the one observed with ICRF heating.

Wavelengths and transition probabilities are presented for the allowed and forbidden n=2-2 transitions belonging to the fluorine I to lithium I isoelectronic sequences which are expected to be observed in Tokamak discharges. The listing is compiled for the elements Ti, Cr, Fe, Co and Ni, these metals being likely contaminants. The emphasis is placed on an accurate compilation of wavelengths from laboratory and solar-flare observations. Where these are inadequate, predictions are given. In the case of the nitrogen-, carbon- and boron-like intercombination lines, these predictions represent a significant improvement on previous data. Theoretical line intensities calculated for conditions typical of a Tokamak plasma are given for each transition. Identifications in Ti of the five strongest forbidden transitions lying in the visible and near-UV spectral regions have been made in spectra from the DITE Tokamak. Such spectral features are useful diagnostic indicators of metallic ion impurity concentrations and mass motions.

Over 400 new identifications of emission lines in the grazing-incidence spectra of the elements from chromium to nickel are reported. The great majority of these lines belong to the 2sn2pk-2sn-12pk+1 transition arrays in the O I to Be I isoelectronic sequences. Some of the spectra belonging to the N I, C I, B I and Be I sequences are identified as intercombination lines. The new data, taken together with other published results in the F I to Li I isoelectronic sequences, allow a complete listing of 2sn2pk-2sn-12pk+1 transitions in the elements from chromium to nickel, with their appropriate term schemes. The term schemes lead to wavelength predictions for various forbidden transitions. The use of these allowed and forbidden transitions has been proposed by other authors for diagnostic purposes in solar-flare and laboratory plasmas. In addition to the n=2-2 transitions the wavelengths of some Na I-like ions of the elements Cr to Ni are presented. The plasma source used to obtain the spectra and to differentiate between the emission from separate ionisation stages was created by focusing the output of a 40 GW neodymium glass laser system onto plane metallic targets.

There have been discussions in the recent literature regarding the accuracy of the available electron impact excitation rates (equivalently effective collision strengths $\Upsilon$) for transitions in Be-like ions. In the present paper we demonstrate, once again, that earlier results for $\Upsilon$ are indeed overestimated (by up to four orders of magnitude), for over 40\% of transitions and over a wide range of temperatures. To do this we have performed two sets of calculations for N~IV, with two different model sizes consisting of 166 and 238 fine-structure energy levels. As in our previous work, for the determination of atomic structure the GRASP (General-purpose Relativistic Atomic Structure Package) is adopted and for the scattering calculations (the standard and parallelised versions of) the Dirac Atomic R-matrix Code ({\sc darc}) are employed. Calculations for collision strengths and effective collision strengths have been performed over a wide range of energy (up to 45~Ryd) and temperature (up to 2.0$\times$10$^6$~K), useful for applications in a variety of plasmas. Corresponding results for energy levels, lifetimes and A-values for all E1, E2, M1 and M2 transitions among 238 levels of N~IV are also reported., Comment: This paper with 5 Figs. and 8 Tables will appear in MNRAS (2016)