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The role of the magnetic field topology for the complex interplay between turbulence and zonal flows (ZFs) leading to the formation of transport barriers is a recent topic in fusion research. Gaining knowledge on their relation in experiments requires a flexible magnetic configuration and a good diagnostic access with high temporal and spatial resolution. WEGA as a small classical stellarator with moderate plasma parameters is well suited for such experiments. Stationary n = 1 islands can be modified in size by external error field compensation coils. Configurations with m/n = 5/1 islands at the plasma edge are compared with configurations showing no major resonances (i.e. no islands). The spatio-temporal structure of fluctuations was studied with Langmuir-probe arrays at different toroidal and poloidal positions. In low density discharges ZF-like patterns were observed in potential fluctuations. A characteristic feature of these oscillations is a long range correlation (LRC) with fluctuations being symmetric on a flux surface. Itwas shown for the first time that the presence of magnetic islands alters the structure of these low frequency fluctuations. At r

• An automated mode recovery for gyrotrons has been implemented on a FPGA. • The algorithm exploits the hysteretic gyrotron behaviour. • The nominal mode is recovered in ≤1 ms after a switch to the competing satellite. • The total output power could be increased by at least 500 kW for the same pulse length. • The solution offers a very attractive cost-benefit ratio. The reliability of gyrotron operation is significantly decreased approaching the individual maximum output power of the tube due to loss of the nominal operating mode. For the first time, this paper proposes an algorithm for an automated mode recovery (MORE) for gyrotrons exploiting the hysteretic gyrotron behaviour. The algorithm has been implemented in a field-programmable gate array (FPGA) controlling the acceleration voltage and is able to recover the nominal operating mode within ≤1 ms after a mode switch to the competing satellite mode. This allows the gyrotron to be operated closer to its stability limits with extended pulse lengths at potentially higher output power. Dedicated experiments to test MORE were conducted at Wendelstein 7-X (W7-X) with two gyrotrons using a beam dump. The nominal mode could be successfully recovered in 99% of 3755 modeloss events during 128 shots up to 20 s. MORE was operational for nine out of ten gyrotrons during the last experimental campaign OP1.2b of W7-X. The overall success rate during W7-X OP1.2b was 91% counting 464 modeloss events in 131 shots performed with seven gyrotrons. Using the mode losses in the working point plane, the cutoff region for the nominal working mode was identified, defining a minimum cathode current for a given acceleration voltage. The total achievable output power of the W7-X ECRH plant could be increased by at least 500 kW for the same pulse length using MORE, assuming a conservative increase of at least 50 kW per gyrotron. Comparing the output power for the same achievable pulse length and reliability level, the output power increase per gyrotron is likely in the order of 100 kW. Since MORE exploits the hysteretic gyrotron behaviour, it could be applied to other gyrotrons of already existing or future ECRH facilities of fusion experiments, like ITER. [ABSTRACT FROM AUTHOR]

Electron cyclotron resonance heating (ECRH) is a powerful and flexible plasma heating technique that serves as the main heater at Wendelstein 7-X (W7-X) and will be used at ITER for start-up, heating, current drive and mitigation of plasma instabilities. In the case of poor or degraded microwave absorption, which is expected in the O2-mode heating scenario, a significant part of the beam directly hits the wall, leading to local overheating and potential damage. The ECRH shine-through power is mostly reflected onto the targets; only a small fraction is really absorbed through ohmic losses (typically 3% for graphite at 140 GHz). The ohmic losses do not only depend on the material properties and the frequency, but also on the polarization of the wave and the angle of incidence. This paper presents a thermographic analysis of ECRH experiments at W7-X, including heat load and temperature simulations of the first wall that include ECRH shine through. Two O-mode ECRH experiments with both a high temperature rise of the first wall and different angles of beam incidence on the wall's surface are depicted. One experiment has 775 kW of power modulation (5 Hz) with mixed polarization (45% O-mode, 55% X-mode) and an EC beam angle almost normal to the first wall. The second has 550 kW of steady EC power with O-mode polarization, a shallow beam angle and increased power absorption by the material. It is shown that infrared thermography is a useful tool for measuring shine-through power and protecting wall components. [ABSTRACT FROM AUTHOR]

In this paper, we present the design and commissioning results of the upgraded collective Thomson scattering diagnostic at the Wendelstein 7-X stellarator. The diagnostic has a new radiometer designed to operate between the second and third harmonics of the electron cyclotron emission from the plasma at 171–177 GHz, where the emission background has a minimum and is of order 10–100 eV. It allows us to receive the scattered electromagnetic field with a significantly improved signal-to-noise ratio and extends the set of possible scattering geometries compared to the case of the original instrument operated at 140 GHz. The elements of the diagnostic are a narrowband notch filter and a frequency stabilized probing gyrotron that will allow measuring scattered radiation spectra very close to the probing frequency. Here, we characterize the microwave components applied to the radiometer and demonstrate the performance of the complete system that was achieved during the latest experimental campaign, OP2.1. [ABSTRACT FROM AUTHOR]

Abstract: The heat flux to the target plates qp is a crucial design parameter for fusion machines. It can be measured by infrared (IR) cameras viewing the target plates. The sheath heat transmission coefficient γ is defined by qp = γkTe jtarget relating qp to particle fluxes and temperature. Sheath theory predicts γ in the range of 5…8 We show γ determined experimentally from a comparison of qp from IR with Te and jtarget measured by Langmuir probes for low to high recycling regimes in L-mode with carbon targets. While γ is in the expected range at low recycling, unphysically low values are observed at high recycling. We compare heat flux measured on carbon and tungsten as target material in reproducible discharges. From these data it could be concluded that there is no measurable difference in γ for the two materials. We finally discuss the influence of γ on simulation results. [Copyright &y& Elsevier]

Here we report the first measurements of the power levels of stray radiation in the vacuum vessel of Wendelstein 7-X using absolutely calibrated sniffer probes. The absolute calibration is achieved by using calibrated sources of stray radiation and the implicit measurement of the quality factor of the Wendelstein 7-X empty vacuum vessel. Normalized absolute calibration coefficients agree with the cross-calibration coefficients that are obtained by the direct measurements, indicating that the measured absolute calibration coefficients and stray radiation levels in the vessel are valid. Close to the launcher, the stray radiation in the empty vessel reaches power levels up to 340 kW/m2 per MW injected beam power. Furthest away from the launcher, i.e., half a toroidal turn, still 90 kW/m2 per MW injected beam power is measured. [ABSTRACT FROM AUTHOR]

Experimental evidence of parametric decay instability (PDI) is observed in the Wendelstein 7-X stellarator, when high-power microwave beams cross a stationary magnetic island at the plasma edge. Here, trapping and build-up of upper hybrid waves within a density bump (measured within the island by alkali beam emission spectroscopy) is responsible for the reduction of the instability power threshold below the maximum gyrotron power. In this paper, we provide the first experimental evidence of the connection between the trapping mechanism in the island density bump and excitation of PDI-related signals. We show correlations of periodic crashes in the PDI-related signals with quasi-continuous fluctuations at the plasma edge, which, additionally, cause a flattening of the density profile in the island. We demonstrate that flattening of the experimental density profiles can suppress the trapping mechanism and inhibit the low-threshold PDI. PDI on the edge island could alter the power deposition profile and reduce the efficiency of the electron cyclotron resonance heating system, simultaneously posing a serious threat to the optimal operation of microwave-based diagnostics and plasma-facing components. [ABSTRACT FROM AUTHOR]

The Wendelstein 7-X stellarator is a large nuclear fusion device based at Max-Planck-Institut für Plasmaphysik in Greifswald in Germany. The main plasma heating system for steady state operation in W7-X is electron cyclotron resonance heating (ECRH). During operation, part of plama facing components will be directly heated by the non-absorbed power of 1 MW rf beams of ECRH. In order to avoid damages of such components made of graphite tiles during the first operational phase, a near infra-red video system has been developed as a protective diagnostic for safe and secure ECRH operation. Both the mechanical design housing the camera and the optical system are very flexible and respect the requirements of steady state operation. The full system including data acquisition and control system has been successfully tested in the vacuum vessel, including on-line visualization and data storage of the four cameras equipping the ECRH equatorial launchers of W7-X. [ABSTRACT FROM AUTHOR]

The electronic structure of Cu2ZnSnS4 has been investigated by S L2,3 x-ray emission and absorption spectroscopy. An empirical assignment of the most prominent spectral features of the valence and conduction band, respectively, can be made based on a comparison with spectra of corresponding binary compounds and with calculated densities of states. We find that the top of the valence band is dominated by hybridized antibonding Cu 3d-S 3p states, while the bottom of the conduction band is dominated by respective Sn 5s-S 3p states. We also determine the electronic band gap near the Cu2ZnSnS4 surface to be 1.5 (±0.3) eV. [ABSTRACT FROM AUTHOR]

An ion cyclotron emission (ICE) diagnostic is prepared for installation into the W7-X stellarator, with the aim to be operated in the 2022 experimental campaign. The design is based on the successful ICE diagnostic on the ASDEX Upgrade tokamak. The new diagnostic consists of four B-dot probes, mounted about 72° toroidally away (one module) from the neutral beam injector, with an unobstructed plasma view. Two of the B-dot probes are oriented parallel to the local magnetic field, aimed to detect fast magnetosonic waves. The remaining two probes are oriented poloidally, with the aim to detect slow waves. The radio frequency (RF) signals picked up by the probes are transferred via 50 Ω vacuum-compatible coaxial cables to RF detectors. Narrow band notch filters are used to protect the detectors from possible RF waves launched by the W7-X antenna. The signal will be sampled with a four-channel fast analog-to-digital converter with 14 bit depth and 1 GSample/s sampling rate. The diagnostic's phase–frequency characteristic is properly measured in order to allow measuring the wave vectors of the picked up waves. [ABSTRACT FROM AUTHOR]

The optimised superconducting stellarator W7-X generates its rotational transform by means of external coils, therefore no toroidal current is necessary for plasma confinement. Electron cyclotron current drive experiments were conducted for strikeline control and safe divertor operation. During current drive experiments periodic and repetitive crashes of the central electron temperature, similar to sawtooth crashes in tokamaks, were detected. Measurements from soft x-ray tomography and electron cyclotron emission show that the crashes are preceded by weak oscillating precursors and a displacement of the plasma core, consistent with a (m, n) = (1, 1) mode. The displacement occurs within , followed by expulsion and redistribution of the core into the external part of the plasma. Two types of crashes, with different frequencies and amplitudes are detected in the experimental program. For these non-stationary parameters a strong dependence on the toroidal current is found. A 1-D heuristic model for current diffusion is proposed as a first step to explain the characteristic crash time. Initial results show that the modelled current diffusion timescale is consistent with the initial crash frequency and that the toroidal current rise shifts the position where the instability is triggered, resulting in larger crash amplitudes. [ABSTRACT FROM AUTHOR]

Numerical calculations based on nonlinear two-fluid equations have been carried out to understand the sawtooth-like crashes of electron temperature observed in Wendelstein 7-X stellarator experiments with electron cyclotron current drive (ECCD). The application of ECCD leads to non-monotonic radial profiles of the safety factor q with two q = 1 surfaces in the plasma core region. Using input parameters similar to the experiment, numerical calculations show two types of crashes of the central electron temperature depending on the growth of the internal kink mode coupled to double tearing modes. When the distance between two equilibrium q = 1 surfaces is sufficiently large, the internal kink mode leads to a fast radial displacement of the hot core from the center to the region of the q = 1 surfaces. In agreement with experimental observation, this causes a full crash of central electron temperature in about 20–30 microseconds. When the distance between the two q = 1 surfaces is not large enough, partial crashes of the central electron temperature are found. During these crashes the hot core only temporally moves outwards. Then it moves back, because the double tearing mode perturbation leads to an increase of the minimum q value about one. [ABSTRACT FROM AUTHOR]

The Wendelstein 7-X (W7-X) experiment is equipped with an Electron Cyclotron Resonance Heating installation consisting of 10 gyrotrons capable of delivering up to 7.5 MW of Electron Cyclotron Wave power at the 140 GHz resonance in the plasma. During the plasma discharge the gyrotron power is delivered in a narrow band of several 100 MHz around the gyrotron frequency. At plasma start-up, i.e. initially in the absence of plasma, the non-absorbed gyrotron power in the vessel ('stray radiation') is very high. During the plasma discharge the stray radiation may also rise depending on the plasma optical depth or due to the challenge to deliver the gyrotron power at the exact required polarisation. At W7-X an additional issue is that at high electron density the second harmonic extra-ordinary heating scheme (X2) is no longer possible and one switches to second harmonic ordinary (O2) heating. But the ECH absorption is not complete in the O2 heating scheme and stray radiation is significantly increased. The stray radiation issue is a problem for sensitive microwave receivers, typically measuring power levels in the sub μ W range. For protection, these generally employ notch filters such as fundamental mode interference cavities [1] , [2]. This paper focuses on the effect of non-absorbed 140 GHz gyrotron power on the Michelson Interferometer at W7-X. [ABSTRACT FROM AUTHOR]

In this paper, we present a method for numerical computation of collective Thomson scattering (CTS). We developed a forward model, eCTS, in the electrostatic approximation and benchmarked it against a full electromagnetic model. Differences between the electrostatic and the electromagnetic models are discussed. The sensitivity of the results to the ion temperature and the plasma composition is demonstrated. We integrated the model into the Bayesian data analysis framework Minerva and used it for the analysis of noisy synthetic data sets produced by a full electromagnetic model. It is shown that eCTS can be used for the inference of the bulk ion temperature. The model has been used to infer the bulk ion temperature from the first CTS measurements on Wendelstein 7-X. [ABSTRACT FROM AUTHOR]

A Collective Thomson Scattering (CTS) diagnostic is installed at Wendelstein 7-X for ion temperature measurements in the plasma core. The diagnostic utilizes 140 GHz gyrotrons usually used for electron cyclotron resonance heating (ECRH) as a source of probing radiation. The CTS diagnostic uses a quasi-optical transmission line covering a distance of over 40 m. The transmission line is shared between the ECRH system and the CTS diagnostic. Here we elaborate on the design, installation, and alignment of the CTS diagnostic and present the first measurements at Wendelstein 7-X. [ABSTRACT FROM AUTHOR]

At the Wendelstein 7-X stellarator, a 174 GHz Collective Thomson Scattering (CTS) diagnostic will be implemented. One of the 140 GHz Electron Cyclotron Resonance Heating (ECRH) gyrotrons will be operated at around 174 GHz in a higher cavity mode, using it as source for the CTS mm-wave probing beam. To prevent any damage to the CTS receiver, a notch filter cuts out the high-power gyrotron signal at the entrance of the receiver. The bandwidth of the gyrotron signal determines the notch filter bandwidth. First proof-of-principle experiments on frequency stabilization were conducted on W7-X ECRH gyrotrons employing Phase-Locked Loop techniques. The gyrotron output frequency was controlled with the accelerating voltage, which is applied between the anode and cathode of the gyrotron diode-type Magnetron Injection Gun. Frequency stabilization experiments with 10 ms pulses were conducted at the gyrotron nominal frequency of 140 GHz as well as at 174 GHz. It is concluded that the gyrotron frequency could be stabilized for at least 3 ms at 140 GHz and 8 ms at 174 GHz. In the frequency spectrum, a clear main peak of the gyrotron frequency at 140 GHz with a full -15 dB linewidth of below 500 Hz was achieved. [ABSTRACT FROM AUTHOR]

In JET-ILW dedicated melt exposures were performed using a sequence of 3MA/2.9T H-Mode JET pulses with an input power of P IN = 23 MW, a stored energy of ∼6 MJ and regular type I ELMs at Δ W ELM = 0.3 MJ and f ELM ∼ 30 Hz. In order to assess the risk of starting ITER operations with a full W divertor, one of the task was to measure the consequences of W transients melting due to ELMs. JET is the only tokamak able to produce transients/ ELMs large enough (>300 kJ per ELM) to facilitate melting of tungsten. Such ELMs are comparable to mitigated ELMs expected in ITER. By moving the outer strike point (OSP) onto a dedicated leading edge the base temperature was raised within ∼1 s to allow transient ELM-driven melting during the subsequent 0.5 s. Almost 1 mm (∼6 mm 3 ) of W was moved by ∼ 150 ELMs within 5 subsequent discharges. Significant material losses in terms of ejections into the plasma were not observed. There is indirect evidence that some small droplets ( ∼ 80 μ m ) were ejected. The impact on the main plasma parameters is minor and no disruptions occurred. The W-melt gradually moved along the lamella edge towards the high field side, driven by j × B forces. The evaporation rate determined is 100 times less than expected from steady state melting and thus only consistent with transient melting during individual ELMs. IR data, spectroscopy, as well as melt modeling point to transient melting. Although the type of damage studied in these JET experiments is unlikely to be experienced in ITER, the results do strongly support the design strategy to avoid exposed edges in the ITER divertor. The JET experiments required a surface at normal incidence and considerable pre-heating to produce tungsten melting. They provide unique experimental evidence for the absence of significant melt splashing at events resembling mitigated ELMs on ITER and establish a unique experimental benchmark for the simulations being used to study transient shallow melting on ITER W divertor PFUs. [ABSTRACT FROM AUTHOR]

The transition reported in L-mode plasmas leading to the formation of a density “shoulder” in the Scrape-off Layer (SOL) of tokamaks is investigated in the three ITER stepladder machines: COMPASS, ASDEX Upgrade (AUG) and JET. In AUG and JET, the broadening of the profiles can be observed at similar relative densities ( n e / n GW ∼ 0.4–0.5) and coincides with a clear increase in the size of SOL filaments and the onset of the outer divertor detachment. In COMPASS, no transition is observed in the same density range: the divertor does not detach and no clear effect is observed neither in the profiles nor in the turbulence. This is explained as the result of an increase of intermittent transport triggered by the increased resistivity along filaments associated to divertor detachment. The transition conditions in AUG and JET are compared to foreseen ITER SOL parameters, and the implications are briefly discussed. [ABSTRACT FROM AUTHOR]

Two classes of nitrogen seeded pulses are here considered: pulses at high triangularity, high confinement with moderate gas puffing and N seeding rates and pulses at low triangularity, lower confinement with higher gas puffing and N seeding. For the high-delta pulses the power radiation fraction, f rad , is normally about 0.5 while at low delta f rad as high as 0.7 is achieved. For the simulations we have used COREDIV code, which self-consistently couples the plasma core with the plasma edge and the main plasma with impurities. To reproduce numerically the main experimental parameters of the two classes of pulses two different settings had to be applied to COREDIV, as the perpendicular transport and recycling in the SOL and the power deposition profile in the core. Simulations suggest that higher values for f rad might be achieved with N seeding only at low density and/or higher heating power. [ABSTRACT FROM AUTHOR]

Combining Lithium beam emission spectroscopy and edge reflectometry, localized Scrape-Off Layer (SOL) density modifications by Ion Cyclotron Range of Frequencies (ICRF) near fields were characterized in JET L-mode plasmas. When using the ICRF wave launchers connected magnetically to the Li-beam chord, the density decreased more steeply 2–3 cm outside the last closed flux surface (mapped onto the outer mid-plane) and its value at the outer limiter radial position was half the ohmic value. The depletion depends on the ICRF power and on the phasing between adjacent radiating straps. Convection due to ponderomotive effects and/or E × B 0 drifts is suspected: during ICRF-heated H-mode discharges in 2013, DC potentials up to 70 V were measured locally in the outer SOL by a floating reciprocating probe, located toroidally several metres from the active antennas. These observations are compared with probe measurements on ASDEX-Upgrade. Their implications for wave coupling, heat loads and impurity production are discussed. [ABSTRACT FROM AUTHOR]

In the recent melt experiments with the JET tungsten divertor, we observe that the heat flux impacting on a leading edge is 3–10 times lower than a geometrical projection would predict. The surface temperature, tungsten vaporisation rate and melt motion measured during these experiments is consistent with the simulations using the MEMOS code, only if one applies the heat flux reduction. This unexpected observation is the result of our efforts to demonstrate that the tungsten lamella was melted by ELM induced transient heat loads only. This paper describes in details the measurements and data analysis method that led us to this strong conclusion. The reason for the reduced heat flux are yet to be clearly established and we provide some ideas to explore. Explaining the physics of this heat flux reduction would allow to understand whether it can be extrapolated to ITER. [ABSTRACT FROM AUTHOR]

An experimental database of the power spreading into the target private flux region (S-factor) from ASDEX Upgrade and JET outer target data has been gathered and analysed. The power spreading, combined with the Scrape-Off Layer heat flux decay length λ q , determines the power exhaust channel width at the divertor target. In terms of main plasma parameters, we find that the poloidal magnetic field plays a major role S ∼ B pol - 1 in both L-mode and H-mode plasmas, similarly to λ q . Combining JET and AUG data in H-mode with open divertor we obtain a strong and beneficial major radius dependency and no plasma density influence is found. For closed AUG divII configurations in L-mode, however, larger upstream n e clearly broadens S . Interestingly, the scaling of S with divertor parameters, reveals that only target T e is needed to account for its variation. SOLPS simulations with fixed perpendicular transport reproduce most of the observed phenomenology, suggesting that the observed larger S in closed versus open divertor configuration is simply due to different target T e generated by larger parallel gradients. [ABSTRACT FROM AUTHOR]

We present the application of an improved EDGE2D-EIRENE SOL transport model for the ELM phase utilizing kinetic correction of the sheath-heat-transmission coefficients and heat-flux-limiting factors used in fluid SOL modelling. With a statistical analysis over a range of similar type-I ELMy H-mode discharges performed at the end of the first JET ITER-like wall campaign, we achieved a fast (Δ t = 200 μs) temporal evolution of the outer midplane n e and T e profiles and the target-heat and particle-flux profiles, which provides a good experimental data set to understand the characteristics of an ELM cycle. We will demonstrate that these kinetic corrections increase the simulated heat-flux-rise time at the target to experimentally observed times but the power-decay time at the target is still underestimated by the simulations. This longer decay times are potentially related to a change of the local recycling coefficient at the tungsten target plate directly after the heat pulse. [ABSTRACT FROM AUTHOR]

The impact of the divertor geometry on global plasma confinement in type I ELMy H-mode has been investigated in the JET tokamak equipped with ITER-Like Wall. Discharges have been performed in which the position of the strike-points was changed while keeping the bulk plasma equilibrium essentially unchanged. Large variations of the global plasma confinement have been observed, the H 98 factor changing from typically 0.7 when the outer strike-point is on the vertical or horizontal targets to 0.9 when it is located in the pump duct entrance. Profiles are mainly impacted in the pedestal but core gradient lengths, especially for the density, are also modified. Although substantial differences are observed in the divertor conditions, none seem to correlate directly with the confinement. Modelling with the EDGE2D-EIRENE and SOLEDGE2D-EIRENE transport codes exhibits differences in the energy losses due to neutrals inside the separatrix, but orders of magnitude are too low to explain simply the impact on the confinement. [ABSTRACT FROM AUTHOR]

Vacuum windows are required in magnetically confined fusion experiments to provide possibilities to observe the plasma in a wide range of electromagnetic wavelengths. The window disk consists of a dielectric, e.g. Fused Silica (SiO 2), Sapphire or Chemically Vapourised Diamond (CVD). As electromagnetic waves pass through the disk, a fraction of the beam power is dissipated resulting in a temperature increase of the disk. In Electron Cyclotron Waves (ECW) heated plasmas the dissipation in the window disk can be very high. The computation of dielectric losses for a collimated beam with known incidence angle, polarisation and loss tangent (measure for the intrinsic dielectric loss) is well established. However, the dielectric losses in diagnostic windows mostly result from microwave stray radiation , which results from a modest, but inevitable, fraction of non-absorbed ECW. This fraction diffuses in the vessel by many reflections into rays with random k-vector and with random polarisation. In this work the thermal load on the window disk by microwave stray radiation is assessed. The load by a collimated beam is studied as a function of incidence angle and polarisation allowing to average over a distribution of incident rays. An experiment was commissioned measuring the loss tangent of a number of commercially available SiO 2 disks at low power in an open resonator, and subsequently measuring the dielectric heating of these disks at high power stray radiation using the facility 'MISTRAL' at Wendelstein-7X. The experimental results are compared to modelling and it is demonstrated that, in the parameter range considered, single-pass fractional absorption may be applied while taking a safety margin that arises from the minima and maxima due to multiple reflections. • Measurement and modelling of stray radiation dielectric losses in vacuum windows. • Measurement data on loss tangent and permittivity at millimetre range frequencies. • Stray radiation exposure of a dielectric slab. [ABSTRACT FROM AUTHOR]

A comparison of the L–H power threshold (Pthr) in JET with all carbon, JET-C, and beryllium/tungsten wall (the ITER-like choice), JET-ILW, has been carried out in experiments with slow input power ramps and matched plasma shapes, divertor configuration and IP/BT pairs. The low density dependence of the L–H power threshold, namely an increase below a minimum density ne,min, which was first observed in JET with the MkII-GB divertor and C wall and subsequently not observed with the current MkII-HD geometry, is observed again with JET-ILW. At plasma densities above ne,min, Pthr is reduced by ∼30%, and by ∼40% when the radiation from the bulk plasma is subtracted (Psep), with JET-ILW compared to JET-C. At the L–H transition the electron temperature at the edge, where the pedestal later develops, is also lower with JET-ILW, for a given edge density. With JET-ILW the minimum density is found to increase roughly linearly with magnetic field, , while the power threshold at the minimum density scales as . The H-mode power threshold in JET-ILW is found to be sensitive both to variations in main plasma shape (Psep decreases with increasing lower triangularity and increases with upper triangularity) and in divertor configuration. When the data are recast in terms of Psep and Zeff or subdivertor neutral pressure a linear correlation is found, pointing to a possible role of Zeff and/or subdivertor neutral pressure in the L–H transition physics. Depending on the chosen divertor configuration, Pthr can be up to a factor of two lower than the ITPA scaling law for densities above ne,min. A shallow edge radial electric field well is observed at the L–H transition. The edge impurity ion poloidal velocity remains low, close to its L-mode values, ⩽5 km s−1 ± 2–3 km s−1, at the L–H transition and throughout the H-mode phase, with no measureable increase within the experimental uncertainties. The edge toroidal rotation profile does not contribute to the depth of the negative Er well and thus may not be correlated with the formation of the edge transport barrier in JET. [ABSTRACT FROM AUTHOR]

For the design and operation of large fusion devices, a detailed understanding of the power exhaust processes is necessary. This paper will give an overview of the current research on divertor power load mechanisms. The results shown are obtained in JET with the ITER-like wall (ILW)and ASDEX-Upgrade with tungsten coated plasma-facing components (PFCs). The challenges of infrared thermography on an ITER-like bulk tungsten divertor are presented. For the steady-state heat load, the power fall-off length λq in JET-ILW is compared to an empirical scaling found in JET and the ASDEX-Upgrade with carbon PFCs. A first attempt to scale the divertor broadening S in the ASDEX-Upgrade with tungsten PFCs is shown. The edge localized mode (ELM) duration tELM in JET-C and JET-ILW is compared. For similar pedestal conditions (Te,ped and ne,ped), similar ELM durations are found in JET-C and JET-ILW. For higher ne,ped at the same pedestal pressure pe,ped, longer ELM durations are found in JET-ILW. The pedestal pressure pe,ped is found to be a good qualifier for the ELM energy fluency in both JET-C and JET-ILW. Improved diagnostic capabilities reveal ELM substructures on the divertor target occurring a few milliseconds before the ELM crash. [ABSTRACT FROM AUTHOR]

The W-transport in the core plasma of JET is investigated experimentally by deriving the W-concentration profiles from the modelling of the signals of the soft x-ray cameras. For the case of pure neutral beam heating W accumulates in the core (r/a < 0.3) approaching W-concentrations of 10−3 in between the sawtooth crashes, which flatten the W-profile to a concentration of about 3 × 10−5. When central Ion cyclotron resonant heating is additionally applied the core W-concentration decays in phases that exhibit a changed mode activity, while also the electron temperature increases and the density profile becomes less peaked. The immediate correlation between the change of magnetohydrodymanic (MHD) and the removal of W from the plasma core supports the hypothesis that the change of the MHD activity is the underlying cause for the change of transport. Furthermore, a discharge from the ASDEX Upgrade is investigated. In this case the plasma profiles exhibit small changes only, while the most prominent change occurs in the W-content of the confined plasma caused by the reduction of the fuelling deuterium gas puff. Concomintantly, the W-concentration profiles in the core plasma r/a < 0.2 steepen up reminescent to the well-known connection between central radiation and transport during cases with strong, established W-accumulation, while in the present analysis such a causality between the two during the onset of W-accumulation could not be pinned down. Both case studies exemplify that small changes of the core parameters of a plasma my influence the W-transport in the plasma core drastically. [ABSTRACT FROM AUTHOR]

This paper reports the impact on confinement and power load of the high-shape 2.5 MA ELMy H-mode scenario at JET of a change from all carbon plasma-facing components to an all metal wall. In preparation to this change, systematic studies of power load reduction and impact on confinement as a result of fuelling in combination with nitrogen seeding were carried out in JET-C and are compared with their counterpart in JET with a metallic wall. An unexpected and significant change is reported on the decrease in the pedestal confinement but is partially recovered with the injection of nitrogen. [ABSTRACT FROM AUTHOR]

This paper covers aspects of long-term evolution of intrinsic impurities in the JET tokamak with respect to the newly installed ITER-like wall (ILW). At first the changes related to the change over from the JET-C to the JET-ILW with beryllium (Be) as the main wall material and tungsten (W) in the divertor are discussed. The evolution of impurity fluxes in the newly installed W divertor with respect to studying material migration is described. In addition, a statistical analysis of transient impurity events causing significant plasma contamination and radiation losses is shown. The main findings comprise a drop in carbon content (x20) (see also Brezinsek et al (2013 J. Nucl. Mater. 438 S303)), low oxygen content (x10) due to the Be first wall (Douai et al 2013 J. Nucl. Mater. 438 S1172-6) as well as the evolution of the material mix in the divertor. Initially, a short period of repetitive ohmic plasmas was carried out to study material migration (Krieger et al 2013 J. Nucl. Mater. 438 S262). After the initial 1600 plasma seconds the material surface composition is, however, still evolving. With operational time, the levels of recycled C are increasing slightly by 20% while the Be levels in the deposition-dominated inner divertor are dropping, hinting at changes in the surface layer material mix made of Be, C and W. A steady number of transient impurity events, consisting of W and constituents of inconel, is observed despite the increase in variation in machine operation and changes in magnetic configuration as well as the auxiliary power increase. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Fluctuations in the JET boundary plasma (from the scrape-off layer up to 1.5 cm inside the last closed flux surface) have been investigated using a multi-pin reciprocating Langmuir probe system. Coherent electrostatic fluctuations consistent with the geodesic acoustic mode (GAM) are observed in the edge plasma. The GAM properties and structure have been investigated and experimental evidence is presented suggesting an interaction between GAMs and the ambient turbulence. [ABSTRACT FROM AUTHOR]

• Accounting for spatial distribution of the stray radiation energy flux, including ports and cryopumps during start-up in the first plasma operation and later. • Distribution of stray radiation energy flux originating from 1 MW 60 GHz diagnostic gyrotron. • Calculation of the effective wall reflectivity for the power balance studies and radiation losses through ECE. [Display omitted] Here we re-evaluate the stray radiation loads due to electron cyclotron resonance heating and diagnostic gyrotrons in ITER, as well as compute the stray radiation energy flux due to electron cyclotron emission. Microwave loads due to stray radiation were assessed in the past. We give here more comprehensive estimates, including spatial variations in the machine. We show that for the start-up phase in the First Plasma campaign the threshold of 100 kW/m 2 , which diagnostics have to be able to withstand, is not exceeded. For the later campaigns, the stray radiation energy flux during start-up and burnthrough will exceed 450 kW/m 2. Stray radiation from electron cyclotron emission and from a diagnostic 60 GHz gyrotron can contribute to more than 140 kW/m 2 combined. [ABSTRACT FROM AUTHOR]

Microscopic turbulence properties in the edge of toroidally confined fusion plasmas are studied by comparative analysis of experimental data from seven devices, collected in an international edge turbulence database. The database contains Langmuir probe measurements of fluctuations in the floating potential and ion saturation current across the last closed flux surface. They are used to address statistical properties and particle transport. Universal features of plasma edge turbulence such as an increase in skewness across the scrape-off layer (SOL) as footprints of density blobs are recovered in all devices. Analysis of the correlation lengths and times reveals power law scaling relations with macroscopic drift-wave parameters, albeit weaker than would be expected for drift-wave turbulence. As a result, the turbulent diffusivity scales with the inverse of the magnetic field strength, which is closer to Bohm-like scaling than to gyro-Bohm scaling. Nearly identical scaling relations are determined in the confined plasma edge and the SOL, pointing to a strong connection between drift-wave turbulence in the edge and blobs in the SOL. The contributions of blobs and holes (negative density spikes) to the radial particle transport are analyzed qualitatively with a conditional averaging approach. Blobs are connected to outward transport in the SOL of all devices whereas holes exhibit no uniform propagation pattern. [ABSTRACT FROM AUTHOR]

The JET scrape-off layer (SOL) has been characterized with a reciprocating probe in inner wall (IW), and outer wall (OW), limited plasmas. Experiments revealed that SOL profiles are substantially broader (by a factor of ∼5–7.5 in the power e-folding length) for IW limited than in OW limited plasmas. Results are consistent with the larger radial turbulent transport found for IW limited plasmas. Major differences are observed between IW and OW limited plasmas on the density and electron temperature e-folding lengths, parallel flow, radial turbulent transport as well as on the temporal and spatial characteristics of the fluctuations. Experimental findings on JET suggest that the differences in the SOL characteristics for both configurations are due to a combination of a poloidal asymmetry in radial transport with a reduced cross-field transport across the last closed flux surface associated with the confinement improvement observed for OW limited plasmas.The dependence of the SOL power e-folding length on the main plasma parameters was also investigated for IW limited plasmas and a modest negative dependence on both the plasma current and the line-averaged density found. Finally, it is shown that the SOL radial transport and the amplitude of the fluctuations increase with plasma current and decrease with line-averaged density for IW limited plasmas. [ABSTRACT FROM AUTHOR]

Abstract: The initial conditioning cycle of JET ILW is analysed and compared with restart and operation in 2008 with a carbon dominated wall. Comparable water and oxygen decay times are observed during bake-out in both cases. Despite a 2×10−3 mbarl/s leak rate during plasma operation, no further wall conditioning has been necessary after plasma restart in ILW, which dramatically contrasts with 2008. Plasma O content is lower with the ILW. Higher O levels are measured after nights or week-ends, BeO layers being formed and re-eroded, but do not impact plasma operation and performance. First results on isotopic wall changeover by GDC on the ILW six months of the first D2 campaign evidence a reservoir of about 3×1022 atoms, i.e. ten time lower than in carbon PFCs. A study in JET of the glow discharge current distribution for different ratios of the ionization mean free paths to the vessel dimensions seems to indicate sufficient toroidal and poloidal homogeneity in ITER. [Copyright &y& Elsevier]

Abstract: The EDGE2D-EIRENE code is applied for simulation of divertor detachment during density ramp experiments in high triangularity, L-mode plasmas in both the carbon and Be/W environments in JET. Emphasis is placed on matching experimental data (upstream and in the divertor) as far as possible. The code runs without drifts and includes either C or Be as impurity, but not W, assuming that the divertor plasma is always cold enough for the W target source to be negligible and that the W targets have to some extent been coated with Be via main chamber migration. The simulations reproduce the observed particle flux detachment as density is raised in both C and Be/W, but not the experimental in/out asymmetry. The main difference between detachment in carbon and Be/W environments is a higher upstream density required in the Be/W case to obtain similar divertor conditions to those when carbon dominates. [Copyright &y& Elsevier]

Abstract: We model the transport of a beryllium (10Be) marker during a sequence of an inner-wall limited and a diverted Ohmic plasma phase in JET with the objective of identifying principal migration pathways. The 3D orbit-following code ASCOT is used for predictive analysis of an experiment during the 2011–2012 campaign on JET where three central pieces of a wall tile enriched with 10Be were installed to an inner wall guard limiter (IWGL) of the tokamak. Assuming erosion during the inner-wall limited plasma, the simulations indicate that 10Be is deposited along the IWGLs during the limiter phase which, when assuming further erosion, can lead to high deposition on the inner (high-field side) divertor during the diverted phase. In contrast, beryllium confined in the core plasma during the limiter phase is seen to be predominantly uniformly deposited during the diverted phase on the outer (low-field side) wall limiters and divertor tiles. [Copyright &y& Elsevier]

Abstract: For the ITER-like wall, the JET mirror link divertor spectroscopy system was redesigned to fully cover the tungsten horizontal strike plate with faster time resolution and improved near-UV performance. Since the ITER-like wall project involves a change in JET from a carbon dominated machine to a beryllium and tungsten machine with residual carbon, the aim of the system is to provide the recycling flux, equivalent to the impinging deuterium ion flux, the impurity fluxes (C, Be, O) and tungsten sputtering fluxes and hence give information on the tungsten divertor source. In order to do this self-consistently, the system provides plasma characterization through the deuterium Balmer spectra measurements of electron density and temperature during high density. L-mode results at the density limit from Stark broadening/line ratio analysis will be presented and compared to Langmuir probe profiles and 2D-tomography of low-n Balmer emission [1]. Comparison with other diagnostics will be vital for modeling attempts with the EDGE2D-EIRENE code [2] as the best possible data sets need to be provided to study detachment. [Copyright &y& Elsevier]

Abstract: The 2D edge plasma transport code EDGE2D-EIRENE has been upgraded to account for the actual material and geometric properties of the newly installed ITER like wall (ILW) at JET. This includes the simulation of beryllium and tungsten impurities as well as a revised treatment of sputtering by main plasma and impurity atoms and ions (including self sputtering). In this work, two L-mode density regimes, a sheath limited and a high recycling regime, are presented with a power scan from 2 to 6MW. Tungsten is self consistently simulated with the scrape of layer (SOL) plasma by the EDGE2D–EIRENE code. A detailed analysis of the tungsten sputtering is presented, resolving the individual contributions of the different atomic and ion species in the simulations. [Copyright &y& Elsevier]

Abstract: The presently favoured option for reactor power handling combines metallic plasma-facing components and impurity seeding to achieve highly radiative scrape-off layer and divertor plasmas. It is uncertain whether tolerable divertor power loads will be obtained in this scenario, necessitating the development of predictive modelling tools. L-mode experiments with N2 seeding have been conducted at both ASDEX Upgrade and JET for benchmarking the critically important impurity radiation models in edge fluid codes. In both machines, particle and power loads are observed to first reduce at the inner target, and only then at the outer target. The outer divertor cools down with increasing N seeding rate, evolving from low-recycling conditions to a regime with peak temperature of 8–10eV in both devices. First SOLPS5.0 simulations of N2 seeding in ASDEX Upgrade geometry show a similar in–out asymmetry in the effect of impurity radiation when drifts are activated in the simulations. [Copyright &y& Elsevier]

Abstract: Material migration and the resulting evolution of plasma facing surfaces were studied at the beginning of the JET ILW campaign using the singular opportunity of well-defined initial conditions with virgin Be and W wall components. In a sequence of identical Ohmically heated discharges the evolution of wall material sources as well as that of residual impurity sources were studied by spectroscopic detection of suitable emission lines of corresponding neutral atom and singly charged ion species in the visible spectral range. The evolution of divertor surface composition resulting from wall material migration occurred at a similar time scale as previously observed in Be migration experiments in the JET carbon wall configuration. In contrast to these experiments with initial Be evaporation on the carbon main chamber wall, the JET ILW migration experiment is characterised by a continuous Be wall source because the main chamber wall now consists of bulk Be components. The experiment further reveals unexpectedly high Be deposition at W divertor surfaces already during preceding limiter discharges for system commissioning, which has implications for predictive modelling of the expected fuel retention in ITER. [Copyright &y& Elsevier]