Plasma Core Research Articles

Assessment of the ITER divertor bolometer diagnostic performance

The ITER bolometer diagnostic will provide measurements of the total radiation emitted from the plasma, a part of the overall energy balance. It will consist of some 550 lines of sight (LOS) of bolometer detector, bundled in multiple individual cameras, which will be located in the gaps between blanket modules on the vacuum vessel wall, in five divertor cassettes, in two upper port plugs and in one equatorial port plug (Meister et al., 2017). The LOS are optimised as much as possible with design constraints to capture the details of different plasma regions in which the intensity and local distribution of radiation will vary over a large range. This is especially true in the divertor, where up to 70 % of the total thermal exhaust power from the core will have to be radiated during burning plasma operation. This radiation will be tightly concentrated meaning that not only will tomographic reconstruction of the distribution be challenging, but the bolometer cameras themselves need to be properly designed to handle the resulting heat fluxes. This paper first presents an assessment of the photonic heat fluxes falling on bolometers in the divertor region during high performance operation (where the heat fluxes will be highest). These heat fluxes are computed using ray-tracing from radiation distributions obtained with the SOLPS (plasma boundary region) and JINTRAC (plasma core region) codes and will serve as essential input to the thermal design studies of bolometers. The same SOLPS and JINTRAC simulations are used as models for the second focus of the paper which examines how closely the radiation distribution obtained from tomographic inversion of synthetic signals from the bolometer LOS matches the model input. Some effort has also been devoted to a study of how the reconstructions are affected by loss of individual LOS.

Design of line array detection scheme for Polarimeter-Interferometer system on EAST

A multi-channel POlarimeter-INTerferometer (POINT) system has been constructed in the core region of Experimental Advanced Superconducting Tokamak (EAST) plasma for electron density profile and current density profile measurements, with a spatial resolution of 8.5 cm. To explore the small spatial scale density fluctuation phenomena in the core plasma induced by MHD instabilities effects, a higher spatial resolution and higher responsivity detection is required.A novel line array detection scheme of POINT is proposed. By replacing the single Schottky mixer with a highly integrated line array HEMT detector at the central channel position, the existing single channel can be extended to 5 channels for spatial continuous measurement in the core plasma, which has a significantly simplified structure design compared with the discrete upgrade scheme. Related optical design, component selection, and bench tests have been carried out to validate the feasibility and reliability of the novel scheme.The IF signals are detected by an AlGaN/GaN line array high electron mobility transistors (HEMT) detector based on five silicon lenses integrated. The average responsivity of the line array detector elements is 40 kV/W. Compared with the existing planar diode Schottky mixer with a sensitivity of 750 V/W, the line array detector can also meet the essential requirements of coherent signal measurement. Benefiting from the highly integrated array detector, the spatial resolution will be improved to 4.5 mm in the novel scheme. Additionally, a circular Plano cylindrical lens is designed as a beam-shaping component to compress the Gaussian beam into a stripe. Bench test results displayed the excellent performance of the cylindrical lens, which is consistent with the simulation results. Furthermore, normalized responsivity coefficients of array elements are determined by test results.

Design study of the collection optics for Polarimetric Thomson Scattering on ITER

The ITER Core Plasma Thomson Scattering (CPTS) system is the primary diagnostic for the electron temperature (Te ) and electron density (ne ) profiles. For high performance ITER operating scenarios in which core Te > 20 keV, the conventional spectral CPTS will not provide measurement accuracy sufficient for advanced machine control. Addition of polarimetric measurement capability to the CPTS diagnostic is projected to increase measurement accuracy in high temperature/low density plasmas to required levels. To determine the feasibility of Polarimetric TS implementation on ITER, the polarization characteristics of the CPTS scattered light collection system have been modeled as part of a conceptual design study. When linearly polarized light is propagated through the CPTS collections optics, polarization is rotated and ellipticized. This characteristic of the ITER collection optical system complicates separation and analysis of the depolarized component of the Thomson scattered light that is produced at high Te . To address this challenge, a custom waveplate and polarizing beamsplitter should be mounted before the final lens in the CPTS collection optical system. Additionally, polarization performance should be one of the criteria by which materials and coatings are selected for the CPTS collection optical system.

Experimental determination of the three components of toroidal momentum transport in the core of a tokamak plasma

A new approach to infer the momentum transport in tokamak core plasmas via perturbation experiments is presented. For the first time, the analysis self-consistently includes all momentum transport components and their time dependencies, which are essential to separate the momentum fluxes and closely match the experiment. The quantitative agreement between the experimentally inferred transport coefficients and the gyrokinetic predictions provides an unprecedented validation. This work shows that the new methodology and gyrokinetic predictions can now be utilized on the route to physics-based prediction of momentum transport in future reactor plasmas.

Analysis and Application of the Electromagnetic Wave Propagation, Fields, and Dispersion Relation in an Elliptical Chiro Plasma Waveguide Immersed in Magnetized Plasma

AbstractThis research proposes and investigates an elliptical waveguide filled with chiro plasma as its core and magnetized plasma as its clad region. The clad region is in the constant infinite magnetic field, and its direction is the same as the wave propagation in the chiro plasma elliptical waveguide. The electromagnetic wave propagation in the considered waveguide is studied. The expressions for electromagnetic field components in the chiro plasma core and magnetized plasma cladding are derived. The dispersion relations for the hybrid modes are calculated considering appropriate boundary conditions at the chiro plasma and magnetized plasma interface. The behavior of the energy flux and the dispersion curves are numerically and graphically studied. It is seen that the magnitude of the power density decreases with the increase of the cyclotron frequency and it increases with the increase of the plasma frequency and supports power flow in the backward direction for the considered mode. Furthermore, it is shown higher values of the chirality parameter cause the magnitude of the power flux to increase in the backward direction.

Minimization of the edge modes and near fields of a travelling wave array antenna for WEST

The travelling wave array (TWA) antenna has been proposed as a promising alternative for ion cyclotron resonant heating in future fusion reactors. In this study, the possibility to make a TWA compatible with a tungsten environment like the WEST tokamak is assessed. For this purpose, two aspects of the antenna are investigated: the power spectrum and the near fields excited by the antenna. The sensitivity of these parameters to load and capacitor layout variations is taken into account while satisfying a proper antenna frequency response. The sensitivity of the power spectrum to frequency variation is also investigated to allow the possibility of fast feedback of the power deposition into the core plasma of WEST. The high resilience of the TWA to these variations is demonstrated and the main parameters of the TWA expected in WEST are compared to a current WEST antenna for the same loading. Possible optimizations of the TWA antenna are discussed. The present study is fully transferable to a fusion reactor like DEMO or ARC.

Exploration of non-resonant divertor features on the Compact Toroidal Hybrid

Non-resonant divertors (NRDs) separate the confined plasma from the surrounding plasma facing components (PFCs). The resulting striking field line intersection pattern on these PFCs is insensitive to plasma equilibrium effects. However, a complex scrape-off layer (SOL), created by chaotic magnetic topology in the plasma edge, connects the core plasma to the PFCs through varying magnetic flux tubes. The Compact Toroidal Hybrid (CTH) serves as a test-bed to study this by scanning across its inductive current. Simulations observe a significant change of the chaotic edge structure and an effective distance between the confined plasma and the instrumented wall targets. The intersection pattern is observed to be a narrow helical band, which we claim is a resilient strike line pattern. However, signatures of finger-like structures, defined as heteroclinic tangles in chaotic domains, within the plasma edge connect the island chains to this resilient pattern. The dominant connection length field lines intersecting the targets are observed via heat flux modeling with EMC3-EIRENE. At low inductive current levels, the excursion of the field lines resembles a limited plasma wall scenario. At high currents, a private flux region is created in the area where the helical strike line pattern splits into two bands. These bands are divertor legs with distinct SOL parallel particle flow channels. The results demonstrate the NRD strike line pattern resiliency within CTH, but also show the underlying chaotic edge structure determining if the configuration is diverted or limited. This work supports future design efforts for a mechanical structure for the NRD.

Simulation study of the influence of E× B drift on tungsten impurity transport in the scrape-off layer

Tungsten (W) is used as the plasma-facing material in the divertor region of future fusion reactors, such as ITER; however, its concentration in the core plasma must be maintained at an extremely low level. W transport in the scrape-off layer (SOL), which is related to the source of core W contamination, has been extensively studied. In this study, the influence of E× B drift on the transport of W impurity in the SOL is studied via numerical simulations of a model case based on EAST upper single-null configuration with high recycling divertor plasma. W transport is simulated using DIVIMP on the background plasma obtained from scape-off layer plasma simulation-ITER simulation including drifts. The E× B drift of W ions is introduced based on the background electric field. Therefore, both the direct E× B drift effect of W ion and the indirect effect via background plasma on W transport in the SOL are studied. The influence on the flux of W impurities entering confined plasma across the last closed flux surface Γ enter is focused on, which is expected to be proportional to the core W concentration. Results reveal that Γ enter is mainly from the outer (inner) target under a favorable (unfavorable) toroidal field B T and can be increased by more than one order of magnitude compared with the case without drifts; this reflects the significant effect of E× B drift. The effects due to the background plasma and the poloidal and radial E× B drift of W ion, as well as the related mechanisms, are analyzed in detail for three stages of W transport in the SOL: effective sputtering from the target, leakage from the divertor, and entry into the confined plasma.

The behavior and control of injected impurity in the closed divertor of HL-2A

Excessive heat load on the divertor target has been identified as a major challenge for present and future tokamaks. Divertor detachment achieved by injecting radiating impurity is an effective method to reduce heat load onto the divertor target surface. However, extremely serious effects on the core plasma could be given by the injected extrinsic impurity, such as fuel dilution and cooling due to energy radiation loss by the impurity in the core plasma region. Therefore, understanding the impurity behavior and then controlling the impurity content during divertor impurity injection are important issues of a tokamak. The closed divertor has the advantage of realization of divertor detachment and the Huan Liuqi-2A (HL-2A) tokamak has a very closed symmetrical divertor structure. In this work, experiments and SOLPS-ITER simulation gave the picture of the impurity behavior and showed that the friction force can play a more key role in screening and controlling radiated impurity, comparing with pressure/temperature gradient force during detachment in HL-2A with the closed divertor. Increasing the degree of divertor detachment (DoD), the screening ability of the divertor is strengthened, which is conducive to the control of impurity ions. It implies that the injected impurity can be confined in the closed divertor under detachment and, to some extent, the effect of DoD or impurity gas flux on main plasma can be attenuated for HL-2A with extrinsic impurity gas injection. During divertor detachment, the screening effect of N+ and Ar+ ions is stronger than that of Ne+ ions. As a result, the behavior and control of impurity with impurity injection in the closed divertor of HL-2A are presented with experimental and simulated results, which gives meaningful understanding and suggestion for heat load mitigation and controlling the effect of impurity in HL-2A and other tokamaks.

Effect of coherent edge-localized mode on transition to high-performance hybrid scenarios in KSTAR

This paper deals with one of the origins and trigger mechanisms responsible for the observed performance enhancements in the hybrid scenario experiments conducted in Korea Superconducting Tokamak Advanced Research (KSTAR). The major contribution to the performance improvement comes from a broader and higher pedestal formation. The increase of fast ion pressure due to a plasma density decrease also contributes substantially to the global beta. Although the reduced core plasma volume resulting from the pedestal expansion has a negative effect on the core thermal energy, a considerable confinement improvement observed in the inner core region limits the degradation. The one significant characteristic of high-performance discharges is the presence of Coherent Edge-localized Mode (CEM) activity. CEM is triggered during the pedestal recovery phase between typical ELM crashes and has been found to be related to the increase of particle and heat transport. It appears to underlie two commonly observed phenomena in high-performance hybrid scenario discharges in KSTAR; pedestal broadening and continuous density decrease. Despite the associated transport increase, CEM activities can induce performance enhancement. With the pedestal broadening, ELM crashes become delayed and weakened, which, in turn, allows for a higher pedestal. Moreover, the density decrease directly increases fast ion pressure by extending the beam-slowing-down time. The linear gyrokinetic analysis reveals that the increase of fast ions could initiate positive feedback loops, leading to the stabilization of Ion Temperature Gradient mode in the inner core region.

Experiments in high-performance JET plasmas in preparation of second harmonic ICRF heating of tritium in ITER

The reference ion cyclotron resonance frequency (ICRF) heating schemes for ITER deuterium–tritium (D-T) plasmas at the full magnetic field of 5.3 T are second harmonic heating of T and 3He minority heating. The wave-particle resonance location for these schemes coincide and are central at a wave frequency of 53 MHz at 5.3 T. Experiments have been carried out in the second major D-T campaign (DTE2) at JET, and in its prior D campaigns, to integrate these ICRF scenarios in JET high-performance plasmas and to compare their performance with the commonly used hydrogen (H) minority heating. In 50:50 D:T plasmas, up to 35% and 5% larger fusion power and diamagnetic energy content, respectively, were obtained with second harmonic heating of T as compared to H minority heating at comparable total input powers and gas injection rates. The core ion temperature was up to 30% and 20% higher with second harmonic T and 3He minority heating, respectively, with respect to H minority heating. These are favourable results for the use of these scenarios in ITER and future fusion reactors. According to modelling, adding ICRF heating to neutral beam injection using D and T beams resulted in a 10%–20% increase of on-axis bulk ion heating in the D-T plasmas due to its localisation in the plasma core. Central power deposition was confirmed with the break-in-slope and fast Fourier transform analysis of ion and electron temperature in response to ICRF modulation. The tail temperature of fast ICRF-accelerated tritons, their enhancement of the fusion yield and time behaviour as measured by an upgraded magnetic proton recoil spectrometer and neutral particle analyser were found in agreement with theoretical predictions. No losses of ICRF-accelerated ions were observed by fast ion detectors, which was as expected given the high plasma density of n e ≈ 7–8 × 1019 m−3 in the main heating phase that limited the formation of ICRF-accelerated fast ion tails. 3He was introduced in the machine by 3He gas injection, and the 3He concentration was measured by a high-resolution optical penning gauge in the sub-divertor region. The DTE2 experiments with 3He minority heating were carried with a low 3He concentration in the range of 2%–4% given the fact that the highest neutron rates with 3He minority heating in D plasmas were obtained at low 3He concentrations of ∼2%, which also coincided with the highest plasma diamagnetic energy content. In addition to 3He introduced by 3He gas injection, an intrinsic concentration of 3He of the order of 0.2%–0.4% was measured in D-T plasmas before 3He was introduced in the device, which is attributed to the radioactive decay of tritium to 3He. According to modelling, even such low intrinsic concentrations of 3He lead to significant changes in ICRF power partitioning during second harmonic heating of T due to absorption of up to 30% of the wave power by 3He.

Tritium neutral beam injection on JET: calibration and plasma measurements of stored energy

Neutral beam injection (NBI) is a flexible auxiliary heating method for tokamak plasmas, capable of being efficiently coupled to the various plasma configurations required in the Tritium and mixed deuterium-tritium experimental campaign on the Joint European Torus (JET) device. High NBI power was required for high fusion yield and alpha particle studies and to provide mixed deuterium-tritium (D-T) fuelling in the plasma core, it was necessary to operate the JET NBI systems in both deuterium and tritium. Further, the pure tritium experiments performed required T NBI for high isotopic purity and reduced 14 MeV neutron yield. Accurate power calibrations are also essential to machine safety. Previously on JET there have been a number of questions raised on the NBI power calibration, in particular following the Trace Tritium Experiments (TTEs). Operator activities on the tokamak NBI system, including calibrations, were performed in 2020. Following these activities, a series of plasma experiments were devised to further corroborate the T NBI power by comparing the plasma response to the D NBI power. A series of stationary, L-mode plasmas were performed on JET with different beam combinations used in different phases of the same pulse. By comparing the plasma response for D and T NBI it was possible to corroborate the T NBI power calibration using the D NBI power calibration. The stored energy as measured by magnetic diagnostics, corrected for fast particle stored energy, show that the uncertainty in NBI power calibration in T is comparable to that in D.

The latest developments of microwave diagnostics for high temperature plasma in ELVA-1 company

For nearly 30 years, we have been designing and supplying instruments for microwave diagnostics of high temperature plasma. This report provides a description of the mm-wave components we utilize to make diagnostics within the frequency range of 26–330 GHz. While most of these components are standard and readily available on the market, we have also developed a few specific devices that simplify the architecture of our instruments. The article includes descriptions of these devices: Backward Wave Oscillators (BWO), Impact Ionization Avalanche Transit-Time diode (IMPATT) sources, IMPATT Active Frequency Multipliers (AFM), Noise Sources, and Electronically Controlled Attenuators. Furthermore, we offer an overview of the microwave plasma diagnostics we have supplied, including ECE radiometers operating at 50–220 GHz, as well as heterodyne interferometers operating at fixed frequency 94 GHz, 140 GHz, or 300 GHz. We also discuss methods employed to ensure measurement stability and present the achieved results. The advent of the new era of modern Monolithic Microwave Integrated Circuit (MMIC) based devices has brought forth exciting possibilities. As an example, we discuss the upgrade of the low noise receiver for the Collective Thomson Scattering (CTS) diagnostic at Wendelstein 7-X, which enables ion temperature measurements in the plasma core [1]. Lastly, we provide a list of MMIC-based devices that are currently available and have garnered the attention of the plasma diagnostics community.

Tangential extreme ultraviolet and soft x-ray diagnostic system for time-resolved temperature measurement on the High Beta Tokamak-Extended Pulse.

The High Beta Tokamak-Extended Pulse has recently incorporated a tangential multi-energy extreme ultraviolet and soft x-ray diagnostic system. This system enables measurements of the electron temperature and the examination of mode dynamics within the tokamak. While other systems have been built for poloidal views over similar temperature ranges, this is the first multi-energy tangential-view system designed to work in a temperature range below 200eV in a tokamak. To facilitate these measurements, a filter wheel comprising five distinct groups of dual-filters has been developed and implemented. By employing a combination of 0.1μm aluminum and 0.2μm titanium filters, the system allows estimation of electron temperature profiles through reconstruction of the emission profile using the standard "double-foil" technique. The influence of impurities and filter oxide layers on measurement outcomes is examined. Results reveal that, while the absolute electron temperature values may exhibit some deviations, key characteristics like the electron temperature profile shape and inversion radius during sawtooth events remain consistent. This consistency confirms the system's suitability for core plasma studies. This system has proven effective in detecting and analyzing internal magnetohydrodynamic phenomena, such as sawteeth.

Development of dual-path multi-pass Thomson scattering system in GAMMA 10/PDX

A dual-path Thomson scattering (DPTS) system was developed to measure electron temperature and density in a GAMMA 10/PDX central cell (CC) and an end cell (EC). The DPTS comprises CC Thomson scattering system (CC-TS) and an EC Thomson scattering system (EC-TS). In the CC-TS, we developed a multipass system to improve the signal intensities and time resolution. In the EC-TS, a double-pass system was developed as a preliminary step in developing a multipass system. In a previous EC-TS, the optical components for the double-pass configuration were placed in a vacuum vessel. To reduce the influence of stray light and facilitate alignment, a hole for the double-pass laser path was made on the partition wall in the vacuum chamber such that the probe laser could be extracted into the atmosphere and the double-path configuration could be easily adjusted. We applied DPTS to detached plasma experiments and successfully measured the core and edge electron temperatures and densities with the detached plasma experiments with the additional application of supersonic molecular beam injection and central electron cyclotron heating in the core plasma to simultaneously introduce intermittent particle flux.

Performance of a liquid Sn divertor target during ASDEX upgrade L-mode and H-mode operation

One of the ways to extend the lifetime of the divertor for DEMO could be to replace the solid tungsten plasma-facing components with liquid tin (Sn) confined in a tungsten capillary porous structure (CPS). Testing a CPS in a divertor plasma configuration is crucial for the development of a liquid metal divertor (LMD) to understand how the main plasma is affected. Only a limited Sn concentration is allowed in the plasma core, due to the high radiative losses associated with the high atomic number of Sn (50). Therefore, it is necessary to test a small-scale LMD filled with Sn in a tokamak environment, which has not previously been done. In ASDEX Upgrade, a liquid tin module (LTM) has been exposed by means of the divertor manipulator. During plasma flat-top, the outer strike point (OSP) was placed onto the pre-heated LTM and held there for a time interval between 2 and 3.4s over multiple discharges. Photographs of the LTM taken after each discharge, revealed macroscopic Sn leakage onto the adjacent tile. Simulations with the HeatLMD code predicted an acceptable tin erosion near the LTM with thermal sputtering dominating over evaporation. However, spectroscopic measurements revealed an order of magnitude higher erosion. Since this remained constant when the OSP was held on the LTM so that the surface temperature increased, evaporation could be excluded as the main source of Sn erosion. Comparison between discharges with different durations of OSP location on the LTM revealed an increase in core radiation up to 1.5MW due to Sn. The 1.5D-impurity transport code STRAHL was used to interpret this increase in total plasma radiation and revealed a Sn concentration in the main plasma of up to 1.4×10-4. Given that the LTM only covered about 1/650 of the outer divertor circumference, extrapolating to a full toroidal divertor implies erosion is above acceptable limits. The unexpectedly high Sn fraction in the main plasma is attributed to the ejection of Sn droplets reaching the main plasma, which may have originated from either the CPS or leaked tin. This conclusion is also supported by splashes of tin droplets, which were observed on the adjacent divertor tile and one ∼0.5m downstream. Therefore, to make a Sn-filled LMD a viable alternative to solid tungsten, the formation of droplets must be reduced by two orders of magnitude.

Role of pinch in Argon impurity transport in ohmic discharges of Aditya-U Tokamak

We present experimental results of the trace argon impurity puffing in the ohmic plasmas of Aditya-U tokamak performed to study the argon transport behaviour. Argon line emissions in visible and Vacuum Ultra Violet (VUV) spectral ranges arising from the plasma edge and core respectively are measured simultaneously. During the experiments, space resolved brightness profile of Ar1+ line emissions at 472.69 nm (3p44s 2P3/2–3p44p 2D3/2), 473.59 nm (3p44s 4P5/2–3p44p 4P3/2), 476.49 nm (3p44s 2P1/2–3p44p 2P3/2), 480.60 nm (3p44s 4P5/2–3p44p 4P5/2) are recorded using a high resolution visible spectrometer. Also, a VUV spectrometer has been used to simultaneously observe Ar13+ line emission at 18.79 nm (2s22p 2P3/2–2s2p22P3/2) and Ar14+ line emission at 22.11 nm (2s21S0–2s2p 1P1). The diffusivity and convective velocity of Ar are obtained by comparing the measured radial emissivity profile of Ar1+ emission and the line intensity ratio of Ar13+ and Ar14+ ions, with those simulated using the impurity transport code, STRAHL. Argon diffusivities ~ 12 m2/s and ~ 0.3 m2/s have been observed in the edge (ρ > 0.85) and core region of the Aditya-U, respectively. The diffusivity values both in the edge and core region are found to be higher than the neo-classical values suggesting that the argon impurity transport is mainly anomalous in the Aditya-U tokamak. Also, an inward pinch of ~ 10 m/s mainly driven by Ware pinch is required to match the measured and simulated data. The measured peaked profile of Ar density suggests impurity accumulation in these discharges.

Tritium burn efficiency in deuterium–tritium magnetic fusion

The controlling parameters regarding tritium burn efficiency (TBE) are derived from first principles and shown to depend fundamentally on the permitted He gas fraction in the divertor and effective pumping speeds of He ash and unburned hydrogenic fuel. The analysis is generic to any equilibrated magnet fusion plasma using a divertor for particle exhaust. The He gas fraction in the plasma limits the maximum TBE due to the link between ash dilution effects in the core plasma and fusion performance. High TBE in magnetic fusion devices is counter-correlated to achieving high gain and power density for commercial fusion. The impact of TBE on fusion performance for several figures of merit are derived, including power density, required n−τe product, and plasma energy gain Qp . The TBE formulation presented here is applied to existing devices, based on published data of enrichment and τHe∗ from research tokamaks. This assessment strongly motivates exploration of technologies that would enhance the effective pumping speed of He to fuel out of the plasma.

Comparison of momentum transport in matched hydrogen and deuterium H-mode plasmas in ASDEX Upgrade

Advanced momentum transport analysis is used to study matched hydrogen (H) and deuterium (D) plasmas in the core of ASDEX Upgrade. The aim is to validate gyrokinetic theory and assess a possible isotope dependence. The methodology extracts momentum diffusion, convection, and intrinsic torque as a function of time from experiments employing neutral beam injection (NBI) modulation. H and D plasma scenarios with comparable ion heat fluxes, NBI torque, electron densities, and several dimensionless parameters were designed to highlight any mass dependency. Linear gyrokinetic simulations predict that, for similar background gradients, the Prandtl and pinch numbers should be similar for H and D. This was confirmed by the experimental momentum transport analyses. The assessed intrinsic torques were found to be similar between H and D, co-current directed and located near the outermost region of the plasma core. The strength of the intrinsic torque is correlated with the amplitude of the plasma pressure gradient in the pedestal. Finally, a robust error analysis demonstrates the uniqueness of the parameters obtained together with their uncertainties. Neglecting the intrinsic torque, or its time dependence, systematically distorts the assessed momentum diffusion and convection. This is the first method to separate all three transport mechanisms from experimental data by retaining their time dependencies, that is found to match, quantitatively, the gyrokinetic predictions for Prandtl and pinch numbers, within experimental uncertainties.

Performance of tungsten plasma facing components in the stellarator experiment W7-X: Recent results from the first OP2 campaign

The transition to reactor-relevant materials for the plasma facing components (PFCs) is an important and necessary step to provide a proof of principle that the stellarator concept can meet the requirements of a future fusion reactor by demonstrating high performance steady-state operation. As a first step to gain experience with tungsten as plasma-facing material in the Wendelstein 7-X (W7-X) stellarator, graphite tiles coated with an approximately 10 µm MedC tungsten layer (NILPRP Bucharest) were installed to complete the ECRH beam dump area in two of the five plasma vessel modules over an area of approximately one square meter each. In addition, tungsten baffle tiles are installed (40 tiles in total) with either bulk tungsten as part of NBI shine-through target or with a tungsten heavy alloy (W95-Ni3.5-Cu1.5) to replace the graphite tiles that were previously thermally overloaded. Based on an advanced diffusive field line tracing method and EMC3-Eirene simulations, the overloaded baffle tiles were redesigned by making the tiles thinner (i.e. moving the plasma-facing surface (PFS) away from the hot plasma region) and by reducing the local angle of incidence through toroidal displacement of the watershed. Significant erosion of the tungsten tiles can only be expected if sputtering by impurity ions such as carbon or oxygen ions contributes. However, the resulting central concentration of tungsten and the corresponding radiation losses are expected to be marginal. The expected deposition of carbon on the tungsten surfaces in the baffle regions mitigates further the possible tungsten enrichment in the core plasma. In OP2.1, no adverse effects of the installed tungsten PFCs on the plasma performance were observed during normal plasma operation. With the design changes made in the baffle area, the predicted heat load reductions could be experimentally confirmed.