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A poloidal array of scintillator-based Fast-Ion Loss Detectors (FILDs) has been installed in the ASDEX Upgrade (AUG) tokamak. While all AUG FILD systems are mounted on reciprocating arms driven externally by servomotors, the reciprocating system of the FILD probe located just below the midplane is based on a magnetic coil that is energized in real-time by the AUG discharge control system. This novel reciprocating system allows, for the first time, real-time control of the FILD position including infrared measurements of its probe head temperature to avoid overheating. This considerably expands the diagnostic operational window, enabling unprecedented radial measurements of fast-ion losses. Fast collimator-slit sweeping (up to 0.2 mm/ms) is used to obtain radially resolved velocity-space measurements along 8 cm within the scrape-off layer. This provides a direct evaluation of the neutral beam deposition profiles via first-orbit losses. Moreover, the light-ion beam probe (LIBP) technique is used to infer radial profiles of fast-ion orbit deflection. This radial-LIBP technique is applied to trapped orbits (exploring both the plasma core and the FILD stroke near the wall), enabling radial localization of internal plasma fluctuations (neoclassical tearing modes). This is quantitatively compared against electron cyclotron emission measurements, showing excellent agreement. For the first time, radial profiles of fast-ion losses in MHD quiescent plasmas as well as in the presence of magnetic islands and edge localized modes are presented.

Internal diamagnetic flux measurements, with measurement loops and compensation magnetic probes inside the vacuum vessel, are now available on the ASDEX Upgrade tokamak. The measured diamagnetic flux is compared to that predicted by simulations and calculated from equilibrium reconstruction. The diamagnetic flux measured at 2 positions separated toroidally by 180° in the vacuum vessel is compared.

Infrared (IR) thermography is widely used in fusion research to study power exhaust and incident heat load onto the plasma facing components. Due to the short pulse duration of today's fusion experiments, IR systems have mostly been designed for off-line data analysis. For future long pulse devices (e.g., Wendelstein 7-X, ITER), a real time evaluation of the target temperature and heat flux is mandatory. This paper shows the development of a real time capable IR system for ASDEX Upgrade. A compact IR camera has been designed incorporating the necessary magnetic and electric shielding for the detector, cooler assembly. The camera communication is based on the Camera Link industry standard. The data acquisition hardware is based on National Instruments hardware, consisting of a PXIe chassis inside and a fibre optical connected industry computer outside the torus hall. Image processing and data evaluation are performed using real time LabVIEW.

In the ASDEX Upgrade tokamak, a radiation measurement for a wide spectral range, based on semiconductor detectors, with 256 lines of sight and a time resolution of 5 μs was recently installed. In combination with the foil based bolometry, it is now possible to estimate the absolutely calibrated radiated power of the plasma on fast timescales. This work introduces this diagnostic based on AXUV (Absolute eXtended UltraViolet) n-on-p diodes made by International Radiation Detectors, Inc. The measurement and the degradation of the diodes in a tokamak environment is shown. Even though the AXUV diodes are developed to have a constant sensitivity for all photon energies (1 eV-8 keV), degradation leads to a photon energy dependence of the sensitivity. The foil bolometry, which is restricted to a time resolution of less than 1 kHz, offers a basis for a time dependent calibration of the diodes. The measurements of the quasi-calibrated diodes are compared with the foil bolometry and found to be accurate on the kHz time scale. Therefore, it is assumed, that the corrected values are also valid for the highest time resolution (200 kHz). With this improved diagnostic setup, the radiation induced by edge localized modes is analyzed on fast timescales.

One of the main approaches to thermonuclear fusion relies on confining high temperature plasmas with properly shaped magnetic fields. The determination of the magnetic topology is, therefore, essential for controlling the experiments and for achieving the required performance. In Tokamaks, the reconstruction of the fields is typically formulated as a free boundary equilibrium problem, described by the Grad-Shafranov equation in toroidal geometry and axisymmetric configurations. Unfortunately, this results in mathematically very ill posed problems and, therefore, the quality of the equilibrium reconstructions depends sensitively on the measurements used as inputs and on the imposed constraints. In this paper, it is shown how the different diagnostics (Magnetics Measurements, Polarimetry and Motional Stark Effect), together with the edge current density and plasma pressure constraints, can have a significant impact on the quality of the equilibrium on JET. Results show that both the Polarimetry and Motional Stark Effect internal diagnostics are crucial in order to obtain reasonable safety factor profiles. The impact of the edge current density constraint is significant when the plasma is in the H-mode of confinement. In this plasma scenario the strike point positions and the plasma last closed flux surface can change even by centimetres, depending on the edge constraints, with a significant impact on the remapping of the equilibrium-dependent diagnostics and of pedestal physics studies. On the other hand and quite counter intuitively, the pressure constraint can severely affect the quality of the magnetic reconstructions in the core. These trends have been verified with several JET discharges and consistent results have been found. An interpretation of these results, as interplay between degrees of freedom and available measurements, is provided. The systematic analysis described in the paper emphasizes the importance of having sufficient diagnostic inputs and of properly validating the results of the codes with independent measurements.

For the new ITER-like wall at JET, two new infrared diagnostics (KL9B, KL3B) have been installed. These diagnostics can operate between 3.5 and 5 μm and up to sampling frequencies of ∼20 kHz. KL9B and KL3B image the horizontal and vertical tiles of the divertor. The divertor tiles are tungsten coated carbon fiber composite except the central tile which is bulk tungsten and consists of lamella segments. The thermal emission between lamellae affects the surface temperature measurement and therefore KL9A has been upgraded to achieve a higher spatial resolution (by a factor of 2). A technical description of KL9A, KL9B, and KL3B and cross correlation with a near infrared camera and a two-color pyrometer is presented.

The new JET ITER-like wall (made of beryllium and tungsten) is more fragile than the former carbon fiber composite wall and requires active protection to prevent excessive heat loads on the plasma facing components (PFC). Analog CCD cameras operating in the near infrared wavelength are used to measure surface temperature of the PFCs. Region of interest (ROI) analysis is performed in real time and the maximum temperature measured in each ROI is sent to the vessel thermal map. The protection of the ITER-like wall system started in October 2011 and has already successfully led to a safe landing of the plasma when hot spots were observed on the Be main chamber PFCs. Divertor protection is more of a challenge due to dust deposits that often generate false hot spots. In this contribution we describe the camera, data capture and real time processing systems. We discuss the calibration strategy for the temperature measurements with cross validation with thermal IR cameras and bi-color pyrometers. Most importantly, we demonstrate that a protection system based on CCD cameras can work and show examples of hot spot detections that stop the plasma pulse. The limits of such a design and the associated constraints on the operations are also presented.

The lithium beam diagnostic at ASDEX Upgrade routinely delivers electron density profiles in the plasma edge by lithium beam impact excitation spectroscopy. An accurate background subtraction requires a periodically chopped lithium beam. A new, improved chopping system was developed and installed. It involves a voltage modulation for the extractor electrode and the beam deflection plates. The modulation of the extractor electrode reduces the unused portion of lithium ions and improves the stability of the beam with respect to its position. Furthermore, the data indicate an extended emitter lifetime. The extractor chopping was also found to be insensitive to sparks. The deflection chopping experiments demonstrated beam chopping in the kilohertz range. The significantly higher modulation frequency of the deflection chopping improves background subtraction of fast transient events. It allows a more accurate density measurements in the scrape off layer during impurity injections and edge localized modes.

Experimental measurements of the SOL power decay length (λ(q)) estimated from analysis of fully attached divertor heat load profiles from two tokamaks, JET and ASDEX Upgrade, are presented. Data was measured by means of infrared thermography. An empirical scaling reveals parametric dependency λ(q) in mm = 0.73B(T)(-0.78)q(cyl)(1.2)P(SOL)(0.1)R(geo)(0), where B(T)(T) describes the toroidal magnetic field, q(cyl) the cylindrical safety factor, P(SOL)(MW) the power crossing the separatrix and R(geo)(m) the major radius of the device. A comparison of these measurements to a heuristic particle drift-based model shows satisfactory agreement in both absolute magnitude and scaling. Extrapolation to ITER gives λ(q) ≃ 1 mm.

A study of the effects of drifts on the particle flow pattern and in-out divertor plasma density asymmetry for L-mode and H-mode plasmas is carried out for EAST discharges by the edge plasma transport codes SOLPS and BOUT++ . The simulation of L-mode plasmas is done by SOLPS while the simulation of H-mode plasmas is done by BOUT++ . The toroidal magnetic field direction for the simulated discharge is artificially reversed in the codes to study the effects of different drift directions on the divertor particle flow pattern and the in-out asymmetry of divertor plasma density. The divertor particle flows induced by diamagnetic and E × B drifts are found to have similar directions in the divertor region for the same discharge. The directions of the flows induced by drifts would be reversed with the reverse of toroidal magnetic field direction. The diamagnetic drift seems to have no effect on the in-out asymmetry of divertor plasma density due to its divergence-free nature. However, the E × B drift could result in a pronounced asymmetry of plasma density between the inner and outer divertor targets. The density in-out asymmetry caused by E × B drift is reversed with the reverse of E × B drift flow direction. Detailed analysis shows that the radial component of the E × B drift flow is the main cause of density asymmetry. The results from the simulation of H-mode plasmas with BOUT++ are similar to those of the L-mode plasmas with SOLPS except that the drift effects seem to be slightly larger in the H-mode plasmas compared to the L-mode plasmas., (© 2023. The Author(s).)