•Focus is on protection of leading edges between toroidally adjacent monoblocks.•Summarizes conclusions of coordinated, multi-device ITPA Divertor and SOL task.•Leading loading found to be well ...described by optical approximation.•Shaping required to prevent deep edge melting during steady state and ELMs.•Shaping implies reduced parallel heat flux to avoid material crystallization.
The key remaining physics design issue for the ITER tungsten (W) divertor is the question of monoblock (MB) front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM) or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA) and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets.
Inherited neuromuscular disorder (NMD) is a wide term covering different genetic disorders affecting muscles, nerves, and neuromuscular junctions. Genetic and clinical heterogeneity is the main ...drawback in a routine gene‐by‐gene diagnostics. We present Czech NMD patients with a genetic cause identified using targeted next‐generation sequencing (NGS) and the spectrum of these causes. Overall 167 unrelated patients presenting NMD falling into categories of muscular dystrophies, congenital muscular dystrophies, congenital myopathies, distal myopathies, and other myopathies were tested by targeted NGS of 42 known NMD‐related genes. Pathogenic or probably pathogenic sequence changes were identified in 79 patients (47.3%). In total, 37 novel and 51 known disease‐causing variants were detected in 23 genes. In addition, variants of uncertain significance were suspected in 7 cases (4.2%), and in 81 cases (48.5%) sequence changes associated with NMD were not found. Our results strongly indicate that for molecular diagnostics of heterogeneous disorders such as NMDs, targeted panel testing has a high‐clinical yield and should therefore be the preferred first‐tier approach. Further, we show that in the genetic diagnostic practice of NMDs, it is necessary to take into account different types of inheritance including the occurrence of an autosomal recessive disorder in two generations of one family.
Divertor infrared thermography on COMPASS Vondracek, P.; Gauthier, E.; Grof, M. ...
Fusion engineering and design,
September 2019, 2019-09-00, 20190901, Letnik:
146
Journal Article
Recenzirano
•A new fast divertor IR system was put into operation on the COMPASS tokamak. The system provides in-situ calibration possibility using a special heated divertor tile.•Radial profiles of the divertor ...heat flux are routinely measured for both the inner and the outer divertor target with spatial resolution ∼1 mm and frequency up to ∼60 kHz.•First experimental divertor heat flux measurements using the new system were successfully performed in both L-mode and H-mode.
A new fast divertor infra-red thermography system was put into operation on COMPASS. It provides full radial coverage of the bottom open divertor with pixel resolution ∼0.6–1.1 mm/px on the target surface and temporal resolution better than 20 μs. The system consists of fast IR camera TELOPS Fast-IR 2K placed in a magnetic shielding box, a positionable holder, a 1 m long IR endoscope consisting of 14 Ge and Si lenses securing off-axis view from an upper inner vertical port and a special graphite divertor tile optimized for IR thermography. The tile is equipped with a heating system allowing tile preheating up to 250 °C. Embedded thermoresistors and a calibration target (a deep narrow hole acting as a black body radiator) allows in-situ calibration of the system including estimation of the target surface emissivity. Furthermore, a roof-top shaped structure on top of the tile increases magnetic field incidence angles above 3 degrees.
Laboratory tests of the system performed during its commissioning are presented. The global transmission of the optical system was found to be τ ≈ 40–50%. Poor spatial resolution compared to the design value was observed. Too large surface error of individual lenses was identified as the main cause and re-manufacturing of the most critical lens was suggested.
First experimental results obtained using the IR system are presented: divertor heat flux profiles in L-mode with the heat flux decay length λqomp=2.1−3.3 mm and average H-mode heat flux profiles in an inter-ELM period and during an ELM heat flux maximum with λqomp≈0.6 mm and λqomp≈6.7 mm, respectively.
•Deuterium retention patterns and other impurities have been measured on 1/8 of the COMPASS divertor tiles with ion beam analysis in all spatial directions, including depth.•Using 2D spatial ...analysis, the deuterium retention pattern appears to be toroidally symmetric; symmetry breaks down when considering also depth information.•Deuterium retention is correlated to different light and heavy impurities, showing different correlations between elements at different depths.•The radial deuterium retention pattern is correlated to divertor probe signals in order to obtain a relationship between radial probe signal patterns and radial deuterium retention pattern.•Simple regression analysis shows that, under certain assumptions, deuterium retention can be predicted by probe signals, given that fit function parameters have been estimated by calibration measurements.•New contributions to the field.•1/8 of the COMPASS divertor has been inspected for impurity and deuterium retention.•The ion beam analysis takes all three spatial dimensions into account to reveal subtle differences in deuterium co-deposition.•Extensive statistical studies have been carried out to reveal correlations between deuterium and impurity deposition, and between deuterium deposition and divertor probe signals.•This paper outlines how an empirical formula for determining fuel retention in future fusion devices can be estimated based on fuel retention calibration measurements and basic divertor probe signals like electron temperature, ion saturation current and parallel heat flux.
Divertor tiles from the COMPASS tokamak have been examined with ion beam analysis for the determination of deuterium retention and co-deposits. A novel approach was used to draw 2D deuterium retention maps at different depths and correlated to the concentrations of other elements: boron, carbon, oxygen and metals. This approach, which employs all three spatial dimensions, revealed that the deuterium retention pattern is not toroidally symmetric at different depths while the overall deuterium pattern is toroidally symmetric. Analysis of visible-light camera records disclosed that parts of the divertor were not wetted by plasma. The radial profile of deuterium in both wetted and non-wetted (“shadowed”) divertor regions was compared to strike point positions, angle of incidence, electron temperature Te, parallel heat flux Qpar and ion saturation current Isat in order to find the influence of these parameters on deuterium retention. There are certain series of consecutive shots for which correlations between deuterium retention and separate plasma parameters (Te, Isat, Qpar) are high. Combining the three parameters in a linear model or a power law model enabled reconstruction of the radial deuterium retention pattern. After fitting the model to one third of the data points, the deuterium retention for shadowed tiles was reproduced at the other two thirds of the data points within 20% deviation. This study shows the need for detailed tile analysis in three dimensions, the great benefits of broad statistical analysis in solving challenges in fusion, and motivates the investigation of COMPASS tiles after dismantling.
•High magnetic field and high plasma density tokamak with SOL decay length close to ITER and PB/qAR ratio 70% of ITER.•Closed divertor that allows snowflake divertor configuration.•Ohmic losses in ...magnets reduced via cryogenic cooling of Cu coils to liquid nitrogen temperatures.•Operation of the first wall at 300°C.•Auxiliary plasma heating systems: 4MW NBI with variable injection geometry, 4MW ECRH using ITER-like gyrotrons.
The paper describes the conceptual design of the COMPASS Upgrade (COMPASS-U) tokamak which represents a compact, medium-size, high-magnetic-field and high-density device with a flexible set of poloidal field coils for generation of single, double null and snowflake configurations. In addition, COMPASS-U will be equipped with a closed divertor, which will be operated at high plasma and neutral density and with high optical opacity similar to future reactors. COMPASS-U is capable of addressing some of the key challenges in the field of the plasma exhaust physics, reactor-relevant edge plasma physics, advanced confinement regimes, advanced magnetic configurations and materials for next-step devices. The conceptual design of the individual COMPASS-U systems is presented with emphasis on magnets, support structure and power supply system enhancement.
The ITER first wall is designed for start-up and ramp-down in limiter configuration. The wall panels are toroidally shaped in order to spread the incident parallel power flux q|| uniformly, assuming ...a single decay length λq whose value is not known from first principles. In order to study the scaling of q|| with plasma parameters, infra-red viewing of specially-designed limiters has been used on the COMPASS tokamak in ∼100 discharges with scans in Ip, ne and for all combinations of magnetic field and Ip directions. The IR measurement clearly shows that in addition to the main SOL heat flux profile with λq>40mm, a steep gradient (λqnear=4±2mm) dominates q|| near separatrix. This appears independently of limiter shaping, insertion with respect to neighbors and incident field-line angles. Good agreement is found between the measured λqnear and the prediction of a heuristic drift-based model.
•Fast IR thermographic system was designed and partially commissioned on COMPASS.•Near SOL λq is constant during a sawtooth cycle.•Fast modulation of first wall heat loading due to runaway electrons ...was observed.
A new fast infrared camera was purchased on the COMPASS tokamak recently. It is equipped with a medium wavelength infrared (3–5mm) InSb detector and is capable of reaching framerate up to 1.9kHz in full frame acquisition mode (320×256px.) and up to 90kHz in sub-windowed mode (64×4px.).
First experimental measurements of plasma heat flux to inner and outer wall limiters of the COMPASS tokamak using the new camera are presented. Time evolution of parallel heat flux during sawtooth instability is studied showing strong modulation of the heat flux absolute value whilst keeping constant its radial decay length. Fast modulation of outboard midplane limiter heat loading due to runaway electrons is also presented.
The camera will be part of a new fast divertor thermographic system with exceptional spatial resolution (∼0.6–1.3mm/px. on the target plane, 0.04–0.14mm/px. mapped to the outer midplane) with a possibility of measurements of radial profiles on the divertor with 320×4px. with temporal resolution better than 20μs. A design of the foreseen optical divertor system is described together with a design of a special divertor graphite tile used for the IR thermography, that will allow in-situ surface emissivity calibration.
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