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Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international C oordination on I nternational C hallenges on L ong duration OP eration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed. [ABSTRACT FROM AUTHOR]

This paper presents an analysis of policy application in Poland affecting landscape and forest management changes designed for an ecological restoration. “Ecological” policy was introduced in Poland in 1991 (Bill on Forests) and 1995-1999 (Decree 11 and 11A). To analyse the effectiveness of the policy and its impact on landscape change, the study uses five selected indicators: 1) species distribution, 2) methods of final felling, 3) forest functions in terms of protection and economic role, 4) cost structure and 5) wood production. The analysis shows that some of the ecological ideas were applied years before official documents were issued: species composition change, in favour of broadleaves has been carried out since 1945; since 1975, considerable growth in protective areas has been observed, doubling from 22.5 to 47.5% by 2006; timber resources have also grown by 50% in the last 35 years. After 1991, when the new policy was introduced, change in these indicators was more rapid. More direct and visible examples of policy applications in the last 15 years are observed in 1) the reduction of clearcuts: in 2005 this was at its lowest level in comparison with the previous 15 years and amounted to 19.9%; 2) the focus on protection: the cost of forest and fire protection in 2006 was the third largest expense in the State Forests budget, special attention was paid to reconstruction in order to increase water retention, 3) the afforestation process: this has doubled in intensity in the last decade compared with the previous decade. Nevertheless, organised landscape change is also accompanied by natural calamities, of which fires and windbreaks are most severe.

To facilitate the estimation of heat loads on plasma-facing components in fusion devices in various different magnetic geometries, a heat load proxy model was developed based on anisotropic diffusion. In this work, this model is compared to the so-called field-line diffusion approach. To facilitate the evaluation of these models, a novel synthetic camera-based approach for obtaining heat load distributions from Monte Carlo samples was also developed and implemented. With the assistance of this synthetic camera, heat load predictions for the Wendelstein 7-X divertor were obtained and compared with infrared camera observations. It was found that the anisotropic diffusion-based model achieved a closer match to infrared camera observations, while still being suitable in computational effort for large magnetic configuration database scans. [ABSTRACT FROM AUTHOR]

W7-X completed its plasma operation in hydrogen with island divertor and inertially cooled test divertor unit (TDU) made of graphite. A substantial set of plasma-facing components (PFCs), including in particular marker target elements, were extracted from the W7-X vessel and analysed post-mortem. The analysis provided key information about underlying plasmaâ€"surface interactions (PSI) processes, namely erosion, transport, and deposition as well as fuel retention in the graphite components. The net carbon (C) erosion and deposition distribution on the horizontal target (HT) and vertical target (VT) plates were quantified and related to the plasma time in standard divertor configuration with edge transform Îą = 5/5, the dominant magnetic configuration of the two operational phases (OP) with TDU. The operation resulted in integrated high net C erosion rate of 2.8 mg sâˆ'1 in OP1.2B over 4809 plasma seconds. Boronisations reduced the net erosion on the HT by about a factor 5.4 with respect to OP1.2A owing to the suppression of oxygen (O). In the case of the VT, high peak net C erosion of 11 μ m at the strike line was measured during OP1.2B which converts to 2.5 nm sâˆ'1 or 1.4 mg sâˆ'1 when related to the exposed area of the target plate and the operational time in standard divertor configuration. PSI modelling with ERO2.0 and WallDYN-3D is applied in an interpretative manner and reproduces the net C erosion and deposition pattern at the target plates determined by different post-mortem analysis techniques. This includes also the 13C tracer deposition from the last experiment of OP1.2B with local 13CH4 injection through a magnetic island in one half module. The experimental findings are used to predict the C erosion, transport, and deposition in the next campaigns aiming in long-pulse operation up to 1800 s and utilising the actively cooled carbon-fibre composite (CFC) divertor currently being installed. The CFC divertor has the same geometrical design as the TDU and extrapolation depends mainly on the applied plasma boundary. Extrapolation from campaign averaged information obtained in OP1.2B reveals a net erosion of 7.6 g per 1800 s for a typical W7-X attached divertor plasma in hydrogen. [ABSTRACT FROM AUTHOR]

For the first time, 2D ion temperature values are derived from coherence imaging spectroscopy (CIS) fringe contrast measurements by taking Zeeman line broadening effects into account during the analysis procedure of a spatial-heterodyne CIS instrument. This allowed 2D images of C2+ ion temperatures (T i) across the 3D-shaped island divertor of the Wendelstein 7-X stellarator. Ion temperatures ranging from 10 to 20 eV are observed for the C2+ impurity species in the region above the divertor targets. During the transition from the attached to the detached plasma state, the C2+ radiation zone moves from close to the divertor target towards the last closed flux surface. Within this radiation zone, C2+ temperature does not decrease significantly. Experimentally, the coherence imaging measurements were cross-calibrated at one poloidal cross-section using a high resolution Echelle spectrometer, that shared its sightlines with the coherence imaging diagnostic. The spectra demonstrated that, apart from Doppler broadening, the Zeeman effect significantly contributes to the spectral line broadening and cannot be neglected when analyzing the CIS contrast data for T i extraction in the edge and scrape-off-layer of Wendelstein 7-X (W7-X), due to the relatively low temperatures (T i < 100 eV) and high magnetic fields (B ≈ 2.5 T). [ABSTRACT FROM AUTHOR]

The fundamental behavior of the W7-X island divertor under detached conditions, which has been theoretically predicted with the EMC3-Eirene code, is re-examined here under the experimental conditions achieved so far and compared with the first experimental results. Both simulations and experiments cover a range of divertor configurations and plasma parameters, and show the following common trends: (1) with rising impurity radiation, the target heat load decreases 'uniformly' over the entire target surface in the sense that both the peak and average heat loads can drop by an order of magnitude. Impurity radiation (mainly from intrinsic carbon) occurs primarily at the plasma edge and the resulting negative impact on the stored energy is less than 10%. (2) When the total radiation exceeds a critical level, the target particle flux (the recycling flux Γrecy) begins to fall and can drop by a factor of 3–5 at high radiation levels without an obvious indication of significant volume recombination. (3) While Γrecy decreases, the divertor neutral pressure continues to build up and reaches a maximum, at which point Γrecy has declined significantly. (4) During detachment, the electron temperature at the last closed flux surface falls in a way that is not quantitatively understandable from parallel classical heat conduction processes. This paper presents a physical explanation of the numerical/experimental results described above. Furthermore, using the EMC3-Eirene code as a diagnostic tool, we are able, apparently for the first time, to provide a full quantitative analysis of each transport channel in the island divertor, aiming to clarify how the island divertor plasma self-regulates to maintain particle, energy, and momentum balance under detached conditions. [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 the Wendelstein 7-X stellarator, the main locations of particle sources are expected to be the carbon divertors, baffles and graphite heat shield first wall. In this paper, the heat shield is implemented in EMC3-EIRENE to understand the expected areas and magnitudes of the recycling flux to this component. It is found that in the simulation the heat shield is not a significant source of recycling neutrals. The areas of simulated recycling flux are shown to correlate well with footprints of plasma-wetting seen in post-experimental campaign in-vessel inspection photos. EMC3-EIRENE reconstruction of line-integrated H-alpha measurements at the heat shield indicate that the majority of emission does not come from local recycling neutrals. Rather, the H-alpha signals at the heat shield are dominated by ionization of neutrals which have leaked from the divertor/baffle region into the midplane. The magnitude of the H-alpha line emission from the synthetic reconstruction is consistent with the experiment, indicating that a large overestimation of heat shield recycling would occur if these measurements were assumed to be from local recycling sources. In the future, it may be possible to obtain some information of local recycling from the heat shield since it was found that the majority of the recycling flux occurs on two well-localized areas. [ABSTRACT FROM AUTHOR]

The investigation of fuel retention in fusion experiments is important in view of plasma density control as well as tritium inventory for future fusion reactors. We present a first gas balance of the stellarator Wendelstein 7-X with its inertially cooled graphite divertor. The gas balance is used to estimate the wall inventory and it is found that the wall plays an important and dynamic role, absorbing or releasing particles depending on the plasma conditions. Several different scenarios are presented and the effect of fueling and heating on the wall inventory is assessed. We find that the record duration plasma experiment of 100 s required previous shorter plasmas to be successfully conducted. [ABSTRACT FROM AUTHOR]

The island divertor concept is an innovative and promising idea to handle heat and particle exhaust in stellarators. At the Wendelstein 7-X (W7-X) stellarator, this divertor concept plays a central role in the device mission to demonstrate reactor relevant plasma confinement for steady-state time scales of up to 30 minutes in the high-performance campaign (OP2) starting in 2022. During the recently concluded first campaign with the inertially cooled island divertor, a large step in the experimental qualification of this divertor concept has been made. In discharges heated with electron cylotron resonance heating of 5−6 MW, central densities in the range of 0.7−1.2 × 1020 m−3 have been reached in combination with full divertor heat flux detachment. Also, significant neutral gas pressures and neutral compression ratios were shown for the first time in combination with reduced divertor particle flux. The divertor heat loads drop by an order of magnitude from >5 MW m−2 to below 0.5 MW m−2 with increasing density, and substantial compression of neutrals reaching neutral pressure in the sub-divertor volume of >6.0 × 10−4 mbar was seen. These elevated neutral pressure levels can be obtained and maintained with an up to 80% reduction of the particle fluxes onto the divertor target tiles. This discharge scenario was held stably detached for up to 28 seconds, which is equivalent to several hundred energy confinement times τE and longer than the time scales for current relaxation. No impurity accumulation was seen at constant Zeff ≈ 1.5 and the stored energy stayed constant at levels of Wdia > 600 kJ. The level of neutral pressure and compression reached in this scenario extrapolates well to the steady-state particle exhaust requirements for high-performance steady-state operation in OP2, in which the fully actively cooled high-heat-flux divertor will be available. An overview of this recently discovered divertor regime is given and the status ofthe physics understanding based on modeling of these regimes with the EMC3-EIRENE code is presented. [ABSTRACT FROM AUTHOR]

In this work we present the first measurements obtained by the V-band Doppler reflectometer during the second operation phase of Wendelstein 7-X to discuss the influence in the velocity shear layer and the radial electric field, Er, of several plasma parameters such as magnetic configuration, rotational transform or degree of detachment. In the first place, we carry out a systematic characterization of the turbulence rotation velocity profile in order to describe the influence of density and heating power on Er under the four most frequent magnetic configurations. The E value in the edge is found to increase with configurations featuring higher ι, although this does not apply for the high mirror configuration, KJM. As well, the Er value in the SOL and the velocity shear near the separatrix are found to display a clear dependence on heating power and density for all configurations. For a number of relevant cases, these results are assessed by comparing them to neoclassical predictions obtained from the codes DKES and KNOSOS, finding generally good agreement with experimental results. Finally, the evolution of Er at the edge is evaluated throughout the island-divertor detachment regime achieved for the first time in the 2018 campaign. After detachment, E is reduced both at the SOL and edge, and the plasma column shrinks, with the shear layer seemingly moving radially inwards from the separatrix. [ABSTRACT FROM AUTHOR]

Electron temperature Te and density ne have been measured in the divertor island of Wendelstein 7-X by means of a new thermal He-beam system utilizing line-ratio spectroscopy. This system is equipped with a poloidal arrangement of five gas valves that allows to infer 2D plasma profiles Te(R,Z) and ne(R,Z) across the magnetic island forming the island divertor. A local maximum in Te(R,Z) was measured along the flux surfaces located between the island center and the outer separatrix. In the island center, characterized by closed field lines, a clear local minimum in Te was measured, indicating that no direct heat deposition occurs in this region and that the perpendicular heat flux from the core plasma into this domain is small. Increasing the island size leads to a reduction of the connection length in the scrape-off layer (SOL) and to a shift of the island center towards the divertor target. In this scenario, the Te(R,Z) are reduced in the entire island domain. Density profiles are found flat across the divertor island for both island size scenarios. These findings are supported by similar observations made with a reciprocating probe plunging through the same SOL island at the mid-plane of W7-X representing the upstream position in this 3D divertor geometry. [ABSTRACT FROM AUTHOR]

The Third IAEA Technical Meeting on Fusion Data Processing, Validation and Analysis took place at the IAEA Headquarters in Vienna, Austria, from 28–31 May and brought together more than 60 scientists and engineers from 19 Member States, the European Commission and ITER Organisation working on data analysis and machine learning (ML) methods for the processing of fusion data, collected either from experimental diagnostics or from plasma simulations. 'Accurate data processing leads to a better understanding of the physics related to nuclear fusion research. It is essential for the careful estimate of the error bars of the raw measurements and processed data,' said D. Mazon (Chair, CEA, France) in his introductory talk and progress has been shown in this direction during the meeting. In particular the meeting discussed new developments in fusion R&D applications in the following areas: inversion techniques, such as tomography; magnetic topology reconstruction, such as equilibrium reconstructions; system identification; scaling laws determination and their accuracy for extrapolation from current machines to fusion reactors; model-based algorithms for control applications; identification of spurious and undesired events, such as disruptive phenomena or hot spots in infra-red images using sophisticated mathematical techniques like neural network and support vector machines. Discussions also focused on the potential use of these techniques for ITER, in particular, on their relevance to the first ITER plasma. The use of Integrated Modelling & Analysis Suite (IMAS) infrastructure—a convenient platform that allows implementation of different simulation codes in the same format—and synthetic diagnostics that could be coupled with this structure, in order to test the different mathematical approaches developed by the meeting participants, were explored in detail. Recent results and progress in the development of new tools will be discussed during the next meeting on the topic, scheduled in 2021. [ABSTRACT FROM AUTHOR]