The largest current stellarator is Wendelstein 7-X at the Max Planck Institute for Plasma Physics in Greifswald, Germany, whose Alkali Beam Emission Spectroscopy diagnostics (ABES) are built and operated by the Fusion Plasma Physics Department of the Centre for Energy Research. Quasi-coherent (QC) oscillations of the plasma density in the edge plasma of the stellarator and in the surrounding magnetic island divertor in the 10-25 kHz range were observed. In addition to ABES, the fluctuations were also observed using other diagnostic techniques such as reflectometry, Mirnov coils and soft X-ray detectors. This phenomenon was found in the spatial range where plasma filaments are formed and ejected from the plasma, and a link between the two phenomena was suggested. Using reflectometry, the parametric dependence of the frequency of the oscillations and the poloidal wavelength of the waves was studied in detail. However, this diagnostic cannot measure the plasma density profile and plasma filaments. In this paper, we investigate the relationship between quasi-coherent modes in the edge plasma and in the edge magnetic island.

The thermal and mechanical loads during disruptions are a major threat for large reactor-class tokamak devices. Therefore, shattered pellet injection (SPI) is selected as the baseline technology for the ITER Disruption Mitigation System (DMS). The aim of DMS Support Laboratory located at the HUN-REN Centre for Energy Research is to study the production, launch and shattering of cryogenic protium, deuterium, and neon pellets in the ITER geometry. This paper reports on the fragment analysis procedure and the first results of the fragment plume investigation performed with pellets made of the above materials and accelerated to a speed ranging between 70 and 500m/s. The experimental results show that the fragment plume consists of macroscopic and microscopic fragments. The shattering of about 500 m/s protium and deuterium pellets resulted in a most common fragment size of a few millimeters, while at lower velocities around 250m/s still the same few millimeters size fragments dominate but larger fragments have a higher proportion in the mass distribution. At low velocities (70 m/s) for neon pellets most of the pellet mass was converted into large fragments (from 10 to 20 mm ). The results are also compared with the Parks pellet fragmentation model.

After a long device enhancement phase, scientific operation resumed in 2022. The main new device components are the water cooling of all plasma facing components and the new water-cooled high heat flux divertor units. Water cooling allowed for the first long-pulse operation campaign. A maximum discharge length of 8 min was achieved with a total heating energy of 1.3 GJ. Safe divertor operation was demonstrated in attached and detached mode. Stable detachment is readily achieved in some magnetic configurations but requires impurity seeding in configurations with small magnetic pitch angle within the edge islands. Progress was made in the characterization of transport mechanisms across edge magnetic islands: Measurement of the potential distribution and flow pattern reveals that the islands are associated with a strong poloidal drift, which leads to rapid convection of energy and particles from the last closed flux surface into the scrape-off layer. Using the upgraded plasma heating systems, advanced heating scenarios were developed, which provide improved energy confinement comparable to the scenario, in which the record triple product for stellarators was achieved in the previous operation campaign. However, a magnetic configuration-dependent critical heating power limit of the electron cyclotron resonance heating was observed. Exceeding the respective power limit leads to a degradation of the confinement.

Shattered Pellet Injection (SPI) is a common technique used in Disruption Mitigation Systems to prevent or minimize the effect of plasma disruptions in Tokamaks. To operate an SPI system, a fast-acting valve is needed for accelerating pellets. An eddy current actuated high-pressure fast valve was developed in 2021, dedicated to the SPI system of the ITER DMS Support Laboratory at the Centre for Energy Research. Based on the experiences of SPI operation, a novel fast valve has been developed to improve durability and reliability, while the design has been simplified in terms of manufacturing and assembly aspects. The results of the development and laboratory testing of this novel eddy current actuated fast valve are discussed in this contribution.

Plasma filaments have been measured with alkali beam emission spectroscopy in the plasma edge, divertor island, and scrape-off layer of Wendelstein 7-X. Due to the high intensity of a 1–2 kHz plasma mode, a new, correlation based conditional averaging algorithm was used to search for filaments in the signals. With that method, effects of different magnetic configurations and density levels on filament properties are observed. In configurations where the islands are small and do not play an important role for the connection length topology, filaments behave similar to tokamaks. In contrast, in configurations with larger magnetic islands and more complex connection length profiles, filaments behave quite differently, for instance they may or may not appear in the inner side of the divertor island depending on the plasma parameters. Coupling between the filaments and lower frequency events are also showed. The role of filaments in the global and local particle transport is briefly discussed.

Measurements of ion temperature profiles are required to assess the energy and particle transport processes in the Wendelstein 7-X stellarator. This device is equipped with a diagnostic alkali beam, which can be utilized to determine local impurity temperatures and densities by Charge Exchange Recombination Spectroscopy (CXRS). It could provide such profiles in the edge plasma, where other diagnostics are less efficient. With this contribution, first results of CXRS measurements on the sodium beam from the scientific operation phase OP2.1 are presented. The spectroscopic system was in commissioning phase lacking some of the final optical components. Thus, the aim of the diagnostics during this campaign was to explore the measurement capabilities. Based on the processed spectra, the prospects of C5+ and C6+ ion temperature and concentration measurements are discussed. The results indicate that with the final optical setup under installation, the diagnostics could provide ion temperature profiles in the edge with 3 mm radial resolution and at least 1 s temporal resolution.

The analysis of the gas expansion in the test bench of the ITER DMS Support Laboratory is discussed. For this purpose, numerical FEM simulations were performed during the design phase of the laboratory. These are compared with experimental data, that have been obtained after the construction of the test bench was finished. The numerical and experimental results have been found to agree well. Both indicate the emergence of pressure waves in the acceleration barrel. These contribute to the pellet detachment from the barrel wall. Supersonic velocities have been also observed in the simulations, and an indirect indication for their presence has been found in the experimental results. The main processes during the propellant gas expansion inside the propellant gas recovery chamber are discussed, and a rough estimate for the amount of gas entering the flight tube of the test bench before the pellet is presented.

ITER is a large-scale experimental nuclear fusion reactor currently under construction. To protect the first wall components of the machine against damages due to a plasma disruption event, ITER needs a Disruption Mitigation System (DMS). The ITER DMS is based on the Shattered Pellet Injector (SPI) technology. One of the issues with SPI injection is that the propellant gas may overtake the pellet and start plasma edge cooling before material deposition by pellet fragments takes place. Edge cooling may increase runaway electron generation, therefore should be avoided. To address this problem, the ITER Organization has launched a project to develop Fast Shutter which serves an important role by closing the orifice right after the pellet passed the component. The shutter must be closed within a few milliseconds to block enough gas according to the simulations. It will be installed close to the torus, where the equipment will be exposed to a high magnetic field, and neutron irradiation. Due to limited accessibility for maintenance, the shutter must operate thousands of cycles. This paper describes the detailed design of the shutter prototype and its present capabilities. The proposed concept addresses the issue of high dynamic loads resulting from the fast-closing time of the device, which can cause fatigue due to cyclic loading. The unique environmental conditions necessitate the use of special actuators and materials. The design space available for the implementation of the concept is relatively small, which requires the optimization of the arrangement of device components to minimize the overall loads and maximize the device's efficiency.

The ITER Disruption Mitigation System will consist of Shattered Pellet Injectors (SPI) which will be fired upon request into the ITER plasma to mitigate the detrimental consequences of disruptions. Each SPI will include a fast shutter to block the propellant gas after the pellet has passed to minimize the gas amount injected before the pellet, which was found to be critical to maintain disruption mitigation efficiency. An electromechanical test bench (ETB) was designed and built where the electromechanical model of the shutter actuator has been validated. Several material combinations were tested, to find the optimal solution for the fast shutter prototype.

A support laboratory has been set up to study pellet production, launch and shattering of cryogenic protium, deuterium, and neon pellets for the ITER disruption mitigation system, which plans to use 28.5 × 57 mm (diameter × length) protium, neon an mixture pellets in the Shattered Pellet Injectors. Such large protium pellets have not been produced and launched before, therefore the desublimation and launch process have been studied in detail in two steps. First 19 mm diameter pellets were produced, followed by the demonstration of the final pellet size. Pellet desublimation recipes were established for all pellet types, and it was found that, under certain conditions, even the large neon pellets can be launched with a propellant gas pulse, without requiring a mechanical punch device. This is attributed to cryogenic snow formation on the surface of the pellet. Conditions for the snow formation are studied and tendencies are understood using simple calculations.

ITER needs a Disruption Mitigation System (DMS) for protection from the consequences of plasma disruptions during high-power operation. The current DMS for ITER is based on the Shattered Pellet Injection (SPI) technology. This works on the basic principle of a cryogenic pipe gun with around a half-meter-long acceleration barrel. It is important to avoid any contact between the pellet and the internal surfaces of the flight line to minimize the risk of premature pellet breakage. Therefore, the internal dimensions of the flight tube of the SPI are relatively large compared to the pellet hence providing a bypass for propellant gas resulting in a significant quantity arriving in the plasma ahead of the pellet. This will cause an undesirable instability in the plasma and compromise the effectiveness of disruption mitigation. The described events appear on a scale of a few milliseconds. One possible solution for this issue could be using a fast-acting shutter in the flight line, which will close after the pellet has passed through to hold back the propellant gas. Due to the environmental requirements, the shutter needs to close an approximately 40 mm aperture within a few ms, the shutter has limited space and limited access and acts under a magnetic field, therefore it should endure a high number of cycles and should perform well in a very demanding situation. This paper describes the optioneering study of the shutter and possible solutions for the demanding criteria. The laboratory testing of the Electromechanical Test Bench and the Prototype design are also detailed in separate papers.

A support laboratory for the ITER Disruption Mitigation System has been set up at the Centre for Energy Research, Budapest, Hungary. It consists of a cryogenic pellet injector capable of producing, launching and shattering 19 x 38 and 28.5 x 55 mm (d x L) cryogenic pellets. Up to now more than 300 pellets made of hydrogen, deuterium, neon and mixtures thereof have been formed and launched . Nearly all pellets flew through the 4 m long, 60 mm diameter flight tube, which represents the ITER flight path between the propellant recovery chamber and the shattering head. Initial analysis of the fragment plumes produced by an open shattering head indicate that they extend 3 times further parallel to the shattering plate than perpendicularly to it. The results show that the large pellets needed for the ITER SPI system can be produced and launched.

On a Wendelstein 7-X (W7-X), an alkali metal beam (AMB) diagnostic system was installed in order to measure the plasma edge electron density profiles and turbulence transport. A sodium beam was injected in the plasma, and the light emission was observed by an optical system. During the last operation phase, OP1.2b campaign trial spectral measurements were performed with a dedicated optical branch. The results showed the emergence of potential CX lines in the light spectra during sodium injection. The lines were identified as Carbon III, which were the dominant lines observed by other diagnostics at the edge plasma. Based on these results, an additional dedicated optical system was developed and installed in 2021 for the upcoming operational phase, OP2. The optics were designed for multiple purposes: spectral measurements for the AMB system and for a He/Ne gas jet. The system was designed to allow implementation of further diagnostics on this port later (e.g., coherence imaging system). The details of the implementation of the design requirements and the main challenges of the manufacturing process and installation are discussed in this paper.

The ITER Disruption Mitigation System (DMS) Support Laboratory at the Centre for Energy Research, Budapest, Hungary aims at the characterization of the de-sublimation, acceleration and shattering of the cryogenic pellets, which will be fired upon request into the ITER plasma to mitigate the detrimental consequences of disruptions. The test bench includes a fast-acting valve for pellet acceleration, which releases high-pressure gas on a millisecond time scale. The valve is a new development since no valve on the market existed which would have met the requirements. The electromechanical modelling of the valve actuator has been validated against a laboratory experiment, the final system was built and thoroughly tested on a test bench to characterize the valve properties, and to demonstrate that the system fulfils the requirements.

Shattered pellet injection is the baseline technology for the ITER Disruption Mitigation System (DMS) [1]. ITER DMS requires the use of large, 28.5 x 57 mm (diameter x length) cryogenic pellets, made of hydrogen, neon, or a mixture of those gasses and accelerated to several hundred m/s velocity. Beside the production, acceleration and transfer of such ITER-size pellets, the characterization of the shattering is one of the main task of the ITER DMS Support Laboratory at center of Energy Research as part of the ITER DMS Task Force program to validate the design choices of the DMS [2]. For this purpose, a fragment plume diagnostic chamber and two optical di-agnostics using fast framing cameras, namely the laser curtain at two distances from the shattering head and a shattering observation were designed and implemented. Preliminary experiments for different pellet material and velocities at 20 degrees and 30 degrees shattering angle showed a strong pellets sublimation during shattering process. The fragments seem to slide along the shattering plate, no significant total reflection was detected. No qualitative difference between 20 degrees and 30 degrees shattering angle was observed, but clear differences for different materials and or pellet speeds.

A method is presented for the swift reconstruction of electron density profiles measured by the alkali beam emission spectroscopy. It is based on the linearization of the governing rate equations and leads to a direct calculation for obtaining the profiles. The uncertainties of the measurement are incorporated into the problem through the utilization of Tikhonov regularization and the generalized least squares method. An approximation for the uncertainty of the reconstructed density data is calculated as well. The applicability of the method is tested against both simulated and real experimental results of the W7-X stellarator.

tokamak J.Galdon-Quiroga1, G.Birkenmeier1,2, V.Olevskaia1,2, M.Sochor1, G.Anda3, K.Bald1, M.Dunne1, M.Garcia-Munoz4, A.Herrmann1, K.Kaunert1, D.Nagy3, J.F.Rivero-Rodriguez4, M.Rodriguez-Ramos 4, V.Rohde1, E.Strumberger1, J.J. Toledo 4, E.Viezzer4, E.Wolfrum1, S.Zoletnik3, U.Stroth1,2 and the ASDEX Upgrade4 Team 1 Max Planck Institute for Plasma Physics, Garching, Germany 2 Physics Department E28, Technical University of Munich, Garching, Germany 3 Wigner RCP, EURATOM Association HAS, Budapest, Hungary 4 Centro Nacional de Aceleradores (CNA), University of Seville, Spain

A 60 keV neutral Alkali beam system was designed, built and installed for beam emission spectroscopy measurement of edge plasma on W7-X. As the decay length of the first excited state of sodium is much shorter than lithium it was selected as beam species. This way the steep density gradient at the bean shaped cross section of the Wendelstein 7-X plasma can be resolved. Unexpectedly a thermal sodium beam was also observed in the far scrape off layer of W7-X plasma during the first measurements in 2017. It was found to originate from the recirculating neutralizer which uses sodium vapour. To resolve this issue the neutralizer material was changed to Potassium. The observation system consists of two parts: a 40 channel avalanche photo diode (APD) camera unit and a CMOS camera which run in parallel: 95% of the collected light goes to the APD unit which is digitized with 2 MHz sampling rate while the CCD camera is operated in the 100 Hz range. Chopping the beam up to 250 kHz is also possible for precise background measurement on the time scale of the turbulence. In this paper the main improvements of the sodium beam system using potassium for ion beam neutralization and the first measurement results are described.

Atomic beam probe is an extension of the routinely used beam emission spectroscopy diagnostic for the plasma edge current fluctuation measurement at magnetically confined plasmas. Beam atoms ionized by the plasma are directed to a curved trajectory by the magnetic field and may be detected close to the wall of the device. The arrival location and current distribution of the ions carry information about the plasma current distribution, the density profile, and the electric potential in the plasma edge. This paper describes a micro-Faraday cup matrix detector for the measurement of the few microampere ion current distribution close to the plasma edge. The device implements a shallow Faraday cup matrix, produced by printed-circuit board technology. Secondary electrons induced by the plasma radiation and the ion bombardment are basically confined into the cups by the tokamak magnetic field. Additionally, a double mask is installed in the front face to limit the ion influx into the cups and supplement secondary electron suppression. The setup was tested in detail using a lithium ion beam in the laboratory. Switching time, cross talk, and fluctuation sensitivity test results in the lab setup are presented along with the detector setup to be installed at the COMPASS tokamak.

The imaging heavy ion beam probe (i-HIBP) developed at the ASDEX Upgrade tokamak is a new diagnostic concept for investigations at the edge of high temperature plasmas. By means of a heavy alkali beam injector, a neutral primary beam of an energy of 70 keV is injected into the fusion plasma, where it is ionized generating a fan of secondary beams. These are deflected by the magnetic field of the tokamak and intersect a scintillator plate in the limiter shadow of the tokamak. The light pattern on the scintillator detected with a high speed camera contains radial information on the density, electrostatic potential and the magnetic field in the edge region of the plasma. For the design of the i-HIBP, a detailed beam model including the 3D tokamak magnetic field and beam attenuation effects for cesium and rubidium atoms is developed in order to find the optimum injection scheme within the limited space of the tokamak environment for maximum signal intensities. Based on the optimized injection, the arrangement of the injector outside the vacuum-vessel and the detailed design of the optical in-vessel system is determined.