The liquid metal shield laboratory (LiMeS-Lab) will provide the infrastructure to develop, test, and compare liquid metal divertor designs for future fusion reactors. The main research topics of LiMeS-lab will be liquid metal interactions with the substrate material of the divertor, the continuous circulation and capillary refilling of the liquid metal during intense plasma heat loading and the retention of plasma particles in the liquid metal. To facilitate the research, four new devices are in development at the Dutch Institute for Fundamental Energy Research and the Eindhoven University of Technology: LiMeS-AM: a custom metal 3D printer based on powder bed fusion; LiMeS-Wetting, a plasma device to study the wetting of liquid metals on various substrates with different surface treatments; LiMeS-PSI, a linear plasma generator specifically adapted to operate continuous liquid metal loops. Special diagnostic protection will also be implemented to perform measurements in long duration shots without being affected by the liquid metal vapor; LiMeS-TDS, a thermal desorption spectroscopy system to characterize deuterium retention in a metal vapor environment. Each of these devices has specific challenges due to the presence and deposition of metal vapors that need to be addressed in order to function. In this paper, an overview of LiMeS-Lab will be given and the conceptual designs of the last three devices will be presented.

Magnum-PSI is a linear plasma generator equipped with a superconducting magnet, assuring fusion devices relevant conditions at plasma–surface interface. The plasma column was diagnosed using 64 probes embedded in the target. The cross-sectional distributions of plasma parameters (floating potential, ion saturation current and electron temperature) were measured for hydrogen and deuterium plasmas under various discharge conditions. The radial profile of the floating potential across the plasma column can be described by a reversed Mexican hat-like wavelet, having the most negative potential at the center of the plasma column. The negativity of the floating potential diminishes when the discharge current increases or the magnetic field decreases. The axial gradient of the floating potential is reduced by increasing the magnetic field. The ion saturation current is maximum at the center of the plasma column, increasing with both the discharge current and magnetic field. The ion flux to the target, estimated from Thomson scattering (TS) data, was confirmed by probe measurements. The electron temperature estimated from the ion branch of the probe current–voltage characteristic is few times larger than that obtained from TS. By increasing the gas pressure in the target chamber, the time-dependent ion saturation current measured by probes changes from a constant average current (when the plasma column is attached to the target) to a fluctuating average current with scattered peaks (in a partially detached regime) which vanishes completely in the fully detached regime. With respect to hydrogen, the plasma column is wider in deuterium and is characterized by less negative floating potential distributions.

Detachment experiments have been carried out in the linear plasma device Magnum-PSI by increasing the gas pressure near the target. In order to have a proper detailed analysis of the mechanism behind momentum and power loss in detachment, a quantitative match is pursued between B2.5-Eunomia solutions and experimental data. B2.5 is a multi fluid plasma code and Eunomia is a Monte Carlo solver for neutral particles, and they are coupled together to provide steady-state solution of the plasma and neutral distribution in space. B2.5-Eunomia input parameters are adjusted to produce a close replication of the plasma beam measured in the experiments without any gas puffing in the target chamber. Using this replication as an initial condition, the neutral pressure near the plasma beam target is exclusively increased during simulation, matching the pressures measured in the experiments. Reasonable agreement is found between the electron temperature of the simulation results with experimental measurements using laser Thomson scattering near the target. The simulations also reveal the effect of increased gas pressure on the plasma current, effectively reducing the current penetration from the plasma source. B2.5-Eunomia is capable of reproducing detachment characteristics, namely the loss of plasma pressure along the magnetic field and the reduction of particle and heat flux to the target. The simulation results for plasma and neutrals will allow future studies of the exact contribution of individual plasma-neutral collisions to momentum and energy loss in detachment in Magnum-PSI.

In this paper, the pixelated phase mask (PPM) method of interferometry is applied to coherence imaging (CI)-a passive, narrowband spectral imaging technique for diagnosing the edge and divertor regions of fusion plasma experiments. Compared to previous CI designs that use a linear phase mask, the PPM method allows for a higher possible spatial resolution. The PPM method is also observed to give a higher instrument contrast (analogous to a more narrow spectrometer instrument function). A single-delay PPM instrument is introduced as well as a multi-delay system that uses a combination of both pixelated and linear phase masks to encode the coherence of the observed radiation at four different interferometer delays simultaneously. The new methods are demonstrated with measurements of electron density ne, via Stark broadening of the Hγ emission line at 434.0 nm, made on the Magnum-PSI linear plasma experiment. A comparison of the Abel-inverted multi-delay CI measurements with Thomson scattering shows agreement across the 3 × 1019 < ne < 1 × 1021 m-3 range. For the single-delay CI results, agreement is found for ne > 1 × 1020 m-3 only. Accurate and independent interpretation of single-delay CI data at lower ne was not possible due to Doppler broadening and continuum emission.

In the direct vicinity of plasma-facing surfaces, the incident plasma particles interact with surface-recombined neutrals. Remarkably high near-surface plasma pressure losses were observed in the high-flux linear plasma generator Magnum-PSI. Combining the incoherent and coherent Thomson scattering diagnostics, we directly measured particle, momentum and energy fluxes down to 3 mm from the plasma target surface. At the surface, the particle and total heat flux were also measured, using respectively an in-target Langmuir probe and thermographic methods. The near-surface momentum and energy losses scale with density, and amount to at least 50% and 20%, respectively, at n e = 8 × 1020 m−3. These losses are attributed to the efficient exchange of charge, momentum and energy between incident plasma and surface-recombined neutrals. In low-temperature plasmas with sufficient density, incident particles go through several cycles of interaction and surface deposition before leaving the plasma, thereby providing an effective alternative dissipation channel to the incident plasma. Parallel plasma parameter profiles exhibit a transition with increasing plasma density. In low-density conditions, the plasma temperature is constant and near-surface ion acceleration is observed, attributed to the ambipolar electric field. Conversely, deceleration and plasma cooling are observed in dense conditions. These results are explained by the combined effect of ion–neutral friction and electron–ion thermal equilibration in the so-called thermalized collisional pre-sheath. The energy available for ambipolar acceleration is thus reduced, as well as the upstream flow velocity. In the ITER divertor, enhanced near-surface p–n interaction is expected as well, given the overlap in plasma conditions. Including these effects in finite-element scrape-off layer models requires a near-surface resolution smaller than the neutral mean free path. This amounts to 1 mm in Magnum-PSI, and possibly an order of magnitude smaller in ITER.

Tungsten and tungsten-based alloys are the leading material choices for the divertor plasma facing components (PFCs) in future fusion reactors. Recrystallization may occur when they undergo high heat loads, drastically modifying the predesigned grain structures and the associated desired mechanical properties. However, the influence of recrystallization on the thermal fatigue behavior of tungsten PFCs still remains unclear. In this study, ITER-grade tungsten was simultaneously exposed to a high-flux hydrogen plasma (∼5 × 10 24 m −2 s −1 ) and high-cycle (10 4 –10 5 ) transient heat loads in the linear plasma device Magnum-PSI. By correlating the surface temperature distribution, obtained by analyzing temperature-, wavelength-, and surface-dependent emissivity, and the surface modifications of the plasma exposed specimens, the crack initiation heat flux factor threshold was found to be ∼2 MW m −2 s 0.5 (equivalently, ∼0.07 MJ m −2 for a 1 ms pulse). Based on electron backscatter diffraction analyses of cross-sections near the crack initiation sites, faster recrystallization kinetics near the surface compared to literature was observed and the surface cracks preferentially initiated at high angle grains boundaries (HAGBs). Upon recrystallization, the yield strength decreases which entails increasing cyclic plastic strains. The HAGBs fraction is increased, which constrains the transfer of plastic strains at grain boundaries. The recrystallization decreases the dislocation density, which promotes heterogeneous deformation. All these mechanisms explain the reduced crack initiation threshold of recrystallized tungsten compared to its as-received counterpart. The results provide new insights into the structural failure mechanisms in tungsten PFCs exposed to extreme fusion plasmas.

For thermionic emission rates exceeding the incident plasma electron flux, recent theory proposes an inverse sheath regime, with promising properties for future application in fusion edge plasmas. With the aim of inducing thermionic emission in fusion-relevant plasma conditions, several lanthanum hexaboride probes were heated in the linear plasma generator Magnum-PSI. During exposures at low plasma power and additional pulsed laser heating, the probe’s floating potential was reduced by up to 12%, providing a possible indication of thermionic emission. However, these observations coincided with rapid erosion of probe material, attributed to enhanced lanthanum self-sputtering. During follow-up experiments with helium plasmas at electron temperatures around 1 eV, the lanthanum ion impact energy and sputtering yield were reduced, and rapid erosion was avoided, thus confirming the thesis of self-sputtering. A parameter scan of plasma power resulted in LaB6 surface temperatures up to 2450 °C, exceeding the theoretical inverse sheath threshold temperature by over 300 °C. However, the probe’s floating potential did not deviate from reference measurements using a probe with high electronic work function, indicating absence of strong thermionic emission. This apparent discrepancy is attributed to the effects of probe surface modifications as observed during these experiments: impurity deposition, erosion and cavity formation. These modifications possibly affected the LaB6 electronic work function, thereby keeping the inverse sheath threshold out of reach. In conclusion, although LaB6 has one of the lowest work functions available, the inverse sheath threshold conditions could not be reached with the present setup in Magnum-PSI. Surface modifications thus do form a limiting factor for the application of LaB6 in fusion-relevant plasma conditions. Moreover, the window of stable operation for LaB6 in dense hydrogen plasmas is limited below ∼1.5 eV, and does not overlap with the conditions expected in the edge region of future fusion devices like ITER.

A cylindrical target with a high degree of closure was exposed to ITER divertor-relevant plasmas with typical electron temperatures of 2 eV, electron densities of 5⋅1020 m−3, and heat fluxes up to 20 MWm−2 in the linear device Magnum-PSI. By terminating the plasma in an unpumped closed volume, neutral pressures were enhanced from about 0.5 to 20 Pa without any increase in the neutral flux returning to the plasma. Such pressures were sustained largely by the pressure exerted by the incoming plasma. By means of hydrogen gas injection, internal neutral pressures of up to 40 Pa were reached during plasma exposure. We find that at these high neutral pressures, a < 1 eV recombination front forms and expands from the back of the cylinder, so that downstream density drops dramatically. Furthermore, in these scenarios, heat deposition to the back plate vanishes and is redirected to the upstream part of the cylinder and to hot neutrals, which can carry 50% of the plasma input power. A power balance analysis reveals that even without additional gas puffing, only about 10% of the incoming heat load reaches the back plate for the 20 MWm−2 plasma. These results demonstrate the important role of closed target configurations and local gas puffing in mitigating plasma heat loads and indicate that the gained experience should be taken into account in next-generation divertor designs.

In this study, enhanced radial transport in a volume-recombining region in detached helium plasmas in a linear device, Magnum-PSI, was investigated. By installing a reciprocating Langmuir probe, electrostatic fluctuations with high spatiotemporal resolutions were measured and analyzed. As a result, the ion-flux profile broadening in the detached state and the coherent plasma structure, which has an internal electric field in the azimuthal direction, were confirmed. By analyzing the emission intensities obtained with a fast framing camera viewing around the probe head, an enhanced fluctuation, which has an azimuthal mode number of m = 1, was found to be correlated with radial plasma ejection. This m = 1 mode rotates by the drift with the radial electric field and magnetic field and is correlated with the m = 0 mode. These two modes behave like a predator and prey; they quasi-periodically appear with about a quarter-period shift. Because the ion flux flowing into the target plate decreases when the radial transport is enhanced, this cross-field transport disperses the ion flux and decreases the maximum heat load applied to the target.

In this lecture an overview of applications of incoherent Thomson scattering (TS) as well as collective Thomson scattering (CTS) will be given. These are the most accurate methods for measuring the electron and ion properties, because the method is direct and non-intrusive. A CTS system based on the fundamental wavelength of a seeded Nd:YAG laser, being developed for the high density, low-temperature plasma of the linear plasma generator Magnum-PSI will be described also. The small Debye length of dense low temperature plasma enables application of CTS at relatively short laser wavelength. The combination of this CTS system and existing incoherent TS system enables determination of electron density and temperature as well as ion temperature and plasma velocity of the near surface plasma. In this lecture, the theoretical background and experimental challenges of the work will be given along with some examples that demonstrate the capabilities of such systems.

Tin (Sn) is an attractive option for a liquid metal wall material for future fusion reactors. Control of tritium inventory is key for the successful operation of these reactors, but little data exists up until now on hydrogen isotope retention in Sn. Free surface Sn targets and Sn-based capillary porous structure targets were exposed to deuterium (D) plasma in nano-PSI and magnum-PSI respectively. The retained D inventory was determined using the methods of thermal desorption spectroscopy and nuclear reaction analysis. The retention dependence is somewhat complex due to the mixed composition of the exposed samples as well as their liquid nature. The D retained in both types of Sn targets was found to increase with increasing D plasma fluence. For free surface liquid Sn targets, both thermal desorption spectroscopy and nuclear reaction analysis measurements showed a negative relationship between D retention and sample temperature. For capillary porous structure Sn targets, D retained in the top layer measured by nuclear reaction analysis decreased with temperature while the total D retained measured by thermal desorption spectroscopy remained approximately constant. By extracting pure Sn pieces from the targets it was found that the amount of D retained in pure Sn was much lower than that in the whole Sn-based targets and was estimated to be about 10(-7)-10(-4) D/Sn. D retained at the Sn-wall interface was found to dominate the total amount of D retained in the whole sample and observed cavities between deposited Sn droplets and the wall are the leading candidates responsible for this. Cavity formation is proposed to be the main retention mechanism for D in liquid Sn targets, although enhanced solubility leading to supersaturation under a D plasma environment is mainly responsible for the observed higher D retention in pure Sn compared with normal solubility under D gas. When compared with tungsten, D in Sn samples is of the same order of magnitude at temperatures below 300 degrees C, but at higher temperatures at least one to two orders of magnitude higher, most likely due to D trapped in cavities.

A robust impurity detection and tracking code, able to generate large sets of dust tracks from tokamak camera footage, is presented. This machine learning-based code is tested with cameras from the Joint European Torus, Doublet-III-D, and Magnum-PSI and is able to generate dust tracks with a 65-100% classification accuracy. Moreover, the number dust particles detected from a single camera shot can be up to the order of 1000. Several areas of improvement for the code are highlighted, such as generating more significant training data sets and accounting for selection biases. Although the code is tested with dust in single two-dimensional camera views, it could easily be applied to multiple-camera stereoscopic reconstruction or nondust impurities.

In this work we investigate the effects induced by the presence of nitrogen in a detached-like hydrogen plasmas in linear plasma machine Magnum-PSI. Detachment has been achieved by increasing the background neutral pressure in the target chamber by means of H-2/N-2 puffing and two cases of study have been set up, i. e. at 2 and 4 Pa. Achieved ne are ITER-relevant i. e. above 10(20) m(-3) and electron temperatures are in the range 0.8-2 eV. A scan among five different N-2/H-2+ N-2 flux ratios seeded have been carried out, at values of 0, 5, 10, 15 and 20%. A ne decrease while increasing the fraction of N2 has been observed for both background pressures, resulting in a plasma pressure drop of.30%. Te remains constant among all scans. The peak intensity of NH*(A(3)Pi->X-3 Sigma(-), Delta v= 0) at 336 nm measured with optical emission spectroscopy increases linearly with the N-2 content, together with the NH3 signal in the RGA. A further dedicated experiment has been carried out by puffing separately H-2/N-2 and H-2/He mixtures, being helium a poorly-reactive atomic species, hence excluding a priori nitrogen-induced molecular assisted recombination. Interestingly, plasma pressure and heat loads to the surface are enhanced when increasing the content of He in the injected gas mixture. In the case of N-2, we observe an opposite behavior, indicating that NeH species actively contribute to convert ions to neutrals. Recombination is enhanced by the presence of nitrogen. Numerical simulations with two different codes, a global plasma-chemical model and a spatially-resolved Monte Carlo code, address the role of NHx species behaving as electron donor in the ion conversion with H+ by means of what we define here to be N-MAR i. e. NHx+ H+ -> NHx+ + H, followed by NHx+ + e(-) -> NHx-1 + H. Considering the experimental findings and the qualitative results obtained by modelling, N-MAR process is considered to be a possible plasma-chemical mechanism responsible for the observed plasma pressure drop and heat flux reduction. Further studies with a coupled code B2.5-Eunomia are currently ongoing and may provide quantitative insights on the scenarios examined in this paper.

The Magnum-PSI facility is unique in its ability to produce and even exceed the heat and particle fluxes expected in the divertor of a fusion reactor, combined with good access to the plasma-material interaction region for diagnostics and relatively easy sample manipulation. In addition, it is possible to study the effects of transient heat loads on a plasma-facing surface, similar to those expected during so called Edge Localized Modes. By virtue of a newly installed superconducting magnet, Magnum-PSI can now maintain these conditions for hours on end for truly long term tests of candidate plasma facing materials. The electron density and temperature in the plasma beam center as a function of different magnetic fields up to 1.6 T, gas flow and source current are determined: particle fluxes greater than 10(25) m(-2)s(-1) and heat fluxes of up to 50 MW m(-2) are obtained. Linear regression and artificial neural network analysis have been used to gain insight in the general behavior of plasma conditions as a function of these machine settings. The plasma conditions during transient plasma heat loading have also been determined. These capabilities are now being exploited to reach fluence of up to 10(30) particles m(-2) at ITER-relevant conditions, equivalent to a significant fraction of the divertor service lifetime for the first time.

In the quest to long-term operation of high-power magnetically confined fusion devices, it is crucial to control the particle and heat loads on the wall. In order to predict these loads, understanding of the plasma-wall interaction is important. Near the wall surface, the plasma is accelerated towards the Debye sheath edge. In plasma conditions with high density and low temperature, the interaction between the incoming plasma and recycled neutrals can become important. In this paper, we present incoherent Thomson Scattering (TS) measurements in the near-surface region of the Magnum-PSI linear plasma generator. To enable TS measurements close to the plasma target of Magnum-PSI, a stray light suppression up to a factor 10(4) was achieved, while retaining high transmission. By incrementally moving the target along the magnetic field, this adapted system was used down to 1.9 mm from the target. In the last 10-15 mm in front of the surface, the electron density as well as temperature were observed to decrease significantly. Under the assumption of constant particle flux in this region, the density drop indicates plasma acceleration. In that case, the measurements can be interpreted to show the plasma presheath, and its lengthscale: similar to 1 cm. The electron cooling indicates an energy loss channel for the electrons near the wall. A reduced electron temperature near the sheath entrance leads to lower estimates of particle and energy flux, as well as ion impact energy, on the target surface.

The interaction of intense plasma impacting on the wall of a fusion reactor is an area of high and increasing importance in the development of electricity production from nuclear fusion. In the Magnum-PSI linear device, an axial magnetic field confines a high density, low temperature plasma produced by a wall stabilized DC cascaded arc into an intense magnetized plasma beam directed onto a target. The experiment has shown its capability to reach conditions that enable fundamental studies of plasma-surface interactions in the regime relevant for fusion reactors such as ITER: 10(23)-10(25)m(-2)s(-1) hydrogen plasma flux densities at 1-5 eV for tens of seconds by using conventional electromagnets. Recently the machine was upgraded with a superconducting magnet, enabling steady-state magnetic fields up to 2.5T, expanding the operational space to high fluence capabilities for the first time. Also the diagnostic suite has been expanded by a new 4-channel resistive bolometer array and ion beam analysis techniques for surface analysis after plasma exposure of the target. A novel collective Thomson scattering system has been developed and will be implemented on Magnum-PSI. In this contribution, the current status, capabilities and performance of Magnum-PSI are presented.

The first comparison of dust radius and surface temperature estimates, obtained from spectroscopic measurements of thermal radiation, with simulations of dust heating and vaporization by the MIGRAINe dust dynamics code is reported. The measurements were performed during controlled tungsten dust injection experiments in the cold and dense plasmas of Pilot-PSI, reproducing ITER divertor conditions. The comparison has allowed us to single out the dominating role of the work function contribution to the dust heating budget. However, in the plasmas of interest, dust was found to enter the strong vaporization regime, in which its temperature is practically insensitive to plasma properties and the various uncertainties in modeling. This makes the dust temperature a poor figure of merit for model validation purposes. On the other hand, simple numerical scalings obtained from orbital-motion-limited estimates were found to be remarkably robust and sufficient to understand the main physics at play in such cold and dense plasmas.

Plasma detachment in tokamak divertors reduces the particle and power fluxes to the plasma facing components and is essential for successful operation of ITER. The linear plasma generator Pilot-PSI can produce a high density (similar to 10(21) m(-3)), low temperature (similar to 1 eV) plasma which is similar to that expected at the ITER divertor strike-points during the partially detached regime [1]. Given the simple geometry of the device, Pilot-PSI allows to diagnose the plasma beam at multiple axial positions. In this study, the incoherent Thomson scattering (TS) diagnostic [2] is exploited to measure the radial plasma T-e and n(e) profiles at two locations of the hydrogen plasma beam: near the plasma source (upstream) and 2 cm from the target plate. At the target, the TS measurements are supported by an embedded Langmuir probe. These measurements prove the existence of parallel plasma pressure loss as well as particle loss, confirming that physical processes believed to cause detachment in tokamak divertors also hold in Pilot-PSI (ion-neutral friction, volume recombination). It is found that the fractional reduction of the plasma pressure varies between similar to 4 and similar to 5000, depending strongly on the pressure of the background neutrals. The importance of individual loss channels is discussed. (C) 2017 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND license. (http://creativecommons.org/licenses/by-nc-nd/4.0/)

The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma–material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.