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Experiments on ASDEX Upgrade (AUG) in 2021 and 2022 have addressed a number of critical issues for ITER and EU DEMO. A major objective of the AUG programme is to shed light on the underlying physics of confinement, stability, and plasma exhaust in order to allow reliable extrapolation of results obtained on present day machines to these reactor-grade devices. Concerning pedestal physics, the mitigation of edge localised modes (ELMs) using resonant magnetic perturbations (RMPs) was found to be consistent with a reduction of the linear peeling-ballooning stability threshold due to the helical deformation of the plasma. Conversely, ELM suppression by RMPs is ascribed to an increased pedestal transport that keeps the plasma away from this boundary. Candidates for this increased transport are locally enhanced turbulence and a locked magnetic island in the pedestal. The enhanced D-alpha (EDA) and quasi-continuous exhaust (QCE) regimes have been established as promising ELM-free scenarios. Here, the pressure gradient at the foot of the H-mode pedestal is reduced by a quasi-coherent mode, consistent with violation of the high-n ballooning mode stability limit there. This is suggestive that the EDA and QCE regimes have a common underlying physics origin. In the area of transport physics, full radius models for both L- and H-modes have been developed. These models predict energy confinement in AUG better than the commonly used global scaling laws, representing a large step towards the goal of predictive capability. A new momentum transport analysis framework has been developed that provides access to the intrinsic torque in the plasma core. In the field of exhaust, the X-Point Radiator (XPR), a cold and dense plasma region on closed flux surfaces close to the X-point, was described by an analytical model that provides an understanding of its formation as well as its stability, i.e., the conditions under which it transitions into a deleterious MARFE with the potential to result in a disruptive termination. With the XPR close to the divertor target, a new detached divertor concept, the compact radiative divertor, was developed. Here, the exhaust power is radiated before reaching the target, allowing close proximity of the X-point to the target. No limitations by the shallow field line angle due to the large flux expansion were observed, and sufficient compression of neutral density was demonstrated. With respect to the pumping of non-recycling impurities, the divertor enrichment was found to mainly depend on the ionisation energy of the impurity under consideration. In the area of MHD physics, analysis of the hot plasma core motion in sawtooth crashes showed good agreement with nonlinear 2-fluid simulations. This indicates that the fast reconnection observed in these events is adequately described including the pressure gradient and the electron inertia in the parallel Ohm’s law. Concerning disruption physics, a shattered pellet injection system was installed in collaboration with the ITER International Organisation. Thanks to the ability to vary the shard size distribution independently of the injection velocity, as well as its impurity admixture, it was possible to tailor the current quench rate, which is an important requirement for future large devices such as ITER. Progress was also made modelling the force reduction of VDEs induced by massive gas injection on AUG. The H-mode density limit was characterised in terms of safe operational space with a newly developed active feedback control method that allowed the stability boundary to be probed several times within a single discharge without inducing a disruptive termination. Regarding integrated operation scenarios, the role of density peaking in the confinement of the ITER baseline scenario (high plasma current) was clarified. The usual energy confinement scaling ITER98( p,y ) does not capture this effect, but the more recent H20 scaling does, highlighting again the importance of developing adequate physics based models. Advanced tokamak scenarios, aiming at large non-inductive current fraction due to non-standard profiles of the safety factor in combination with high normalised plasma pressure were studied with a focus on their access conditions. A method to guide the approach of the targeted safety factor profiles was developed, and the conditions for achieving good confinement were clarified. Based on this, two types of advanced scenarios (‘hybrid’ and ‘elevated’ q -profile) were established on AUG and characterised concerning their plasma performance.

We reply to the comment by Li et al (submitted to Nucl. Fusion with this response) on our recent paper Nishijima et al (2023 Nucl. Fusion 63 126003). In this response, we address the existence of an incident ion energy, E _i , threshold for the deuterium (D) supersaturated surface layer (DSSL) formation with a newly conducted D plasma exposure experiment at E _i ∼ 20 eV. It is also further demonstrated, based on new experiments where the ion flux, fluence, and sample temperature are scanned, that non-kinetic (ballistic) processes play a role in the DSSL formation and growth. In addition, the effect of impurities in our plasma is discussed also with a new analysis of the surface composition made after a D plasma exposure.

Future fusion reactors will have to breed enough tritium (T) to sustain continuous operation and to produce excess T to power up other fusion reactors. Therefore, T is a scarce resource that must not be lost inside the fusion power plants systems. The factor that describes the T production is the ‘tritium breeding ratio’ (TBR) which is the ratio of the breading rate in atoms per second to the burn rate in atoms per second. Its value is calculated from neutronics analyses of the breeding process in the blanket and coupled dynamics of the T processing plant. However, these calculations generally ignore the T transport and loss in the first wall by assuming essentially instantaneous recycling of the impinging T in-flux. In this paper the transport and retention of T in the main chamber first wall of a future EU-DEMO reactor is investigated based on the available material data and expected particle loads onto the wall. Two breeding blanket concepts are compared WCLL (water cooled lithium lead) and HCPB (helium cooled pebble bed) and the resulting wall-loss probabilities are compared with a simple balance model that describes the maximum allowable wall loss given a TBR to achieve T-self-sufficiency.

Properties of deuterium (D) supersaturated surface layers (DSSLs) formed in tungsten (W), such as thickness, internal microstructures, and D retention, are experimentally investigated as a function of the incident ion energy, E _i . W samples were exposed to D plasmas in the PISCES-A linear plasma device in a range of E _i ∼ 45–175 eV, while other plasma exposure parameters were fixed: sample temperature, T _s , ∼423 K, ion flux, Γ _i , ∼1.2 × 10 ^21 m ^−2 s ^−1 , and fluence, Φ _i , ∼3.0 × 10 ^24 m ^−2 . High-resolution, cross-sectional, transmission electron microscopy observations confirm that (1) a DSSL forms even at the lowest E _i ∼ 45 eV, (2) the DSSL thickness, Δ t _DSSL , is found to decrease with decreasing E _i from ∼11–12 nm at E _i ∼ 175 eV to ∼5–6 nm at ∼45 eV, and to agree with approximately the maximum implantation depth calculated using SDTrimSP, and (3) high-density D nanobubbles with a diameter of ∼1 nm or less exist inside the DSSL, which is deemed to validate a theory-predicted vacancy stabilization process due to trapping of a solute D atom(s). Utilizing a D areal density of ∼4.2 × 10 ^19 m ^−2 in the first 14 nm from the surface at E _i ∼ 75 eV from nuclear reaction analysis and the measured E _i dependence of Δ t _DSSL , our previous laser-induced breakdown spectroscopy data is updated: both dynamic and static D retention increase with decreasing E _i , and the D/W atomic fraction during plasma exposure reaches ∼0.3 at E _i ∼ 45 eV. A possible DSSL formation mechanism is proposed.

Single crystalline tungsten samples were irradiated by He, D, and H at room temperature, with a calculated damage level in the peak maximum up to 0.04 dpa and with He up to 0.5 dpa. The dislocation structure of the whole damage zone was investigated perpendicular to the irradiated surface by Transmission Electron Microscopy (TEM) under four different two-beam diffraction conditions g = −200, 020, −110, 110 close to the [100] zone axis, taking bright-field kinematical images. The depth of the damage zone observed by TEM is in good agreement with the damage depth calculated by SRIM. The damage zones of the tungsten samples irradiated by He, D, H up to 0.04 dpa can be subdivided into four subzones. Significant differences in the dislocation structures between the different ion species despite comparable calculated damage levels were found at the depth close to the damage peak maximum. In this depth, the damage zone of tungsten irradiated by He consists of small dense dislocation loops (⩽3 nm) with high strain fields. Irradiation by D creates large dislocation loops with an average diameter of 15–22 nm depending on the set diffraction condition. The damage zone of H-irradiated tungsten is dominated by long dislocation lines at the depth of the damage peak maximum. No transition from dislocation loops to lines was found in the sample irradiated with He up to 0.5 dpa. This is probably due to He bubbles, which hinder the loops to aggregate into lines. This work shows that the displacement damage obtained by the irradiation of tungsten at room temperature with light ions is significantly different to tungsten irradiated to the same calculated damage level by medium to high-mass ions (Wielunska et al 2022 Nucl. Fusion 62 096003 [1]).

The effect of deuterium (D) presence on the amount of displacement damage created in tungsten (W) during high-energy W-ion irradiation is investigated. For this purpose, we have performed modelling of experimental results where W was sequentially or simultaneously irradiated by 10.8 MeV W ions and exposed to 300 eV D ions. A novel displacement damage creation and stabilization model was newly developed and introduced into the MHIMS-Reservoir (migration of hydrogen isotopes in materials) code. It employs macroscopic rate equations (MREs) for solving the evolution of solute and trapped D concentrations in the material. The new displacement damage creation and stabilization model is based on spontaneous recombination of Frenkel pairs and stabilization of defects that are occupied by D atoms. By using the new model, we could successfully replicate the measured D depth profiles and D thermal desorption data, where a higher defect concentration was observed when D was present during W irradiation as compared to when no D was present. For this we utilized parameters, which include the number of distinct defect types, the de-trapping energies of their fill-levels, their saturation concentrations and their probability for stabilization if they contain a D during the W-ion irradiation. To successfully replicate the experimental results three distinct defect types were needed with several fill-levels. By comparing the de-trapping energies of the defect fill-levels with data available from the literature, the defect types were identified as single-vacancies, small vacancy clusters and large vacancy clusters. The effect of D presence was found to be largest in single vacancies as its concentration increased by about a factor of three, while the concentration of small vacancy clusters increased by about a factor of two. Large vacancy clusters were found to be largely unaffected as they showed very little increase in concentration when D was present. [ABSTRACT FROM AUTHOR]

Polycrystalline tungsten (W) samples were simultaneously irradiated by 10.8 MeV W ions and exposed to 300 eV deuterium (D) ions at different temperatures ranging from 450 K to 1000 K. After simultaneous W ion irradiation and D ion exposure the samples were additionally exposed to low-energy D ions at 450 K in order to populate all the defects created beforehand. The amount of damage created was evaluated by measuring D depth profiles and D thermal desorption spectra. Results were compared with data obtained in a sequential experiment where samples were first irradiated by 10.8 MeV W ions and only afterwards exposed to 300 eV D ions at 450 K to populate the created defects. At 450 K we observed a two times higher maximum D concentration for the simultaneous case as compared with the sequential case. At 600 K and 800 K the ratio between the simultaneous and sequential case decreases to about 1.6 and 1.2, respectively, and increases again to a factor of 2 at 1000 K. We attribute this dependence on temperature to the change in the concentration of mobile and trapped D during the simultaneous exposures, which is in line with theoretical calculations predicting that trapped D in a vacancy prevents vacancy annihilation with self-interstitials. [ABSTRACT FROM AUTHOR]

The possible mutual influence and synergistic effect between defect production and the presence of hydrogen isotopes in the crystal lattice of tungsten is studied. For this purpose, we perform modeling of experimental data where polycrystalline tungsten samples were in one case sequentially irradiated by 10.8 MeV tungsten ions followed by low-energy deuterium exposure and in the other case simultaneously irradiated by tungsten ions, while exposed to deuterium atoms. Modeling of the measured deuterium depth profiles and thermal-desorption spectra for different irradiation temperatures is performed by the MHIMS (migration of hydrogen isotopes in materials) code. A model of trap creation due to tungsten ion irradiation during the deuterium atom exposures is implemented. In both experimental series, the deuterium desorption peaks corresponding to defects induced by tungsten irradiation are described by the same two de-trapping energies of 1.83 and 2.10 eV. The experiments give unambiguous proof that the presence of deuterium increases the overall trap density. The modeling reveals that the two trap concentrations are affected differently by the temperature and presence of deuterium: the concentration of the low-energy trap is significantly higher in the case of simultaneous exposure compared to sequential exposure, especially at high temperature (2.2 times higher at 1000 K). The concentration of the high-energy trap is only weakly affected by the presence of hydrogen. [ABSTRACT FROM AUTHOR]

Measurements are conducted to identify the motion of tungsten and helium atoms during the formation of tungsten fuzz. In a first series of experiments the mobility of helium within the growing fuzz was measured by adding 3He to the different stages of plasma exposure under conditions that promoted tungsten fuzz growth. Ion beam analysis was used to quantify the amount of 3He remaining in the samples following the plasma exposure. The results indicate that the retention of helium in bubbles within tungsten is a dynamic process with direct implantation rather than diffusion into the bubbles, best describing the motion of the helium atoms. In the second experiment, an isotopically enriched layer of tungsten (~92.99% 182W) is deposited on the surface of a bulk tungsten sample with the natural abundance of the isotopes. This sample is then exposed to helium plasma at the conditions necessary to support the formation of tungsten ‘fuzz’. Depth profiles of the concentration of each of the tungsten isotopes are obtained using secondary ion mass spectrometry (SIMS) before and after the plasma exposure. The depth profiles clearly show mixing of tungsten atoms from the bulk sample toward the surface of the fuzz. This supports a physical picture of the dynamic behavior of helium bubbles which, also, causes an enhanced mixing of tungsten atoms. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The effect of near-surface blisters on deuterium transport in tungsten is studied by means of nuclear reaction analysis (NRA) and scanning electron microscopy (SEM). Gentle deuterium plasma loading of different durations and subsequent NRA depth profiling is performed in heavily pre-blistered and unblistered areas on self-damaged tungsten samples. Comparison of the deuterium depth profiles reveals a considerable reduction of the deuterium transport into the bulk due to the presence of near-surface blisters. SEM and NRA results identify the enhanced re-emission of deuterium from the sample due to open blisters as the underlying mechanism, which reduces the deuterium flux into the bulk. Based on a simple analytical hydrogen retention model, the re-emitted deuterium flux by open blisters is determined to be 80% of the implanted deuterium flux in the present conducted experiment. In addition, the deuterium flux into the bulk is reduced by 60% compared to the unblistered case. Hence the presence of blisters is not a general disadvantage in the context of retention, but can be beneficial in slowing down the build up of a certain hydrogen inventory and in reducing the permeation flux. [ABSTRACT FROM AUTHOR]

The influence of helium (He) on deuterium (D) transport and retention was studied experimentally in tungsten (W). Helium was implanted 1 µm deep into W to a maximum calculated concentration of 3.4 at.%. To minimize the influence of displacement damage created during the He implantation on D retention, so-called self-damaged W was used. W was damaged by 20 MeV W ion bombardment and defects were populated by low-temperature D plasma at room temperature before He implantation. Deuterium depth profiling was performed in situ during isochronal annealing in the temperature range from 300 K to 800 K. It is shown for the first time unambiguously that He attracts D and locally increases D trapping. Deuterium retention increased by a factor of two as compared to a non-He implanted W reference after sample annealing at 450 K. Rate equation modelling can explain the measured D depth profiles quantitatively when keeping the de-trapping parameters unchanged but only increasing the number of traps in the He zone. This bolsters the confidence in the theoretical calculations predicting that more hydrogen isotopes can be stored around a He cluster zone. [ABSTRACT FROM AUTHOR]

Plasma-wall interactions have been studied in the full-W ASDEX Upgrade during its dedicated helium campaign. Relatively clean plasmas with a He content of  >80% could be obtained by applying ion cyclotron wall conditioning (ICWC) discharges upon changeover from D to He. However, co-deposited layers with significant amounts of He and D were measured on W samples exposed to ICWC plasmas at the low-field side (outer) midplane. This is a sign of local migration and accumulation of materials and residual fuel in regions shadowed from direct plasma exposure albeit globally D was removed from the vessel. When exposing W samples to ELMy H-mode helium plasmas in the outer strike-point region, no net erosion was observed but the surfaces had been covered with co-deposited layers mainly consisting of W, B, C, and D and being the thickest on rough and modified surfaces. This is different from the typical erosion-deposition patterns in D plasmas, where usually sharp net-erosion peaks surrounded by prominent net-deposition maxima for W are observed close to the strike point. Moreover, no clear signs of W nanostructure growth or destruction could be seen. The growth of deposited layers may impact the operation of future fusion reactors and is attributed to strong sources in the main chamber that under suitable conditions may switch the balance from net erosion into net deposition, even close to the strike points. In addition, the absence of noticeable chemical erosion in helium plasmas may have affected the thickness of the deposited layers. Retention of He, for its part, remained small and uniform throughout the strike-point region although our results indicate that samples with smooth surfaces can contain an order of magnitude less He than their rough counterparts. [ABSTRACT FROM AUTHOR]

Simulations of deuterium (D) atom exposure in self-damaged polycrystalline tungsten at 500 K and 600 K are performed using an evolution of the MHIMS (migration of hydrogen isotopes in materials) code in which a model to describe the interaction of D with the surface is implemented. The surface-energy barriers for both temperatures are determined analytically with a steady-state analysis. The desorption energy per D atom from the surface is 0.69  ±  0.02 eV at 500 K and 0.87  ±  0.03 eV at 600 K. These values are in good agreement with ab initio calculations as well as experimental determination of desorption energies. The absorption energy (from the surface to the bulk) is 1.33  ±  0.04 eV at 500 K, 1.55  ±  0.02 eV at 600 K when assuming that the resurfacing energy (from the bulk to the surface) is 0.2 eV. Thermal-desorption spectrometry data after D atom exposure at 500 K and isothermal desorption at 600 K after D atom exposure at 600 K can be reproduced quantitatively with three bulk-detrapping energies, namely 1.65  ±  0.01 eV, 1.85  ±  0.03 eV and 2.06  ±  0.04 eV, in addition to the intrinsic detrapping energies known for undamaged tungsten (0.85 eV and 1.00 eV). Thanks to analyses of the amount of traps during annealing at different temperatures and ab initio calculations, the 1.65 eV detrapping energy is attributed to jogged dislocations and the 1.85 eV detrapping energy is attributed to dislocation loops. Finally, the 2.06 eV detrapping energy is attributed to D trapping in cavities based on literature reporting observations on the growth of cavities, even though this could also be understood as D desorbing from the C-D bond in the case of hydrocarbon contamination in the experimental sample. [ABSTRACT FROM AUTHOR]

Surface morphology and deuterium retention in tungsten exposed at surface temperature of ~550 K to mixed deuterium–neon plasmas of different neon concentrations are investigated. It is found that the addition of neon up to 20% mitigates blistering on the surface. Cross-section view of the surface shows the formation of pores near the surface in the depth less than 100 nm. Deuterium depth profile is featured by an enhanced deuterium concentration within a depth of 16 nm but a mitigated penetration in depth larger than 1 µm. Deuterium retention is reduced by up to a factor of four. It is suggested the open pores formed in the surface serves as an escaping channel, mitigates deuterium penetration towards bulk and retention in the bulk. [ABSTRACT FROM AUTHOR]

Fundamental understanding of hydrogen–metal interactions is challenging due to a lack of knowledge on defect production and/or evolution upon hydrogen ingression, especially for metals undergoing hydrogen irradiation with ion energy below the displacement thresholds reported in literature. Here, applying a novel low-energy argon-sputter depth profiling method with significantly improved depth resolution for tungsten (W) surfaces exposed to deuterium (D) plasma at 300 K, we show the existence of a 10 nm thick D-supersaturated surface layer (DSSL) with an unexpectedly high D concentration of ~10 at.% after irradiation with ion energy of 215 eV. Electron back-scatter diffraction reveals that the W lattice within this DSSL is highly distorted, thus strongly blurring the Kikuchi pattern. We explain this strong damage by the synergistic interaction of energetic D ions and solute D atoms with the W lattice. Solute D atoms prevent the recombination of vacancies with interstitial W atoms, which are produced by collisions of energetic D ions with W lattice atoms (Frenkel pairs). This proposed damaging mechanism could also be active on other hydrogen-irradiated metal surfaces. The present work provides deep insight into hydrogen-induced lattice distortion at plasma–metal interfaces and sheds light on its modelling work. [ABSTRACT FROM AUTHOR]

Magnetron-sputtered tungsten nitride (WNx) films were used as a model system to study the behaviour of re-deposited WNx layers which could form in fusion devices with tungsten (W) wall during nitrogen seeding. The interaction of such WNx layers with deuterium (D) plasmas was investigated in dedicated laboratory experiments. D retention and N removal due to D plasma exposure (D flux: 9.9  ×  1019 D m−2 s−1, ion energy 215 eV) at different temperatures were measured with ion beam analysis (IBA). Low-energy argon sputtering followed by IBA was applied to resolve the D distribution in the top-most surface of WNx with significantly improved depth resolution compared with the standard D depth profiling method by nuclear reaction analysis. Experimentally determined thicknesses for the penetration of D in WNx were compared with the penetration depth for D calculated in SDTrimSP simulations. Results show that D is only retained within the ion penetration range for samples exposed at 300 K. In contrast to the 300 K case, D diffuses beyond the implantation depth in a sample exposed at 600 K. However, the D penetration depth is much lower than in pure W at comparable conditions. The total amount of retained D in WNx at 600 K is by 50% lower than for implantation at 300 K with the same D fluence. Nitrogen is removed only within the D ion range. [ABSTRACT FROM AUTHOR]