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The influence of defects induced by plastic deformation of tungsten (W) on deuterium (D) retention has been studied. Recrystallized W samples were subjected to tensile deformations at temperatures of 573 K and 873 K to strains in the range of 3%–38%. The dislocation density measured by transmission electron microscopy increased by about 40 times after deformation to the highest strain. The introduced defects were decorated with D by exposure to a low-flux D plasma at sample temperatures of 370 K and 450 K. D retention in the samples was studied using nuclear reaction analysis and thermal desorption spectroscopy. The trapped D concentrations after the plasma exposures were low (up to a few times 10−4 at. fr.) and increased more slowly with strain than the dislocation density. Small vacancy-like defects and large vacancy clusters were detected in the samples by positron annihilation lifetime spectroscopy. Their concentrations also increased with strain more weakly than the dislocation density. It was concluded that these defects governed the D retention under plasma exposure at 450 K, while dislocations gave only a small contribution. It was also found that deformation already to the lowest strains significantly facilitates the formation of blister-like structures under D plasma exposure at 370 K. The defects associated with blister-like structures presumably gave a substantial contribution to D retention at 370 K. [ABSTRACT FROM AUTHOR]

Hydrogen-induced cracking (HiC) or blistering is a commonly observed feature in plasma-loaded material surfaces. HiC exhibits a strong dependence on the irradiation parameters, such as incident flux and fluence, particle energy, and sample temperature. However, the underlying physics of this process is still not understood. Focusing on HiC with intragranular cavities in tungsten (W) exposed to deuterium (D) plasma, we apply a one-dimensional (1D) flux-balance model and further propose the crucial role of the solute D distribution in the subsurface region for initiating HiC formation in plasma-loaded surfaces. Within this proposal, HiC features previously observed only under high-flux (~1024 D m−2 s−1), elevated-temperature (~500 K) exposure conditions—the coexistence of protrusions with intragranular cavities and small-sized, dome-shaped blisters with intergranular cracking at the surfaces—were reproduced in our low-flux experiments (~1020 D m−2 s−1) by loading W samples at low sample temperatures (230 K). The presence of protrusions in low-flux experiments is attributed to the comparable local solute D distribution in the corresponding blistering-relevant depth in both types of D plasma exposure. Applying the 1D flux-balance model to the interpretation of HiC formation in plasma-loaded surfaces, the present work allows us to further explore the underlying physics of HiC formation under well-defined experimental conditions. [ABSTRACT FROM AUTHOR]

Repetitive melting of tungsten by power transients originating from edge localized modes (ELMs) has been studied in ASDEX Upgrade. Tungsten samples were exposed to H-mode discharges at the outer divertor target plate using the divertor manipulator II (DIM-II) system (Herrmann et al 2015 Fusion Eng. Des. 98–9 1496–9). Designed as near replicas of the geometries used also in separate experiments on the JET tokamak (Coenen et al 2015 J. Nucl. Mater. 463 78–84; Coenen et al 2015 Nucl. Fusion55 023010; Matthews et al 2016 Phys. Scr. T167 7), the samples featured a misaligned leading edge and a sloped ridge respectively. Both structures protrude above the default target plate surface thus receiving an increased fraction of the parallel power flux. Transient melting by ELMs was induced by moving the outer strike point to the sample location. The temporal evolution of the measured current flow from the samples to vessel potential confirmed transient melting. Current magnitude and dependency from surface temperature provided strong evidence for thermionic electron emission as main origin of the replacement current driving the melt motion. The different melt patterns observed after exposures at the two sample geometries support the thermionic electron emission model used in the MEMOS melt motion code, which assumes a strong decrease of the thermionic net current at shallow magnetic field to surface angles (Pitts et al 2017 Nucl. Mater. Energy12 60–74). Post exposure ex situ analysis of the retrieved samples show recrystallization of tungsten at the exposed surface areas to a depth of up to several mm. The melt layer transport to less exposed surface areas leads to ratcheting pile up of re-solidified debris with zonal growth extending from the already enlarged grains at the surface. [ABSTRACT FROM AUTHOR]

In fusion reactors, surfaces of plasma facing components (PFCs) are exposed to high heat and particle flux. Tungsten and Copper alloys are primary candidates for plasma facing materials (PFMs) and coolant tube materials, respectively, mainly due to high thermal conductivity and, in the case of tungsten, its high melting point. In this paper, recent understandings and future issues on responses of tungsten and Cu alloys to fusion environments (high particle flux (including T and He), high heat flux, and high neutron doses) are reviewed. This review paper includes; Tritium retention in tungsten (K. Schmid and M. Balden), Impact of stationary and transient heat loads on tungsten (J.W. Coenen and Th. Loewenhoff), Helium effects on surface morphology of tungsten (Y. Ueda and A. Ito), Neutron radiation effects in tungsten (A. Hasegawa), and Copper and copper alloys development for high heat flux components (C. Hardie, M. Porton, and M. Gilbert). [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]

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]

Deuterium (D) retention and surface modifications of hot-rolled polycrystalline tungsten (W) exposed to a low-energy (~40 eV D−1), high-flux (2–5  ×  1023 D m−2 s−1) D plasma at temperatures of ~380 K and ~1140 K to fluences up to 1.2  ×  1028 D m−2 have been examined by using nuclear reaction analysis, thermal desorption spectroscopy, and scanning electron microscopy. The samples exposed at ~380 K exhibited various types of surface modifications: dome-shaped blister-like structures, stepped flat-topped protrusions, and various types of nanostructures. It was observed that a large fraction of the surface was covered with blisters and protrusions, but their average size and the number density showed almost no fluence dependence. The D depth distributions and total D inventories also barely changed with increasing fluence at ~380 K. A substantial amount of D was retained in the subsurface region, and thickness correlated with the depth where the cavities of blisters and protrusions were located. It is therefore suggested that defects appearing during creation of blisters and protrusions govern the D trapping in the investigated fluence range. In addition, a large number of small cracks was observed on the exposed surfaces, which can serve as fast D release channels towards the surface, resulting in a reduction of the effective D influx into the W bulk. On the samples exposed at ~1140 K no blisters and protrusions were found. However, wave-like and faceted terrace-like structures were formed instead. The concentrations of trapped D were very low (<10−5 at. fr.) after the exposure at ~1140 K. [ABSTRACT FROM AUTHOR]

In this paper, reduced-activation ferritic/martensitic (RAFM) steels including Eurofer (9Cr) and oxide dispersion strengthening (ODS) steels by the addition of Y2O3 particles with different amounts of Cr, namely, (9-16)Cr were exposed to low energy deuterium (D) plasma (~20–200 eV per D) up to a fluence of 2.9  ×  1025 D m−2 in the temperature range from 290 K to 700 K. The depth profile of D in steels was measured up to 8 µm depth by nuclear reaction analysis (NRA) and the total retained amount of D in those materials was determined by thermal desorption spectroscopy (TDS). It was found that the D retention in ODS steels is higher compared to Eurofer due to the much higher density of fine dispersoids and finer grain size. This work shows that in addition to the sintering temperature and time, the type, size and concentration of the doping particles have an enormous effect on the increase in the D retention. The D retention in undamaged ODS steels strongly depends on the Cr content: ODS with 12Cr has a minimum and the D retention in the case of ODS with (14-16)Cr is higher compared to (9-12)Cr. The replacing of Ti by Al in ODS-14Cr steels reduces the D retention. The formation of nano-structure surface roughness enriched in W or Ta due to combination of preferential sputtering of light elements and radiation-induced segregation was observed at incident D ion energy of 200 eV for both Eurofer and ODS steels. Both the surface roughness and the eroded layer enhance with increasing the temperature. The surface modifications result in a reduction of the D retention near the surface due to increasing the desorption flux and can reduce the overall D retention. [ABSTRACT FROM AUTHOR]

The reduced-activation ferritic/martensitic (RAFM) steels including Eurofer (9Cr) and oxide dispersion strengthened (ODS) steels by the addition of Y2O3 particles investigated in Part I were pre-damaged either with 20 MeV W ions at room temperature at IPP (Garching) or with high heat flux at FZJ (Juelich) and subsequently exposed to low energy (~20–200 eV per D) deuterium (D) plasma up to a fluence of 2.9  ×  1025 D m−2 in the temperature range from 290 K to 700 K. The pre-irradiation with 20 MeV W ions at room temperature up to 1 displacement per atom (dpa) has no noticeable influence on the steel surface morphology before and after the D plasma exposure. The pre-irradiation with W ions leads to the same concentration of deuterium in all kinds of investigated steels, regardless of the presence of nanoparticles and Cr content. It was found that (i) both kinds of irradiation with W ions and high heat flux increase the D retention in steels compared to undamaged steels and (ii) the D retention in both pre-damaged and undamaged steels decreases with a formation of surface roughness under the irradiation of steels with deuterium ions with incident energy which exceeds the threshold of sputtering. The increase in the D retention in RAFM steels pre-damaged either with W ions (damage up to ~3 µm) or high heat flux (damage up to ~10 µm) diminishes with increasing the temperature. It is important to mention that the near surface modifications caused by either implantation of high energy ions or a high heat flux load, significantly affect the total D retention at low temperatures or low fluences but have a negligible impact on the total D retention at elevated temperatures and high fluences because, in these cases, the D retention is mainly determined by bulk diffusion. [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]

ASDEX Upgrade became a full tungsten experiment in 2007 by coating its graphite plasma facing components with tungsten. In 2013 a redesigned solid tungsten divertor, Div-III, was installed and came into operation in 2014. The redesign of the outer divertor geometry provided the opportunity to increase the pumping efficiency in the lower divertor by increasing the gap between divertor and vessel. In parallel, a by-pass was installed into the cryo-pump in the divertor region allowing adapting of the pumping speed to the required edge density.Safe divertor operation and heat removal becomes more and more significant for future fusion devices. This requires developing ‘tools’ for divertor heat load control and to optimize the divertor design. The new divertor manipulator, DIM-II, allows retracting a relevant part of the outer divertor into a target exchange box without venting ASDEX Upgrade. Different front-ends can be installed and exposed to the plasma. At present, front-ends for probe exposition, gas puffing, electrical probes and actively cooled prototype targets are under construction.The installation of solid tungsten, the control of the pumping speed and the flexibility for testing divertor modifications on a weekly base is a unique feature of ASDEX Upgrade and offers together with the extended set of diagnostics the possibility to investigate dedicated questions for a future divertor design. [ABSTRACT FROM AUTHOR]

Blistering of tungsten (W) surfaces due to deuterium (D) implantation was investigated by a sequence of exposures at two different temperatures—230 and 450 K—and by the reversed sequence. A single exposure at 230 K leads to a high areal density of small dome-shaped blisters (up to 3 µm in diameter) together with much smaller flat-topped structures, while a single 450-K exposure produces large dome-shaped blisters up to 40 µm in diameter without the flat-topped structures. Most of the small dome-shaped blisters from 230 K exposure disappeared after annealing at 450 K for 17 h, but survived and even grew in size if the surface was exposed to D plasma during annealing. Sequential exposure at the two temperatures reveals a non-commutative behaviour: after a first exposure at 450 K the second exposure at 230 K leads to superposition of the observed blister structures without changing the large blisters from the first exposure. By contrast, a first exposure at 230 K almost completely suppresses the formation of large blisters during a second exposure at 450 K. Obviously, the presence of the small blisters strongly influences the penetration of D into the W bulk. [ABSTRACT FROM AUTHOR]