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An atomistic description of tin deposition on ruthenium and its effect on blistering damage is of great interest in extreme ultraviolet (EUV) lithography. In EUV machines, tin debris from the EUV-emitting tin plasma may be deposited on the mirrors in the optical path. Tin facilitates the formation of hydrogen-filled blisters under the ruthenium top layer of the multi-layer mirrors. We have used Density Functional Theory (DFT) to show that tin deposition on a clean ruthenium surface exhibits a film-plus-islands (Stranski–Krastanov) growth mode, with the first atomic layer bonding strongly to the substrate. We find that a single tin layer allows hydrogen to reach the ruthenium surface and subsurface more easily than on clean ruthenium, but hydrogen penetration through the tin film becomes progressively more difficult when more layers are added. The results indicate that hydrogen penetration and blistering occur when only a thin layer of tin is present., A monolayer of Sn makes the Ru surface more permeable to H, but more layers inhibit H penetration.

Self-passivating Metal Alloys with Reduced Thermo-oxidation (SMART) are under development for the primary application as plasma-facing materials for the first wall in a fusion DEMOnstration power plant (DEMO). SMART materials must combine suppressed oxidation in case of an accident and an acceptable plasma performance during the regular operation of the future power plant. Modern SMART materials contain chromium as a passivating element, yttrium as an active element and a tungsten base matrix. An overview of the research and development program on SMART materials is presented and all major areas of the structured R&D are explained. Attaining desired performance under accident and regular plasma conditions are vital elements of an R&D program addressing the viability of the entire concept. An impressive more than 104-fold suppression of oxidation, accompanied with more than 40-fold suppression of sublimation of tungsten oxide, was attained during an experimentally reproduced accident event with a duration of 10 days. The sputtering resistance under DEMO-relevant plasma conditions of SMART materials and pure tungsten was identical for conditions corresponding to nearly 20 days of continuous DEMO operation. Fundamental understanding of physics processes undergone in the SMART material is gained via fundamental studies comprising dedicated modeling and experiments. The important role of yttrium, stabilizing the SMART alloy microstructure and improving self-passivating behavior, is under investigation. Activities toward industrial up-scale have begun, comprising the first mechanical alloying with an industrial partner and the sintering of a bulk SMART alloy sample with dimensions of 100 mm × 100 mm × 7 mm using an industrial facility. These achievements open the way to further expansion of the SMART technology toward its application in fusion and potentially in other renewable energy sources such as concentrated solar power stations.

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 × 1024 m−2 s−1) and high-cycle (104–105) 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 s0.5 (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.

It is highly desirable to understand the long term evolution of the divertor material under the extreme steady-state and transient heat and particle loads expected during ITER operation. Here the impact of ELM-like transient loading under combined high-flux plasma and transient ELM-like heat loading in Magnum-PSI was explored to determine how plasma affects the fatigue cracking threshold of tungsten due to ELMs. Mock-ups consisting of five ITER-like monoblocks in a chain were simultaneously exposed to high flux plasma and a high power pulsed laser which closely simulated the ELM impact in terms of heat flux and duration. Loading conditions were chosen to enable comparison to existing data from electron-beam loading, while the influence of surface base temperature (750 °C, 1150 °C or 1500 °C) and impurity seeding (addition of 6.5 ion% He+ and/or 8 ion% Ne+) were also investigated. The plasma loading leads to differences in surface morphology and indicates synergistic effects on the extent of the surface damage. Base temperatures at or above 1150 °C are found to lead to a significant reduction in the fatigue cracking threshold by a factor of two or more compared to at 750 °C. Cracked surfaces are found to be more than ten times rougher than the original microstructure, and additionally when seeding impurities are added surface roughness can be significantly increased by up closely factor of two compared to roughening using pure H plasma. Overall the results indicate that avoiding fatigue cracking in ITER will be very challenging, and that understanding the level to which this can therefore be tolerated is vital for anticipating divertor lifetime and reliability.

One of the issues faced by future fusion devices will be high divertor target heat loads. Alternative divertors can promote detachment, flux expansion and dissipation mechanisms to mitigate these heat loads. They have been investigated in several devices including TCV and DIII-D, and will be investigated on MAST-U. To evaluate their effectiveness, accurate target heat flux and power balance measurements are required in these machines. Infrared (IR) thermography is a widely used technique to determine the target heat flux, but is susceptible to surface effects and emissivity in carbon-walled machines. In this work, the effect of plasma exposure on graphite is assessed to understand what may happen in MAST-U. A sample of fine grain graphite, as used on MAST-U, is exposed to 30min plasma exposures, with density ne=6 × 1018m−3 and temperature Te=0.08eV as measured by Thomson scattering. During these pulses, the temperature is measured by a medium wave IR camera and is seen to decrease by ≈70 °C over the course of 3h of plasma exposure. Pyrometer measurements suggest that the IR camera data is affected by a change in the surface emissivity. Profilometry confirms erosion of graphite at the tile centre to a depth of ≈100µm, and a larger region of deposition further out, amounting to ≈40µm of material.

Secondary electron emission (SEE) yields of tin (Sn) and tin-lithium (SnLi) eutectic (20 at.% Li) samples were measured in He-plasma at a mean incoming electron energy up to 120 eV. SnLi shows a maximum yield of about 1.45 at 110 eV electron energy while the yield of the Sn surface was measured to be maximally 1.05 at 120 eV. X-ray photoelectron spectroscopy (XPS) analysis demonstrated the segregation effect of Li to the surface of the eutectic, both after melting in the argon atmosphere and in molten state with simultaneous He-plasma exposure. At least the top 10 nm of the SnLi samples were heavily enriched with Li, and Sn/Li ratios varied in the range 0.8–5% depending on eutectic treatment conditions. After the plasma exposure Sn3d is detected predominantly in the oxidized state while after extended atmospheric oxidation there was still a significant amount of Sn3d detected in the metallic state. The liquid surface of SnLi indicated a possible decrease of SEE yield. All measurements gave values of SEE yield close to or above unity. Such values can lead to significant plasma sheath disturbances and subsequent additional heat flux from electrons on such a plasma-facing material, thus, should be accounted for in designing fusion reactors using these components.

Hydrogen-graphite interactions are relevant to a wide variety of applications, ranging from astrophysics to fusion devices and nano-electronics. In order to shed light on these interactions, atomistic simulation using Molecular Dynamics (MD) has been shown to be an invaluable tool. It suffers, however, from severe time-scale limitations. In this work we apply the recently developed Collective Variable-Driven Hyperdynamics (CVHD) method to hydrogen etching of graphite for varying inter-impact times up to a realistic value of 1 ms, which corresponds to a flux of ∼1020 m-2 s-1. The results show that the erosion yield, hydrogen surface coverage and species distribution are significantly affected by the time between impacts. This can be explained by the higher probability of C-C bond breaking due to the prolonged exposure to thermal stress and the subsequent transition from ion- to thermal-induced etching. This latter regime of thermal-induced etching-chemical erosion-is here accessed for the first time using atomistic simulations. In conclusion, this study demonstrates that accounting for long time-scales significantly affects ion bombardment simulations and should not be neglected in a wide range of conditions, in contrast to what is typically assumed.

The key remaining physics design issue for the ITER tungsten (W) divertor is the question of monoblock (MB) front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM) or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA) and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets. Keywords: ITER, Tungsten, Divertor, Shaping, Melting, MEMOS

Tungsten samples were heat-treated to achieve partial recrystallization and exposed to high ion flux deuterium plasma at different temperatures and fluences. Continuous stiffness nanoindentation measurements of near-surface hardness were performed in the grains of specific annealing states and of specific crystallographic orientation, determined by electron backscatter diffraction (EBSD); indentation pile-up was investigated using surface profilometry. Bulk hardness of unexposed tungsten does not strongly depend on grain orientation, but depends on the annealing state of the grain, with values between ∼4.3 GPa for recrystallized grains and ∼5.5 for non-recrystallized ones. Grains with surface normal orientation feature the least pile-up, while grains with orientation the most; pile-up also depends on the annealing state, being generally lower in recrystallized grains. Plasma exposure leads to the increase of hardness, most significantly near the surface. The width of plasma-affected zone increases with the increase of exposure temperature and fluence, as well in recrystallized grains, correlating with the increase of diffusion depth. Plasma exposure does not lead to the emergence of orientation-dependence of hardness. Both indentation pile-up and near-surface indentation pop-ins are generally suppressed by plasma exposure.

A small-scale tungsten monoblock mockup was exposed to continuous Magnum-PSI plasma beams under six different conditions, including pure hydrogen, deuterium and helium (He) plasmas as well as mixed D/He plasmas with low electron temperature (1.5–5 eV), with maximum surface temperatures up to 1600 °C, and with ion fluences up to 10 30 m−2. The recrystallization in the centre of the hottest exposure spots, the absence of cracking on all monoblocks, the formation of a nanostructured fuzz layer several microns thick by pure He exposure, and the presence of various impurities from the exposure in Magnum-PSI on the surface were observed and analysed with a dedicated scanning electron microscope able to handle the entire mockup without any prior cutting.

W–Y2O3 alloy with low DBTT (ductile-brittle transition temperature) and high RCT (recrystallization temperature), processed by high energy rate forging (HERF) was exposed to ITER-like steady-state and transient hydrogen plasma in the linear plasma generator Magnum-PSI. The steady-state heat fluxes were in the range of 8.35–16.32 MW . m−2, resulting in surface base temperatures of the samples in the range of 1271 °C to 1982 °C. The applied transient peak heat flux with a frequency of 5 Hz (a total of 1000 pulses) was about 0.50 GW . m−2. The exposure time was ~220 s. No obvious morphological change of the exposed sample with a base temperature of 1271 °C was observed, except the preferential erosion of the W/Y2O3 interface. However, cracks along grain boundaries were formed on the surface of the exposed samples with base temperatures above 1389 °C. Pronounced recrystallization and grain growth also occurred for the samples with base temperatures of 1666 °C and 1982 °C. It is desirable to find that no wide and deep crack with preferential propagation direction with regard to the sample dimension was observed, even extensive recrystallization and cracking along grain boundaries occurred. This indicates that decreasing DBTT and increasing RCT simultaneously is desirable to broaden the safe operational temperature window of W as plasma-facing material, and therefore to increase the power handling capability of the plasma-facing units of a tungsten-based divertor in future fusion reactors. However, the formation of W/Y2O3 composite, melting and depletion of the Y2O3 particles gradually occurred with increasing the surface temperatures. It implies that the doping of Y2O3 particles complicates the plasma-material interaction, compared to the pure W case. For example, it raises concerns about the formation of complex dust which will potentially be a significant issue for the safe operation of future fusion reactors (e.g. core plasma contamination and fuel recycling).

Due to a lack of direct experimental results, the detailed mechanisms that govern the blistering behavior of tungsten (W) exposed to ITER-relevant condition in nuclear fusion remain unclear. The growth mechanism of hydrogen (H) blisters is one example. In this work, recrystallized W was exposed to H plasma at 50 eV, 1.5 × 10 26 m − 2 , and 573 K. Transmission electron microscopy (TEM) samples were prepared using plasma-focused ion beam (FIB) followed by flash-polishing to effectively remove surface damages induced by FIB. The TEM images revealed that the general blisters observed on the exposed surface are associated with underlying cavities. A considerable amount of dislocations were found in the vicinity of the cavities. Prismatic dislocation loop arrays were observed, including small size 'coffee-bean' prismatic loops and large size prismatic loops. Near the tip of surfaces cavities, evidences for the emission of shear loops were also found. Based on the experimental findings, a multi-stage growth mechanism of H cavities was proposed. The loop-punching mechanism is operative for both very small cavities and cavities with sizes larger than several hundreds of nanometers. Whereas at intermediate sizes, cavities grow by emitting shear loops from the cavity tip.

The ITER divertor will be exposed to extremely high plasma fluences over its lifetime, and it is known that plasma exposure can lead to a variety of particle-induced surface-morphology and microstructure changes in tungsten. However, no data exists at fluences comparable to those expected over extended ITER operations (10 30−31 m−2) and so it is uncertain how these changes will evolve and affect the divertor performance over such long timescales. Six monoblocks were exposed to high flux plasma comparable to partially-detached plasma conditions in the ITER divertor in Magnum-PSI. Different exposures used different plasma species (H, He, D or D + He) and aimed to replicate conditions similar to those during different phases of the ITER staged approach. The highest fluence achieved was 10 30 D m−2, comparable to around one year of ITER Fusion Power Operation. Post-mortem analysis by Nuclear Reaction Analysis revealed very low deuterium retention throughout the blocks, while surface analysis showed no cracking or damage, but did observe helium fuzz growth at low ion energies of 8–18 eV, below typically assumed ion energy requirements for such growth to occur. Metallographic sectioning revealed recrystallization up to 2.2 mm below the surface of monoblocks exposed at peak surface temperatures of up to 1580 °C for different durations up to ~20 h. Finite Element Method analysis coupled to metallographic and Vickers Hardness identification of the boundary of the recrystallized region identified a faster recrystallization process compared to literature expectations, reinforcing that recrystallization dynamics is an important criterion for tungsten grade selection for the ITER divertor. Overall, no major damage or failure was identified, indicating that the design is capable of fulfilling its steady-state performance requirements under high flux, high fluence plasma loading conditions in the ITER divertor.

To develop realistic liquid lithium divertors for future fusion reactors, this paper aims to improve the understanding of their power handling capabilities. A liquid lithium divertor target prototype, designed to facilitate liquid metal experiments in tokamaks, was tested in Magnum-PSI. The target has an internal reservoir pre-filled with lithium and aims to passively re-supply the textured plasma facing surface during operation. To assess the power handling capability the target was exposed to helium plasmas with increasing power flux density in the linear plasma device Magnum-PSI. The temperature response of lithium targets was recorded via an infrared camera, and compared to finite element method modeling taking into account dissipation via lithium in the plasma. It was found that the target works as intended and can take up to 9 1 MW m−2 for 10 s before the mesh layer was damaged, and could continue operating at higher power densities even after being damaged. The total lifetime of the targets was up to 100 s. Overall the targets are found suitable for use in tokamak experiments. Additionally, a central surface temperature evolution indicative of vapor shielding was observed on intact targets. Predicting the target temperature (and consequently the evaporation rates and thermal stresses) is considered very relevant for the design of lithium divertor targets for DEMO. The observed temperature response could indeed be replicated through modeling, which showed that a significant power fraction was dissipated by the lithium in the plasma.

Liquid metal divertors aim to provide a more robust alternative to conventional tungsten divertors. However, they still require a solid substrate to confine the liquid metal. This work proposes a novel design philosophy for liquid metal divertor targets, which allows for a two orders of magnitude reduction of thermal stresses compared to the state-of-the-art monoblock designs. The main principle is based on a 3D-printed tungsten structure, which has low connectedness in the direction perpendicular to the thermal gradient, and as a result also short length scales. This allows for thermal expansion. Voids in the structure are filled with liquid lithium which can conduct heat and reduce the surface temperature via vapor shielding, further suppressing thermal stresses. To demonstrate the effectiveness of this design strategy, an existing liquid metal concept is re-designed, fabricated, and tested on the linear plasma device Magnum-PSI. The thermo-mechanical finite element method analysis of the improved design matches the temperature response during the experiments, and indicates that thermal stresses are two orders of magnitude lower than in the conventional monoblock designs. The relaxation of the strength requirement allows for much larger failure margins and consequently for many new design possibilities.

From the helium (He) plasma irradiations to tungsten performed in the Magnum-PSI device, the effects of deposition of metals on the helium-plasma induced fiberform nanostructures (fuzz) are discussed. It was found that fuzz was not formed at the center of the plasma cylinder if there were significant metallic impurities from the source. Deposition of metallic impurities (mainly molybdenum and copper) counteracted the growth of fuzz. In addition to the effects of metals from the source, we installed a sputtering source near the sample to replicate the deposition environment in fusion devices. The thickness of fuzzy layer was ∼ 7 µm, which was about five times greater than that without deposition, at the He flux of 1.3 × 10 26 m − 2 , suggesting that the growth rate of fuzz layer was significantly accelerated due to the deposition of tungsten.

A collective Thomson scattering system has been developed for measuring ion temperature, plasma velocity and impurity concentration in the high density magnetized Magnum-PSI plasma beam, allowing for measurements at low temperature (4 × 1020 m−3, while avoiding laser plasma heating caused by inverse Bremsstrahlung. The collective Thomson scattering system is based on the fundamental mode of a seeded Nd:YAG laser and equipped with an LIVAR M506 camera (EBABS technology). The first collective Thomson scattering measurements are taken at the linear plasma generator Pilot-PSI, 40 mm downstream of the cascaded arc source. At this location, the ion temperature is about equal to the electron temperature in the bulk of the plasma beam.

The InfraRed Video Bolometer (IRVB) is a powerful tool to measure radiated power in magnetically confined plasmas due to its ability to obtain 2D images of plasma emission using a technique that is compatible with the fusion nuclear environment. A prototype IRVB has been developed and installed on NSTX-U to view the lower divertor. The IRVB is a pinhole camera which images radiation from the plasma onto a 2.5 μm thick, 9 × 7 cm2 Pt foil and monitors the resulting spatio-temporal temperature evolution using an IR camera. The power flux incident on the foil is calculated by solving the 2D+time heat diffusion equation, using the foil’s calibrated thermal properties. An optimized, high frame rate IRVB, is quantitatively compared to results from a resistive bolometer on the bench using a modulated 405 nm laser beam with variable power density and square wave modulation from 0.2 Hz to 250 Hz. The design of the NSTX-U system and benchtop characterization are presented where signal-to-noise ratios are assessed using three different IR cameras: FLIR A655sc, FLIR A6751sc, and SBF-161. The sensitivity of the IRVB equipped with the SBF-161 camera is found to be high enough to measure radiation features in the NSTX-U lower divertor as estimated using SOLPS modeling. The optimized IRVB has a frame rate up to 50 Hz, high enough to distinguish radiation during edge-localized-modes (ELMs) from that between ELMs.

The orientation dependence of blister formation induced by D plasma exposure at low temperature (about 523 K) on rolled tungsten and chemical vapor deposition (CVD) W samples was studied by scanning electron microscopy and electron backscatter diffraction. Severe blistering was observed on grains with surface normal directions close to [111], while the [001] surfaces are the most resistant to blister formation. Cavities induced by D2 gas were observed beneath [111], [110] and [001] surfaces, independently on whether blisters were observed on the surface or not. The [111] surface is more prone to blister formation, because it is easily plastically deformed by the D2 gas pressure. Some blister edges and steps were perpendicular to [110] directions, which may be induced by the slipping of dislocations on {110} planes. The blister morphology induced by D plasma can be well explained by the blister model based on plastic deformation mechanism.

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.

The first wall is one of the key components in a tokamak, in particular, for DEMO in the future. The conceptual first wall for a DEMO reactor is designed, fabricated and tested in this paper. It has a sandwich-like structure, which consists of a reduced activation ferritic/martensitic (RAFM) substrate, an interlayer and a plasma-facing material of tungsten produced by chemical vapor deposition (CVD-W) due to its characteristics of high density, high purity and superior thermal shock resistance. The interlayer is required to not only have good adhesion between CVD-W and the CLF-1 substrate, which is a type of RAFM developed by the Southwestern Institute of Physics, but also the capability of tritium transport prevention due to the great importance of controlling the tritium buildup in the first wall, improving the fuel efficiency and conforming to the safety regulations of tritium. Titanium nitride (TiN), a well-known tritium barrier, is selected as the interlayer material in this work. In total, eight sandwich-like first wall samples have been fabricated, and thermal cycle tests and plasma exposure tests are implemented. The results of material analysis show that the quality of the TiN coating layer is critical for good adhesion and tritium prevention. At the defects of the TiN interlayer, fluorine (F) penetrated into the CLF-1 during the CVD-W fabrication and, as a result, the sandwich-like structure was broken near the CLF-1 side. For some samples, the TiN interlayer made by CVD has sufficient adhesion as an interlayer between CVD-W and CLF-1 during tests so far. These encouraging test results support further examination of TiN as an interlayer material for W on an RAFM steel as a first wall material. Preliminary results indicate that the design and fabrication of the conceptual first wall are feasible. Further tests, including the improvement of the interlayer quality and the tritium permeation test, are necessary in the future.

The surface nano structures on W surface exposed to low-energy D plasma was studied by transmission electron microscopy. It was found that the surface nano structures on W surface induced by D plasma were mainly due to two reasons. First, the subsurface nano-sized bubbles will induce nano scale blistering of the surface, resulting in surface nano morphology. The nano scale blistering behaviour is most obvious on [1 1 1] surface at low temperature (500 K). Secondly, during the plasma exposure, oxide layer on W surface will be sputtered by low energy D particles, and the sputtering can also induce nano structures on some surfaces. At low temperature (500 K), the surface nano structures were formed due to the nano scale blistering behaviour and the sputtering of surface oxide layer. At high temperature (1000 K), the surface nano structures were mainly caused by the sputtering of surface oxide layer.

The authors exposed a radiatively cooled, Li-filled tantalum (Ta) heat pipe (HP) to a H plasma in Magnum PSI continuously for ˜2 h. We kept the overall heat load on the inclined HP constant and varied the tilt to give peak heat fluxes of ˜7.5–13 MW/m2. The peak temperature reached ˜1250 °C. This paper describes the post-test analysis and discusses Li HPs with materials other than Ta for fusion. A companion paper describes the experiment.

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.

Tungsten (W) substrates with adhered beryllium (Be) proxy dust - copper, chromium, aluminium - have been exposed in the Magnum-PSI linear device. Their interaction with transient and stationary plasmas has been systematically studied under varying heat fluxes and magnetic field topologies. The dust remobilization activities, macro-morphological changes and chemical modifications induced by the plasma incidence are documented. Aluminium is identified to be the most suitable surrogate material due to the similar binary phase diagram and nearly identical evaporation rates. Extrapolation suggests that Be dust cannot survive on hot W surfaces but it can trigger mixed Be/W effects prior to its plasma removal. Keywords: Transient heat loads, Dust remobilization, Dust survivability, Dust adhesion, Mixed material effects

Both the physics of divertor detachment and vapour shielding are characterized by a relatively large amount of radiation produced in the divertor. The linear plasma generator Magnum-PSI is well-suited to study such processes due to its ITER-divertor relevant plasma conditions, simplified geometry and diagnostic accessibility. The need the quantify the plasma radiated power close to the target surface motivated the development of a 4-channel resistive bolometer for Magnum-PSI, and marks the first deployment of such a diagnostic on a linear device. An axially resolved measurement of plasma emission at arbitrary distances from the target surface is now possible. The radial position of the detector can be varied, hereby viewing the full diameter of the plasma column or down to a central region. The overall system design is discussed alongside a comparison of the spectral absorbance of carbon-coated versus non-coated Au/Al bolometer sensors. Despite low electron temperatures of the plasma (1-5 eV), the observed power densities were found to be 10-37 times the sensor noise floor of ∼0.1 W m-2. A synthetic diagnostic based on collisional radiative model calculations from ADAS could well match observed values from H and Ne plasmas while the measured values for Ar and He were more difficult to reproduce. The obtained findings allow for approximate power balance calculations in Magnum-PSI indicating that maximally ∼47% and ∼14% of the total power is lost by radiation in the cases of Ar and Ne/He respectively. The results demonstrate the feasibility of resistive bolometry in low temperature high density plasma regions and on long timescales (>450 s) which is of relevance to ITER. Due to long-term temperature drifts which were observed, a recent upgrade involved the installation of a shutter and FPGA-based electronics for increased accuracy.

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: 1023–1025 m−2s−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 2.5 T, 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.

A model is formulated to make a first estimate of the maximum tolerable power of liquid lithium (Li) divertor targets, and to gain insight into their behaviour in terms of Li loss rate and surface temperature. The model, formulated as a simple analytical expression, states that the incoming power is balanced by heat conduction through the target and by the Li which dissipates energy via evaporation, radiation and ion-neutral friction. A target is considered to fail when the net Li loss flux from the surface exceeds the available supply. The model is evaluated over a range of input parameters: Li supply rate, surface layer thickness, redeposition coefficient, and dissipated energy per Li particle lost to the plasma. Based on the results, first, surface temperature locking is expected above a deposited power of ∼10 MW m-2. Second, Li targets are expected to be extremely robust against power deposited during short transient events. A surface layer thickness of 50 micron is sufficient to withstand 60 MJ m-2 vertical displacement events or 20 MJ m-2 disruptions.

For the final design of the ITER divertor it is important to determine whether shaping of each tungsten monoblock to eliminate leading edges is required or not. In order to aid this decision, two experiments were performed in DIFFERs linear plasma devices to study heat loads on misaligned water cooled blocks at glancing incidence. First, a series of tungsten blocks were exposed to a high parallel heat flux (26 MW m-2) hydrogen plasma beam in the Magnum-PSI linear device. The blocks were exposed at an oblique angle between 4° and 7° with respect to the plasma beam, approaching the low angle of incidence expected in the ITER divertor strike-point regions. One block was vertically misaligned with respect to the others by a value between 0 and 1.2 mm, and the surface temperature evolution monitored. This was compared to a finite element thermo-mechanical model of the system, to investigate any discrepancies between the two. The importance of Larmor orbit smoothing effects could also be considered, as the Larmor radius was larger than half the misalignment height. Next, a row of identical tungsten blocks were exposed at an angle of 3° to ∼ 500-700 μs pulsed hydrogen plasmas of peak power between 275-625 MW m-2 in the Pilot-PSI linear device, to simulate the effect of edge localized mode (ELM) transients. One block was misaligned with respect to the others by 1 mm. In all cases, the model and experiment were found to be in excellent agreement (to ∼1%), demonstrating clearly that an optical approximation for plasma power loading on misaligned edges is appropriate. A simple model including Larmor smoothing was applied, the results of which give a similar prediction to the optical approximation for the temperature of the misaligned edges, within the error of the temperature measurements, though a small decrease in edge temperature is calculated. Therefore it can be concluded that the application of the optical approximation is generally appropriate as an input physics assumption for the study of shaping alternatives.

International audience; This study demonstrates a fast one-step synthesis method for nanostructured carbon microparticles on graphite samples using high-flux plasma exposure. These structures are considered as potential candidates for energy applications such as Li-ion batteries and supercapacitors. The samples were exposed to plasmas in the linear plasma generator Pilot-PSI with an average hydrogen ion-flux of ~10 24 m-2 s-1. The parameter window was mapped by varying the ion energy and flux, and surface temperature. The particle growth depended mainly on the sample gross-erosion and the resulting hydrocarbon concentration in the plasma. A minimum concentration was necessary to initiate particle formation. The surface of the sample was covered with microparticles with an average growth rate of 0.2 μm/s, which is significantly faster than most chemical methods. The particles were initially volumetrically grown in in the gas-phase by a multi-phase process and after deposition on the sample their growth proceeded. Scanning and transmission electron microscopy reveal that the core of these microparticles can be made of an * Corresponding author. E-mail: d.aussems@differ.nl (D.U.B. Aussems) 2 agglomeration of nanoparticles, surrounded by crumpled layers of carbon nanowalls. Gas sorption analysis shows sufficient meso-and macropores for fast mass transport. In conclusion, this processing technique could be a novel synthesis route to nanostructure surfaces for electrochemical applications.

Providing an efficacious plasma facing surface between the extreme plasma heat exhaust and the structural materials of nuclear fusion devices is a major challenge on the road to electricity production by fusion power plants. The performance of solid plasma facing surfaces may become critically reduced over time due to progressing damage accumulation. Liquid metals, however, are now gaining interest in solving the challenge of extreme heat flux hitting the reactor walls. A key advantage of liquid metals is the use of vapour shielding to reduce the plasma exhaust. Here we demonstrate that this phenomenon is oscillatory by nature. The dynamics of a Sn vapour cloud are investigated by exposing liquid Sn targets to H and He plasmas at heat fluxes greater than 5 MW m−2. The observations indicate the presence of a dynamic equilibrium between the plasma and liquid target ruled by recombinatory processes in the plasma, leading to an approximately stable surface temperature., Vapour shielding is one of the interesting mechanisms for reducing the heat load to plasma facing components in fusion reactors. Here the authors report on the observation of a dynamic equilibrium between the plasma and the divertor liquid Sn surface leading to an overall stable surface temperature.

Cavity-ring-down spectroscopy (CRDS) was implemented to measure the re-deposition of liquid tin under a high flux plasma beam in the linear plasma device Pilot-PSI. A capillary porous system (CPS) consisting of a molybdenum cup and tungsten meshes (pores diameters of 0.2 mm and 0.44 mm) was filled with tin and exposed to argon plasma. The absorption of a UV laser-beam at 286.331 nm was used to determine a number of sputtered neutral tin atoms. The incoming flux of argon ions of ∼50 eV was 1.6-2.7 × 1023 m-2 s-1, and the sample temperature measured by pyrometry varied from 850 °C to 1200 °C during exposures. The use of CRDS for measuring absolute number of particles under such plasma exposure was demonstrated for the first time. The number of sputtered tin particles in the cavity region assuming no losses would be expected to be 5.5 × 1011-1.2 × 1012 while CRDS measurements showed only 5.7-9.9 × 108. About 98-99.8% of sputtered particles were therefore found to not reach the CRDS observation volume. Spectroscopic ratios of Sn I to Sn II ions, as well as equilibrium considerations, indicate that fast ionization as well as plasma entrainment of neutrals is responsible for the discrepancy. This would lead to high re-deposition rates, implying a lowered contamination rate of core plasma and lower required replenishment rates at high-flux conditions than would otherwise be expected.

Feasibility of an ion sensitive probe was evaluated in high electron density (Formula Presented) helium plasmas produced in the Magnum-PSI device. The ion sensitive probe showed that the ion temperature was 1 eV and almost equal to the electron temperature. Increasing the neutral pressure to efficiently lead to the electron-ion recombination processes, the electron currents were collected at the ion collector. A secondary electron emission in the guard electrode might have an effect on the ion current.

We have performed a combined study using transmission electron microscopy (TEM), nuclear reaction analysis (NRA) and nano-indentation (NI) techniques to reveal the impact of high flux plasma exposure on the properties of a sub-surface region of the commercially available pure tungsten fabricated following the ITER specification. TEM examination revealed the formation of a dense dislocation network and dislocation tangles, resulting in a strong increase in the dislocation density by at least one order of magnitude as compared to the bulk density. The plasma-induced dislocation microstructure vanishes within a depth of about 10–15 μm from the top of the exposed surface. Surface hardness after the plasma exposure was characterized by NI and was found to increase significantly in the sub-surface region of 1.5–3 μm. That was attributed to the resistance of the plasma-induced dislocation networks and deuterium-induced defects, whose presence within a depth of ∼1 μm was unambiguously detected by the NRA measurements as well.

The surface damage of W under D plasma and ELMs-like transient heat loads was studied by combined stationary and pulsed D plasma. Low-flux transient heat loads will promote blister formation due to the gas expansion inside the blisters. On the contrary, high-flux transient heat loads will mitigate blistering due to the high surface temperature. Therefore, blistering on W surface first increased and then decreased with the increasing transient heat loads. The promotion effect of pulsed plasma on blistering is more obvious on [001] and [110] surfaces than on [111] surface, and the orientation dependence of blisters was mitigated by the transient heat loads. Surface modification induced by transient heat loads only formed on [001] and [110] surfaces, but did not form on [111] surface. The orientation dependence of surface modification was mainly due to the slipping system of dislocations.

Liquid metal infused capillary porous structures (CPSs) are considered as a potential divertor solution for DEMO due to their potential power handling capability and resilience to long term damage. In this work the power handling and performance of such Sn-based CPS systems is assessed both experimentally and via modelling. A Sn-CPS target was exposed to heat fluxes of up to 18.1 MW m−2 in He plasma in the Pilot-PSI linear device. Post-mortem the target showed no damage to nor any surface exposure of the underlying W-CPS felt. The small pore size (∼40µm) employed resulted in no droplet formation from the target in agreement with calculated Rayleigh-Taylor and Kelvin-Helmoholtz instability thresholds. The temperature response of the Sn-target was used to determine the thermal conductivity of the mixed Sn-CPS material using COMSOL modelling. These values were then used via further finite element analysis to extrapolate to DEMO relevant monoblock designs and estimate the maximum power handling achievable based on estimated temperature windows for all component elements of the design. For an optimized design a heat-load of up to 20 MW m−2 may be received while the use of CPS also offers other potential design advantages such as the removal of interlayer requirements. Keywords: Liquid-metal, DEMO, CPS, Tin, Power-exhaust, Plasma-wall interaction

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 × 10 23 D m −2 s −1 ) D plasma at temperatures of 380 K and 1140 K to fluences up to 1.2 × 10 28 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 (

Samples made of tungsten (W) doped either with titanium carbide (W-1.1TiC) or tantalum carbide (W-3.3TaC) were exposed to a low-energy (40 eV/D), high-flux (1.8–5 × 10 23 D/m 2 s) deuterium (D) plasma at temperatures of about800 K, 1050 K, and 1250 K to a fluence of about1 × 10 27 D/m 2 . The deuterium (D) inventory in the samples was examined by nuclear reaction analysis and thermal desorption spectroscopy. At 800 K the D bulk concentrations and total D inventories in W-1.1TiC and W-3.3TaC were more than one order of magnitude higher compared to that in pure polycrystalline W. At 1050 K and 1250 K the D concentrations in all types of samples were very low (≤10 −5 at. fr.); however the D inventories in W-1.1TiC were significantly higher compared to those in W-3.3TaC and pure W. It is suggested that D trapping inside the carbide precipitates and at their boundaries is essential at high temperatures and high incident fluxes, especially in W-1.1TiC.

The deuterium (D) bubbles formed in W exposed to high flux D plasma were researched by scanning electron microscopy and transmission electron microscopy. After D plasma exposure at 500 K and 1000 K, a layer of nano-sized bubbles were homogenously distributed in W subsurface region. The D bubbles were homogenously nucleated due to the high D concentration, and the nucleation process is not related to the vacancy defects. At low temperature (500 K), D bubbles can grow by surface blistering, which caused different nano scale morphologies on different surfaces. At high temperature (1000 K), D bubbles mainly grow by vacancy clustering, which caused pinholes on the surface.

Materials exposed to plasma may undergo various forms of surface modifications. Among the important phenomena for tungsten - as the prime candidate plasma-facing material for fusion devices – is a formation of helium-induced fibreform nanostructure, so-called tungsten fuzz. In this paper, we report direct observations of the interaction of the pre-prepared fuzzy tungsten surfaces with ELMy H-mode plasmas in the COMPASS tokamak as well as consequent ex-situ morphological analyses, with a particular focus on arcing as a potential erosion mechanism. Arcing events are documented from high-speed camera imaging. The sample surfaces are examined by scanning electron microscopy. Arc traces were observed on all samples, while their number was dependent on the sample position and orientation. Inside the arc traces, localized melting and densification of the original fuzz was observed, resulting in thickness reduction. The modified structure still retained some porosity and did not extend into the bulk.

In this study, we developed a revised stationary force balance model for particles in the regime a/labda D

A steady-state high-flux H or He plasma beam was balanced against the pressure of a Sn vapor cloud for the first time, resulting in a self-regulated heat flux intensity near the liquid surface. A temperature response of the liquid surface characterized by a decoupling from the received heating power and significant cooling of the plasma in the neutral Sn cloud were observed. The plasma heat flux impinging on the target was found to be mitigated, as heat was partially dissipated by volumetric processes in the vapor cloud rather than wholly by surface effects. These results motivate further exploration of liquid metal solutions to the critical challenge of heat and particle flux handling in fusion power plants.

The effect of the addition of rhenium to tungsten on the formation of a nanostructure referred to as ‘fuzz’ when exposed to helium plasmas at fusion relevant ion fluxes was investigated in the Magnum and Pilot PSI devices at the FOM Institute DIFFER. The effect rhenium had on fuzz growth was seen to be dependent on time, temperature and flux. Initial fuzz growth was seen to be highly dependent on grain orientation, with rhenium having little effect. Once the fuzz was fully developed, the effect of grain orientation disappeared and the rhenium had an inhibiting effect on growth. This could be beneficial for inhibiting fuzz growth in a future fusion reactor, where transmutation of tungsten to rhenium is expected. It also appears that erosion or annealing of the fuzz is limiting growth of fuzz at higher temperatures in the range of ∼1340 °C.

We have investigated plasma detachment phenomena of high-density helium plasmas in the linear plasma device Pilot-PSI, which can realize a relevant ITER SOL/Divertor plasma condition. The experiment clearly indicated plasma detachment features such as drops in the plasma pressure and particle flux along the magnetic field lines that were observed under the condition of high neutral pressure; a feature of flux drop was parameterized using the degree of detachment (DOD) index. Fundamental plasma parameters such as electron temperature (Te) and electron density in the detached recombining plasmas were measured by different methods: reciprocating electrostatic probes, Thomson scattering (TS), and optical emission spectroscopy (OES). The Te measured using single and double probes corresponded to the TS measurement. No anomalies in the single probe I – Vcharacteristics, observed in other linear plasma devices [ 16, 17,36], appeared under the present condition in the Pilot-PSI device. A possible reason for this difference is discussed by comparing the different linear devices. The OES results are also compared with the simulation results of a collisional radiative (CR) model. Further, we demonstrated more than 90% of parallel particle and heat fluxes were dissipated in a short length of 0.5 m under the high neutral pressure condition in Pilot-PSI.

Impurity seeding of noble gases is an effective way of decreasing the heat loads onto the divertor targets in fusion devices. To investigate the effect of noble gases on deuterium retention, tungsten targets have been implanted by different noble gas ions and subsequently exposed to deuterium plasma. Irradiation induced defects and deuterium retention in tungsten targets have been characterized by positron annihilation Doppler broadening and thermal desorption spectroscopy. Similar defect distributions are observed in tungsten irradiated by neon and argon, while it is comparatively low in the case of helium. The influence of helium pre-irradiation on deuterium trapping is found to be small based on the desorption spectrum compared with that of the pristine one. Neon and argon pre-irradiation leads to an enhancement of deuterium trapping during plasma exposure. The influence on deuterium retention is found to be argon > neon > helium when comparing at a similar crystal damage level.

Results are presented of large-scale Molecular Dynamics simulations of low-energy He bombardment of W nanorods, or so-called ‘fuzz’ structures. The goal of these simulations is to see if ballistic He penetration through W fuzz offers a more realistic scenario for how He moves through fuzz layers than He diffusion through fuzz nanorods. Instead of trying to grow a fuzz layer starting from a flat piece of bulk W, a new approach of creating a fully formed fuzz structure 0.43 µm thick out of ellipsoidal pieces of W is employed. Lack of detailed experimental knowledge of the 3D structure of fuzz is dealt with by simulating He bombardment on five different structures of 15 vol% W and determining the variation in He penetration for each case. The results show that by far the most important factor determining He penetration is the amount of open channels through which He ions can travel unimpeded. For a more or less even W density distribution He penetration into fuzz falls off exponentially with distance and can thus be described by a ‘half depth’. In a 15 vol% fuzz structure, the half depth can reach 0.18 µm. In the far sparser fuzz structures that were recently reported, the half depth might be 1 µm or more. This means that ballistic He penetration offers a more likely scenario than He diffusion through nanorods for how He moves through fuzz and may provide an adequate explanation for how He penetrates through the thickest fuzz layers reported so far. Furthermore, the exponential decrease in penetration with depth would follow a logarithmic dependence on fluence which is compatible with experiments. A comparison of these results and molecular dynamics calculations carried out in the recoil interaction approximation shows that results for W fuzz are qualitatively very different from conventional stopping power calculations on W with a similarly low but homogeneous density distribution.

For tungsten exposed to low-energy hydrogen-plasmas, it has been thought that grains with 〈 1 1 1 〉 surface normal are most susceptible to blistering while those with 〈 0 0 1 〉 surface normal are virtually impervious to it. Here, we report results showing that non-uniformity of blister distribution depends on the state of the surface due to polishing. In electrochemically polished material blisters appear on the grains with all orientations, while in mechanically polished material blister-free areas associated with particular orientations emerge. On the other hand, blistering is shown to have a strong dependence on the level of deformation within particular grains in partially recrystallized material.

Helium retention is measured in tungsten samples exposed to mixed H/He plasma in the Magnum-PSI linear plasma device. It is observed that there is very little He retention below helium ion impact energies of 9.0 ± 1.4 eV, indicating the existence of a potential barrier which must be overcome for implantation to occur. The helium retention in samples exposed to plasma at temperatures >1000 K is strongly correlated with nano-bubble formation measured using grazing incidence small-angle x-ray scattering. The diameters of nano-bubbles were not found to increase with increasing helium concentration, indicating that additional helium must be accommodated by increasing the bubble concentration or an increase in bubble pressure. For some samples pre-irradiation with heavy ions of 2.0 MeV energy is investigated to simulate the effects of neutron damage. It is observed that nano-bubble sizes are comparable between samples pre-irradiated with heavy-ions, and those without heavy-ion pre-irradiation.

Lithium-coated high-Z substrates are planned for use in the NSTX-U divertor and are a candidate plasma facing component (PFC) for reactors, but it remains necessary to characterize the gross Li erosion rate under high plasma fluxes (>1023 m−2 s−1), typical for the divertor region. In this work, a realistic model for the compositional evolution of a Li/D layer is developed that incorporates first principles molecular dynamics (MD) simulations of D diffusion in liquid Li. Predictions of Li erosion from a mixed Li/D material are also developed that include formation of lithium deuteride (LiD). The erosion rate of Li from LiD is predicted to be significantly lower than from pure Li. This prediction is tested in the Magnum-PSI linear plasma device at ion fluxes of 1023–1024 m−2 s−1 and Li surface temperatures ⩽800 °C. Li/LiD coatings ranging in thickness from 0.2 to 500 μm are studied. The dynamic D/Li concentrations are inferred via diffusion simulations. The pure Li erosion rate remains greater than Langmuir Law evaporation, as expected. For mixed-material Li/LiD surfaces, the erosion rates are reduced, in good agreement with modelling in almost all cases. These results imply that the temperature limit for a Li-coated PFC may be significantly higher than previously imagined.

The effect of severe plastic deformation on the deuterium retention in tungsten exposed to high-flux low-energy plasma (flux ∼ 1024 D/m2/s, energy ∼ 50 eV, and fluence up to 3 × 1026 D/m2) at the plasma generator Pilot-PSI was studied by thermal desorption spectroscopy and scanning electron microscopy. The desorption spectra in both reference and plastically deformed samples were deconvolved into three contributions attributed to the detrapping from dislocations, deuterium-vacancy clusters, and pores, respectively. The plastically induced deformation, resulting in high dislocation density, does not change the positions of the three peaks, but alters their amplitudes as compared to the reference material. The appearance of blisters detected by scanning electron microscopy and the desorption peak attributed to the release from pores (i.e., deuterium bubbles) were suppressed in the plastically deformed samples but only up to a certain fluence. Beyond 5 × 1025 D/m2, the release from the bubbles in the deformed material is essentially higher than in the reference material. Based on the presented results, we suggest that a dense dislocation network increases the incubation dose needed for the appearance of blisters, associated with deuterium bubbles, by offering numerous nucleation sites for deuterium clusters eventually transforming into deuterium-vacancy clusters by punching out jogs on dislocation lines.