In tokamak devices, impurities generated by plasma-wall interactions (PWI) can cause erosion of plasma-facing components (PFCs), reduce their operational lifetime, and adversely affect fusion plasma performance. As a critical component directly exposed to edge plasma, the movable limiter was subjected to intense PWI during discharges, resulting in substantial impurity accumulation on its surface. In this study, an integrated fiber-optic LIBS system was employed to characterize the impurity deposition on the surface of the carbon fiber composite (CFC) movable limiter from the EAST tokamak. Depth-resolved LIBS spectra combined with Focused Ion Beam (FIB) and Scanning Electron Microscopy (SEM) cross-sectional imaging were used to estimate the thickness of the impurity layer. The average laser ablation rate was determined to be approximately 250 nm/pulse. Spatial distribution analysis of impurity elements reveals that impurity accumulation is more pronounced on the electron-side regions compared to the ion-side. An abnormally high concentration of copper (Cu) was observed in the bottom region, which is likely associated with local damage to the limiter and subsequent migration of the Cu bonding layer under intense plasma exposure. Calibration-free LIBS (CF-LIBS) was employed to quantitatively analyze two samples collected from the upper and bottom regions to verify the accuracy of the spatial distribution measurements. The calculated Cu content in the bottom sample reached 87.28%, significantly higher than the 33.27% in the upper sample, which is consistent with the results obtained from the spatial distribution measurements. These results provide valuable insights for understanding PWI and optimizing impurity control strategies in fusion devices and verify the feasibility of in situ measurements.

Fuzz growth under high-flux and low-energy helium plasma environment has been of intense interest for the plasma-facing materials research in nuclear fusion. This work investigated the surface morphology and subsurface microstructure changes in a gradient nanostructured tungsten-rhenium (25 atomic %) alloy exposed to 90 eV He plasma (flux: similar to 2.8 & times; 10(22) He/m(2)s) at 1173 K. Compared with coarse-grained counterparts, the nanostructured grains, introduced by ultrasonic shot peening technique, can effectively retard fuzz nucleation at low fluence (similar to 5.0 & times; 10(24) He/m(2)), yet undergo full fuzz formation at high fluence (similar to 5.0 & times; 10(25) He/m(2)). In addition, at the low fluence, fuzz initiates as nanoscale protrusions with pronounced convex curvature relative to the substrate; at the high fluence, near-surface He bubble saturation precedes morphological transition. Our findings suggest that fuzz growth may be a self-sustained process governed by surface diffusion, which provides new insights into the fundamental understanding of fuzz formation in plasma-facing materials.

A dynamic radiation belt—exhibiting characteristics of intense radiation, high density, and low electron temperature, akin to multifaceted asymmetric radiation from the edge (MARFE)—was observed and investigated in the Experimental Advanced Superconducting Tokamak (EAST). This belt forms in the divertor region with impurity injection and subsequently migrates towards the X -point and the high-field side (HFS) of the plasma boundary after the transition from H mode to L mode (H–L transition) or confinement degradation. The belt traverses the inner mid-plane, reaches the opposite divertor, and subsequently retreats back to its initial divertor location. This migration is accompanied by a concurrent evolution of poloidal asymmetry in key plasma parameters, including electron density, radiation profiles, neutral pressure, and wall heat load distribution. The migration of the radiation belt is closely related to the evolution of the edge neutral pressure. Reversing the toroidal magnetic field ( B _T ) direction inverts both the initial position of the radiation belt and its trajectory. The reversal of neutral pressure asymmetry between upper and lower divertors following B _T direction change demonstrates the role of B -direction-dependent drift effects. Furthermore, the migration of the radiation belt is consistently accompanied by localized fluctuations with frequencies up to 10 kHz in bothplasma radiation and density signals. During belt migration, the magnetic oscillations transition from a broadband spectrum to a coherent narrowband emission, identified as an m / n = 2/1 magnetohydrodynamic (MHD) instability mode. The frequency of the mode is positively correlated with the distance of the radiation belt from its initial position. Excessive increases in the magnetic oscillation frequency typically precede plasma disruptions, which may suggest a critical link between oscillation spectral characteristics and disruption onset mechanisms.

Abstract Boronization, a critical wall conditioning technique in fusion devices, is employed to suppress light impurities like carbon (C) and oxygen (O) and mitigate material sputtering, with future applications planned for ITER. This study examines the effects of two boronization methods—glow discharge boronization (GDB) and real-time boron powder injection (BPI)—on tungsten (W) sputtering behavior in the EAST divertor, utilizing spectroscopic diagnostics to monitor both W atoms and tungsten deuteride (WD) molecules. Results demonstrate that GDB effectively reduces C/O impurities and suppresses physical W sputtering but unexpectedly enhances WD sputtering due to increased deuterium retention on the W surface, while BPI simultaneously suppresses both W and WD sputtering through radiative cooling and reduced impurity levels. These findings provide valuable insights for optimizing wall conditioning strategies in all-metal first-wall fusion devices, highlighting the importance of method selection for impurity and erosion control.

High-temperature and high-pressure (HTHP) water is employed as the core coolant and heat transfer medium for water-cooled ceramic breeder (WCCB) blankets. This environment significantly aggravates the surface oxidation of reduced activation ferritic/martensitic (RAFM) steels for WCCB blankets, which further modulates the hydrogen isotopes permeation behavior through the steel. Dissolved oxygen (DO) has been widely recognized as the key governing factor determining the microstructure and phase composition of the oxide scale formed during oxidation. Accordingly, it is of paramount importance to elucidate the influence mechanism of oxide scales formed under different DO conditions on the hydrogen isotopes permeation behavior of RAFM steels. In this work, the oxidation behavior of the CLF-1 steel exposed to simulated HTHP aqueous environments under two DO levels (10 ppb and 200 ppb) was investigated. Experimental results demonstrate that the 10 ppb DO group yielded an oxide scale with larger grain size and an inner oxide layer thickness of 708 nm, yet no Cr was detected in its near-surface region. By contrast, the 200 ppb DO group presented more pronounced oxidation weight gain, a porous oxide morphology with distinct cracks at the inner oxide/steel matrix interface, and uniform Cr distribution across the entire oxide scale. Measurements also confirmed that oxidized CLF-1 steel exhibited a significantly lower hydrogen isotopes permeability than bare unoxidized steel. The 200 ppb double-sided oxidized specimen delivered superior hydrogen isotopes permeation resistance over the 10 ppb counterpart.

Developing barrier coatings with high resistance to hydrogen isotope permeation in the structural materials of fusion reactor blankets is a critical priority. In this work, we propose a novel approach to address this challenge by utilize the oxide layer of a modified Reduced Activation Ferritic-Martensitic (RAFM) steel after proper oxidization process. The SIMP steel, which contains 1.43% silicon and 10.8% chromium, is distinguished from the 9% chromium CLF-1 steel by its additional silicon. Gas-driven permeation (GDP) experiments show that, after oxidation at 700 °C in air for 30 h, the SIMP steel forms an oxide layer with a thickness of approximately 100 nm. Deuterium GDP tests reveal a permeation reduction factor (PRF) of 2 × 10 ^4 at 560 °C for oxidized SIMP steel, compared to a PRF of approximately 100 for oxidized CLF-1 steel. This high PRF is attributed to the dense Cr–Mn oxide layer due to the presence of Si, independent of the substrate heat treatment, as confirmed by microstructural analysis. Following irradiation with Au ions up to 12 dpa, the PRF of the oxidized SIMP steel decreases by less than one order of magnitude, with partial recovery observed at elevated temperatures. This work provides new insight into the development of tritium permeation barriers, highlighting a promising route that combines low activation, cost-effectiveness, and strong manufacturability.

Fuzz formation in plasma-facing materials remains insufficiently understood. We correlate subsurface helium-bubble evolution with surface morphological transitions in tungsten-rhenium (25 atomic %) exposed to high-flux, 90 eV He plasma at similar to 1173 K. Microstructural characterization reveals three sequential surface stages-incubation (ridges), nucleation (bundles of nanocrystalline protrusions, 'tendril embryos'), and growth (tendrils)-and three corresponding types of bubbles: substrate bubbles beneath wavy or smooth surfaces, embryo bubbles at grain boundaries, and tendril bubbles within the fuzz. Substrate bubbles attain size and density saturation prior to nucleation, indicating a dynamic subsurface bubble layer under continued plasma impact; embryo bubbles undergo anomalous growth that drives surface roughening and protrusion bifurcation, while polycrystalline tendrils emanate from a nanocrystalline near-surface layer, suggesting outward tip-driven growth. We show that elevated local surface temperature and a reduced surface melting point at protrusion and fiber tips generate capillary forces which, together with adatom surface diffusion, sustain fuzz nucleation and growth. The proposed thermodynamic, capillary-driven framework integrates subsurface bubble dynamics, surface diffusion, and plasma-wall interactions into a unified model of fuzz growth that accounts well for the experimental observations.

The tritium release behavior of Li2TiO3 breeder pebbles is governed by irradiation-induced defects, microstructure, and thermal transport processes. In fusion blankets, these materials are exposed not only to neutrons but also to gamma-rays generated during nuclear reactions. In this work, Li2TiO3 pebbles were pre-irradiated with lowdose gamma-rays (100 and 500 kGy), subsequently irradiated with thermal neutrons, and then examined by tritium thermal desorption spectroscopy (TDS) under heating and isothermal conditions. The dominant tritium desorption peak shifted non-monotonically with gamma-ray dose, whereas the apparent activation energies obtained from peak-shift analysis changed only slightly, from 0.21 eV for the 0 kGy sample to 0.16 and 0.13 eV for the 100 and 500 kGy samples, respectively. These values represent effective kinetic parameters for the coupled release process rather than single diffusion or de-trapping barriers. Isothermal TDS showed that gamma-pre-irradiation slightly increased the fraction of tritium released during the initial holding stage, while the overall effect remained limited. The release curves can be approximated by an effective diffusion model after the early transient stage; however, the initial deviation from the model indicates the contribution of trapping or de-trapping, defect recovery, and surface processes. The results suggest that, under the present low-dose gamma-pre-irradiation conditions, the intrinsic defects and porous microstructure of Li2TiO3 pebbles still dominate the overall tritium release behavior.

The retention of tritium (T) in plasma-facing materials significantly impacts the operational lifespan and safety of fusion reactors. Hydrogen isotope (HI) exchange offers a viable strategy for T removal in such reactors. While hydrogen (H) clusters are known to influence HI exchange behavior, the specific mechanisms and patterns of their effects remain unclear. In this study, we employed extended nanosecond-scale molecular dynamics (MD) simulations to investigate the influence of temperature, H concentration, and vacancy concentration on the formation of H clusters and subsequent HI exchange in tungsten (W). Our results indicate that increasing the temperature consistently enhances the rate of T removal. As the H concentration rises, H evolves from isolated atoms into rod-shaped clusters along the {100} and {110} crystal plane families, and further expands into platelet-like structures. During the isolated atom stage, the T removal rate continues to increase. However, in the platelet-like cluster stage, the removal efficiency initially rises and then declines. Additionally, high vacancy concentrations promote the widening of the isolated H atom stage. These findings provide valuable insights for optimizing T removal parameters in fusion reactors.

To meet the demanding requirements for high-precision, spatially-resolved diagnostics in plasma-material interaction (PMI) studies under the high-flux, high-magnetic-field environment of the SPARROW linear plasma device, an actively water-cooled Langmuir probe array system has been designed and developed. This design synergizes actively water-cooling with array layout requirements, with a focus on optimizing the cooling channel structure. Through systematic computational fluid dynamics (CFD) simulations, the thermal performance of the probe was quantitatively evaluated under Gaussian-distributed heat fluxes of 10 MW/m2, 15 MW/ m2, and 20 MW/m2, along with the impact on the probe body and the key insulating material (alumina ceramic). Under the 15 MW/m2 heat flux, the maximum temperatures of the tungsten tip and alumina sleeve are maintained at approximately 62% and 61% of their respective safety limits. Even under the extreme 20 MW/m2 condition, these key diagnostic components remain below 75% of their limits, demonstrating a substantial safety buffer that accommodates potential CFD uncertainties. By integrating innovative design with comprehensive thermal analysis, this research establishes key technical foundations for achieving efficient and reliable arrayed active diagnostics in extreme fusion-relevant plasma environments. It provides vital support for future highparameter plasma physics experiments.

Abstract This work investigates the effects of H 2 /He ratio and gas flow rate on cascaded arc H 2 /He mixed plasma by measuring discharge voltage and diagnosing downstream plasma parameters—including electron excitation temperature ( T exc ) and electron density ( n e )—via optical emission spectroscopy (OES). Experimental results show that with discharge current kept constant, the discharge voltage of the cascaded arc source first decreases and then increases as gas flow rate rises. Increasing the H 2 /He ratio (within the range of 0%–10%) in the gas flow elevates the discharge voltage and shifts the turning point (the gas flow rate corresponding to the minimum voltage) toward lower flow rates. Spectral data suggest that the addition of H 2 may promote molecular activated recombination (MAR) reactions, thereby altering the discharge behavior of the plasma source. At a location 290 mm downstream of the cascaded arc source, OES measurements reveal that n e increases continuously as gas flow rate rises from 200 sccm to 2000 sccm. In contrast, T exc exhibits a non-monotonic trend of first increasing and then decreasing, with the transition point shifting toward higher flow rates as H 2 content increases, which induces a similar evolution of ion flux. Such phenomena could be ascribed to the combined effects of cascaded arc source characteristics and plasma transport dynamics, which are jointly regulated by gas flow rate and H 2 content.

The effects of helium (He) irradiation on tungsten (W) surfaces have been of intense interest for plasma-facing materials research in nuclear fusion. This study explores the synergistic effects of rhenium (Re) alloying and ion channeling on irradiation blistering in W under 30 keV He ion exposure. Using in situ helium ion microscopy, coupled with electron backscatter diffraction and cross-sectional transmission electron microscopy, we have examined the morphological evolution of both channeling and non-channeling grains in pure W and W–25Re (atomic %). The findings show that crystallographic orientation primarily dictates the pathways of He implantation and accumulation, while Re alloying alters the material’s mechanical properties and He–vacancy interactions. Channeling grains in pure W demonstrate larger surface blisters and deeper subsurface bubbles and cracks. By comparison, while the addition of Re tends to reduce the He implantation range, it simultaneously promotes blistering deformation through enhanced plasticity. Consequently, the synergistic effects of ion channeling and Re alloying leads to the formation of the largest blisters in the channeling grains of W–25Re, offering new insights into the performance of W-based materials in fusion environments.

The transport behavior of hydrogen isotopes through tungsten (W) is a critical factor for the safety and economics of fusion reactors. This study investigates the impact of intense helium (He) plasma irradiation on the deuterium (D) transport characteristics in W. High-flux He plasma irradiation was utilized to induce a nanoscale fuzz structure on the W surface. The effects of this structure on the permeation and retention of D were subsequently examined using plasma-driven permeation (PDP) and thermal desorption spectroscopy (TDS). The TDS results reveal that the presence of fuzz significantly enhances D retention due to an increased density of trapping sites introduced by surface nanostructure and He bubble formation. In contrast, PDP results indicate only a slight reduction in the steady-state permeation flux. Moreover, comparative analysis between the first and second PDP cycles demonstrates a more substantial increase in the apparent diffusion coefficient for fuzz samples, indicating the formation of a higher density of irreversible traps. These findings suggest that while the fuzz structure acts as an effective near-surface reservoir for hydrogen isotopes, it has limited influence on long-term permeation behavior, which remains governed by bulk diffusion and reversible trapping mechanisms.

The reliable joining of plasma-facing tungsten (W) to structural reduced activation ferritic/martensitic steel is a key challenge in manufacturing the first wall for future fusion reactor blankets. The large mismatch in their coefficients of thermal expansion leads to significant residual stress, and brittle intermetallic compounds readily form at the interface, both of which severely degrade joint performance. Introducing a copper (Cu) interlayer between W and steel is an effective strategy to mitigate stress concentration and suppress brittle phases. However, achieving their high-strength metallurgical bonding is difficult due to the immiscibility of W/Cu and the limited solubility of Cu/steel. In this study, a robust W/Cu/steel joint was successfully fabricated by brazing steel to a high-performance W/Cu slice using a Cu-Ge filler metal. The results show that a body-centered cubic-structured interdiffusion layer, approximately 10 nm thick, forms at the W/Cu interface, with its interplanar spacing about clear larger than that of pure W. At the Cu/steel interface, a discontinuous network-like metallurgical structure is established. All joints fractured within the Cu interlayer regardless of the holding time, demonstrating high interfacial bonding strength. A maximum shear strength of 240 MPa was achieved, accompanied by good ductility, indicating an excellent strength-ductility synergy. This work reveals the atomic-scale interdiffusion mechanism at the W/Cu interface and the bonding behavior at the brazed Cu/steel interface. It provides an efficient route for fabricating first-wall components and offers key insights into the interfacial heat transfer and load-transfer behavior in W/Cu or W/Cu/steel plasma-facing components under fusion-relevant conditions.

There were several errors in figure 11 of the original article. While the first panel, showing the bubble distribution for the low fluence of 1.1 × 10 ^24 He m ^−2 , is correct, the second panel was mistakenly duplicated from the first. Additionally, the third panel should display the distribution data for the higher fluence of 5.0 × 10 ^24 He m ^−2 , and the data for the highest fluence of 5.0 × 10 ^25 He m ^−2 was completely omitted. These errors have been addressed in the corrected figure 11 provided in this corrigendum.

Tungsten, fabricated via powder metallurgy (PM) for its use as a plasma-facing material (PFM), inherently contains residual porosity that significantly influences its service performance. This study investigates the influence of initial porosity on the mechanical properties and deuterium (D) retention of ITER-grade tungsten. The results indicate a significant degradation in mechanical properties associated with a microstructure characterized by higher porosity, larger grain sizes, and a lower fraction of high-angle grain boundaries (HAGBs). Specifically, as the microstructure coarsened and porosity increased, flexural strength and flexural modulus decreased by up to 58.5 % and 8.3 %, respectively. The deformation of pores during rolling creates sharp tips that serve as stress concentration sites, while their presence at grain-boundary triple junctions impairs load sharing, ultimately promoting crack propagation and exacerbating intrinsic brittleness of tungsten. Regarding D retention, the total amount of trapped deuterium in as-rolled specimens varied by nearly an order of magnitude depending on the initial porosity. While recrystallization lowered D retention by a factor of 1.39 to 9.67, the retained amount remained positively correlated with the initial porosity. Hydrogen Isotope Diffusion and Trapping (HIDT) simulations, which show excellent agreement with experimental data, confirm that pores are responsible for at least 40 % of the total deuterium trapping.

Understanding the hydrogen isotope transport behavior (HITB) in the first wall (FW) is significant whereas simulations have been extensively employed. Most previous simulations use hydrogen transport parameters of pure tungsten (W) and reduced activation ferritic/martensitic (RAFM) steel, while largely neglecting hydrogen diffusion at their interface. Joining W and RAFM steel inevitably leads to the formation of a bonding region characterized by elemental interdiffusion, Fe-W binary phase precipitates, and increased densities of dislocations and microvoids. The HITB of the bonding region where complex microstructural features and trapping mechanisms coexist remains insufficiently understood. Therefore, this work develops a comprehensive approach to quantify the influence of the bonding region on the HITB in the FW by explicitly incorporating both diffusion and trapping effects. The diffusivity of the bonding region is obtained by minimizing the deviation between gas-driven permeation (GDP) experimental data and simulation results, while trap properties are evaluated using macroscopic rate equation modeling based on thermal desorption spectroscopy (TDS) data. Results show that the diffusivity of the bonding region lies between that of pure W and RAFM steel, and its trap density is at least an order of magnitude higher than that of pure W and RAFM steel. The permeation flux is delayed by the bonding region. Simulated retention indicates about 20% higher retention as considering the bonding region in the FW model, however the overall retention in the FW is not significantly affected due to the limited thickness of the bonding region. Additionally, to improve the predictive accuracy of the HITB model, the quasi-intrinsic diffusion coefficient ( $D$ ) of RAFM steel is obtained. This is achieved by employing an error minimization function that adjusts the value of $D$ to minimize the discrepancy between the simulated permeation flux and the experimental flux obtained from GDP measurements. This study establishes a refined HITB model for the FW.

The deuterium (D) retention in reduced-activation ferritic/martensitic (RAFM) steels under simultaneous hydrogen (H) and D plasma exposure was investigated, with a focus on the coupling effects of pre-existing displacement damage. Controlled damage profiles were introduced via heavy-ion irradiation using 4.5 MeV and 387 MeV Fe ions. The results indicate that heavy-ion irradiation significantly enhances D retention and induce high-temperature desorption above 1000 K. Crucially, the co-introduction of H effectively suppresses D retention, especially in the high-temperature regime. These findings imply that in a D-tritium (T) fusion environment, the co-existence of multiple hydrogen isotopes may naturally mitigate T retention, offering positive implications for T management and safety in future fusion reactors.

Tungsten (W) is a leading plasma-facing material for future fusion reactors, yet the crystallographic-orientation-dependent mechanisms of helium (He) irradiation induced surface swelling, blistering and roughening during high-temperature annealing remain insufficiently understood. Here we combine helium ion microscopy (HIM) with scanning electron microscopy (SEM), atomic force microscopy (AFM) and cross-sectional transmission electron microscopy (TEM) to systematically compare the surface and subsurface evolution of He-implanted single-crystalline W (100) and W (110) before and after vacuum annealing at 1273 K. Our results show that under room-temperature irradiation, both orientations exhibit a fluence-dependent transition from uniform swelling to blistering, with W (100) consistently developing larger and higher blisters than W (110). Upon annealing, new blisters emerge at previously non-blistering fluences and surface roughness increases substantially. This is attributed to thermally induced bubble coarsening and near-surface stress accumulation. Moreover, compared with W (110), W (100) exhibits deeper He ion channeling, higher slip symmetry and a larger number of operative slip systems, as well as a lower surface diffusion barrier, which together account for its greater swelling height, larger blister size and more pronounced annealing-induced roughening. This study provides new experimental evidence and insights into orientation-dependent damage mechanisms in W under He ion irradiation.