First Wall Research Articles

Wall heating by subcritical energetic electrons generated by the runaway electron avalanche source *

Subcritical energetic electrons (SEEs) produced by the runaway electron (RE) avalanche source at energies below the runaway threshold are found to be the primary contributor to surface heating of plasma-facing components (PFCs) during final loss events. This finding is supported by theoretical analysis, computational modeling with the Kinetic Orbit Runaway electrons Code (KORC), and qualitative agreement with DIII-D experimental observations. The avalanche source generates significantly more secondary electrons below the runaway threshold, which thermalize rapidly when well-confined. However, during a final loss event, the RE beam impacts the first wall, and SEEs are deconfined before they can thermalize. Additionally, because the energy deposition length decreases faster than energy, the deposited energy density, and thus the maximum PFC surface temperature change, is larger for SEEs than REs. KORC simulations employ an analytic first wall to model particle deconfinement onto a non-axisymmetric wall composed of individual tiles. PFC surface heating is calculated using a 1D model extended to include an energy-dependent deposition length scale. Simulations of DIII-D qualitatively agree with infrared (IR) imaging only when SEEs from the avalanche source are included. These results demonstrate that SEEs are the dominant contributor to PFC surface heating and indicate that the avalanche source plays a critical role in the PFC damage caused during final loss events. The prominence of SEEs also has important implications for interpreting IR imaging, one of the primary diagnostics for RE-wall interaction diagnosis, despite REs dominating the energy and current density. This result improves predictions of wall damage due to post-disruption REs to estimate material lifetime and design RE mitigation systems for ITER and future reactors.

RFX-mod2 diagnostic capability enhancements for the exploration of multi-magnetic-configurations

The RFX-mod2 device, the upgraded version of the previous RFX-mod with a modified magnetic boundary, is presently under realization and will start to be operated in 2025. Significant upgrades of the diagnostic capabilities have been proposed and are under development. These include a largely increased number of in-vessel magnetic and electrostatic sensors, a new fast reciprocating manipulator for the exploration of the edge plasma in a wide range of experimental conditions, the improved Thomson scattering and soft x-ray diagnostics system for a detailed determination of the behavior of the electron temperature profile, new dedicated systems for the space and time resolved analysis of x-ray spectra and neutron rate, a reflectometric diagnostic for real-time determination of plasma position, two diagnostics devoted to the imaging of light impurities and influxes behavior along with arrays of halo current sensors. These diagnostic upgrades will be accompanied by a significant effort to improve the control of the electron density and of the impurity influxes by means of proper treatment of plasma facing components with in-vessel fixed electrodes distributed over the first wall. The described advancements will allow a deeper understanding of physics phenomena in the wide variety of magnetic configurations, including the tokamak, the reversed-field pinch and the Ultra-low q, which can be produced in RFX-mod2 thanks to its flexibility and unique MHD control capabilities.

Electron beam profile measurement using enhanced dual-techniques in high heat flux test facility at Institute for Plasma Research.

During operation of an ITER-like plasma fusion device, Plasma Facing Components (PFCs) of a divertor are subjected to steady-state and transient heat loads of up to 20 MW/m2. At High Heat Flux Test Facility (HHFTF) at Institute for Plasma Research, PFCs are subjected to such heat flux scenarios using 200 kW electron beam (EB). A heat flux distribution on PFCs is significantly influenced by the EB profile and scanning technique. A novel dual-technique approach is used to measure EB cross-sectional profiles in HHFTF. A single setup is designed, fabricated, and installed in HHFTF consisting of (1) the knife-edge technique involving graphite and tungsten rectangular tiles bonded on a copper block adjacent to each other to form a sharp edge and (2) the wire scanner technique wherein tungsten wires are placed at equal distances. Current collected by the setup is measured as EB is rastered across the wire and tile boundaries. Current measured as a function of EB position is used to calculate the EB profile using a customized LabVIEW software application. Experiments are conducted by varying EB power from 28 to 200 kW and correspondingly the full width at half maximum of the EB profile grows from 5.30 to 16.04mm, as observed by both these techniques. The COMSOL Multiphysics simulation of the wire scanner technique is performed to verify the experimental results. X-ray and infra-red imaging diagnostics of HHFTF are also used to measure the EB profile in a unique and first-of-its-kind way, and they are found to agree well with the dual-technique measurements.

Competitive barrier and trapping effects of helium bubbles on hydrogen isotopes migration behavior in tungsten

In fusion reactors, plasma-facing materials will be exposed to low-energy and high-flux plasma of the hydrogen isotopes fuel, i.e., deuterium (D) and tritium (T), along with helium (He), the product of nuclear fusion reactions. He irradiation is likely to trigger the formation of He bubbles beneath the material surface, exerting complex effects on the migration patterns of hydrogen isotopes. In this work, molecular dynamics simulations were employed to investigate the influence of nano-sized He bubbles on the D migration behavior in tungsten (W) during the continuous implantation and diffusion processes, which is more consistent with the fusion environment for plasma-facing materials. It is found that the influence of He bubbles on D migration behavior can be classified into barrier and trapping effects. These effects show a competitive relationship at different processes. During the implantation process, the barrier effect takes precedence over the trapping effect, leading to the reduced D retention. Conversely, during the diffusion process, the trapping effect dominates over the barrier effect, resulting in increased D retention. Due to the surface of He bubbles serves as the trapping sites for hydrogen isotopes, augmenting the size and density of He bubbles notably enriches the strong trapping sites in W. When the location of He bubble is closer to the surface, a higher frequency of interaction events occurs between the implanted D atoms and He bubbles, consequently increasing D retention. Additionally, it was also found that the lattice distortion in pure W is more pronounced than that in He bubble-containing W at high temperatures due to the robust interaction between D and He, which diminishes collisions between D atoms and the W lattice. This study gives a valuable insight into the influence of He bubbles on hydrogen isotopes migration behavior in the fusion environment.

Studies of erosion-deposition of plasma-facing materials due to plasma-wall interactions in EAST tokamak

Material erosion, migration, mixing, dust formation, and co-deposition were investigated on plasma-facing materials in the experimental advanced superconducting tokamak (EAST). The test samples (TSs) of molybdenum (Mo), tungsten (W), and carbon (C) were irradiated during EAST operations. All exposed TS surfaces are modified by erosion and deposition processes resulting in the formation of thin layers containing a mixture of PFCs (Mo, W, Cu, Cr, Fe) and light elements (Li, C, Ca, N, O). The majority of chemical species are in a layer of thickness <650–900 nm. The particle size and structure of co-deposition were observed on each TS. Depth profiling by laser-induced breakdown spectroscopy (LIBS) suggests the presence of local and global impurities on the TSs resulting from the plasma-wall interaction (PWI). The processes involved in the PWI that caused erosion of EAST materials are heat flux and high energy excited impurity particles. The material is sputtered from the first wall plasma-facing components. Erosion and deposition processes occur simultaneously in the EAST. The migrated impurities get deposited globally on remote surfaces. A crack was observed in one of the W TSs due to a high heat load. The spectral intensity of co-deposition and substrate continuously changed during successive laser shots.

Neutron irradiation-induced shutdown dose rate estimation in EU DEMO divertor

The EU DEMO fusion reactor, which is part of the EUROfusion PPPT program and the successor to ITER, aims to facilitate the transition of fusion from the science laboratory to industrial use. Its key objective is net energy production. Its critical component, the divertor, manages power exhaust, ash, and helium removal. The key elements of the divertor include Plasma Facing Components (PFCs), shielding liners, reflector plates, divertor cassettes, and a vacuum vessel locking system. It is imperative that the shutdown dose rates are assessed to ensure workers' radiation protection during maintenance, plan the disposal of radioactive material, and ensure the overall safety of the plant. The shutdown dose rate tool employed in the present study is R2Smesh, which couples MCNP for radiation transport and FISPACT for neutron activation to calculate neutron-induced shutdown dose rates, utilizing an advanced, semi-heterogeneous EU DEMO divertor model, presenting a challenging task in neutron transport calculations.

Massive, long-lived electrostatic potentials in a rotating mirror plasma

Hot plasma is highly conductive in the direction parallel to a magnetic field. This often means that the electrical potential will be nearly constant along any given field line. When this is the case, the cross-field voltage drops in open-field-line magnetic confinement devices are limited by the tolerances of the solid materials wherever the field lines impinge on the plasma-facing components. To circumvent this voltage limitation, it is proposed to arrange large voltage drops in the interior of a device, but coexist with much smaller drops on the boundaries. To avoid prohibitively large dissipation requires both preventing substantial drift-flow shear within flux surfaces and preventing large parallel electric fields from driving large parallel currents. It is demonstrated here that both requirements can be met simultaneously, which opens up the possibility for magnetized plasma tolerating steady-state voltage drops far larger than what might be tolerated in material media.

Evolution of hydrogen isotopes retention behavior of in-situ boronization films in EAST

Effective management of hydrogen isotopes retention in plasma-facing materials (PFMs) is crucial, particularly when utilizing tritium (T) as fuel, for the success of burning plasma operations. Boronization, a widely employed technique for controlling fuel recycling and mitigating impurity influx from plasma-surface interactions into the core of burning plasma, significantly influences hydrogen isotopes retention in PFMs. In this study, boronization films were generated by ion cyclotron range of frequency (ICRF) discharge assisted boronization with C2B10H12 as boron source on tungsten substrates at Experimental Advanced Superconducting Tokamak (EAST) which employing ITER-like water-cooled W monoblock PFMs and components (PFCs), followed by in-situ glow discharge (GD) cleaning for 2 h and 20 shots (180 s) edge-plasma exposure. Employing the Material and Plasma Evaluation System (MAPES), representative samples were analyzed after each process. The resultant carbon–boron films, dense and continuous, exhibited thickness up to 120 nm and were identified as amorphous in structure. It was observed that the D2-GD cleaning resulted in a significant isotopic exchange effect, effectively reducing the hydrogen (H) retention in the carbon–boron films. This hydrogen isotope replacement efficiency was found to be influenced by the thickness of the films. Notably, after boundary plasma exposure, samples with thicker films demonstrated an enhanced capacity to capture deuterium (D), adsorbing 10 times more D than bare tungsten (W). Our findings offer transformative insights for T recycling analysis and the plasma operation of devices like International Thermonuclear Experimental Reactor (ITER), highlighting the impact of boronization and subsequent treatments on hydrogen isotope retention behavior in PFMs.

Exploring the effect of ELM and code-coupling frequencies on plasma and material modeling of dynamic recycling in divertors

Integrated modeling of plasma-surface interactions provides a comprehensive and self-consistent description of the system, moving the field closer to developing predictive and design capabilities for plasma facing components. One such workflow, including descriptions for the scrape-off-layer plasma, ion-surface interactions and the sub-surface evolution, was previously used to address steady-state scenarios and has recently been extended to incorporate time-dependence and two-way information flow. The new model can address dynamic recycling in transient scenarios, such as the application presented in this paper: the evolution of W samples pre-damaged by helium and exposed to ELMy H-mode plasmas in the DIII-D DiMES. A first set of simulations explored the effect of ELM frequency. This study was discussed in detail in this conference’s proceedings and is summarized here. The 2nd set of simulations, which is the focus of this paper, explores the effect of code-coupling frequency. These simulations include initial SOLPS solutions converged to the inter-ELM state, ion impact energy (E in) and angles (A in) calculated by hPIC2, and an improved heat transfer description in Xolotl. The model predicts increases in particle fluxes and decreases in heat fluxes by 10%–20% with the coupling time-step. Compared with the first set of simulations, the less shallow impact angle leads to smaller reflection rates and significant D implantation. The higher fraction of implanted flux (and deeper), in particular during ELMs, increases the accumulated D content in the W near-surface region. Future expansion of the workflow includes coupling to hPIC2 and GITR to ensure accurate descriptions of E in and A in, and W impurity transport.

Laser-induced breakdown spectroscopy for helium detection in beryllium coatings

Laser Induced Breakdown Spectroscopy (LIBS) method is considered to be a promising tool for analyzing the retention of hydrogen isotopes (D and T) and helium (He) on the first walls and divertor regions of future fusion reactors. Helium will be produced in DT reactions but could also be used in the initial non-nuclear phases of DEMO concepts. The present study investigates the He detection by LIBS method in the Be coatings simulating the deposits on the divertor plasma-facing components of JET while the results are also relevant for He detection in the deposits of other wall materials. The study was carried out in a vacuum vessel filled with 2–40 mbar argon background gas. It was shown that 2.8 at. % of He was confidently detectable by LIBS at optimized measurement conditions and the estimated limit of detection at used experimental conditions is approximately 0.7 at. %. The intensity of the He emission line at 587.56 nm was the strongest at the center of the laser-induced plasma plume. The He line intensity increased with the pressure of Ar gas but the broadening of the He line and the increase of the background emission and noise set an upper limit to the Ar background pressure usable for He detection. The application of the calibration-free LIBS procedure resulted in the overestimation of the He/Be ratio by several orders of magnitude. The overestimation can be explained by the deviation of LIBS plasma from the local thermodynamic equilibrium, which is caused by the very high excitation energy of He atoms.

Exploring of two connection forms of the tungsten tile and CuCrZr heat sink for first wall mockup

The first wall (FW) of a blanket or divertor faces high temperature plasma. Thus, the FW needs to withstand extremely high heat fluxes. Periodic plasma pulses bring some reliability challenges to FW structures. The FW is composed of tiles, heat sink and a bolt with different connection between the tile and the heat sink. In this study, the performances of two connection forms for the tile and the heat sink were investigated. The FLUENT software was used to evaluate the heat transfer performances for the two types structure. Elastic-plastic analysis was subsequently adopted to estimate the structural properties based on the thermal hydraulic results. At the same time, the thermal and structural results under a transient high thermal load (2 MW/m2) were also analyzed. Finally, structural fatigue analysis under a cyclic thermal load with 0.5 MW/m2 was also performed. The results indicated that there was a large discrepancy in the tile results for the two connections, while the differences in the heat sink result were not significant. The advantages and disadvantages of two connections also be discussed at last.

Multi-scale lumped modeling of micro-channels cooling structure for W7-X divertor unit target module

A novel concept for the cooling system of the W7-X divertor unit target module has been proposed, which involves tiles equipped with parallel arrays connected by hundreds of sub-millimeter rectangular micro-channels (MCs), obtained using Additive Manufacturing techniques. The structural material proposed for the heat sink substrate is galvanized copper, while the plasma facing material is tungsten. To reduce the high computational cost of thermal-hydraulic simulations of the tiles, a multi-scale lumped modeling approach has been built in previous studies. This involves replacing a group of hydraulic parallel MCs with a porous strip (PS), suitably calibrated to reproduce similar thermal-hydraulic behavior of the MCs. The aim of the current work is to verify that the PS-model is also suitable for the thermo-mechanical stress evaluation, comparing the results from arrays with MCs and PS. Since the heat sink and plasma-facing tiles are bonded, the interfacial delamination and shear stresses are analyzed, being crucial for structural integrity. The analyses, carried out in the elastic regime, show that the PS model returns conservative results when compared to the MCs model, with an overestimation of the delamination and shear stresses at the free edge. Results from both models reveal nearly identical interfacial stress predictions at the free surface edge. Moreover, it is found that both MCs and PS blocks, located near the bond interface, contribute to high-stress fluctuations that could lead to the delamination of the bond interface, suggesting that the distance of the microchannel from the interface should be increased. The PS model can reliably be used to design and verify the most convenient series/parallel connection of the tiles in the divertor unit target module, taking operational constraints into consideration.

In-situ TEM characterization and atomistic simulation of cavity generation and interaction in tungsten at 800 °C under dual W2+/He+ irradiation

In this work, we have investigated the role of pre-existing nano-scale cavities and irradiation sequence in helium clustering, and in the evolution of the cavity population in tungsten at 800 °C, by performing in-situ Transmission Electron Microscopy (TEM) characterizations under dual-beam W2+(600 keV)/He+(10 keV) irradiation to a total dose of 0.351 dpa, and supported by atomistic simulations. Pre-existing cavities, induced by single W2+ irradiation to 0.256 dpa, were seen to attract induced defects from subsequent irradiations, resulting in enhanced cavity size and a reduced number density in sequential irradiation when compared to concurrent irradiation. Such cavities developed faceted geometries and saturated in number density at an additional 0.051 dpa induced by He+ exposure in the sequential irradiation. In contrast, concurrent irradiations generate a denser cavity distribution that showed no sign of saturation or faceting up to the total irradiation damage of 0.351 dpa. In addition, Molecular Dynamics and Statics simulations were performed using LAMMPS software. These showed the pre-existing cavities to be trapping sites for helium atoms. Despite the large binding energy of ∼6 eV of those cavities, He-He interactions still occurred beyond the effective radius of such sinks, leading to additional He-He clustering not seeded by neighbouring cavities. These results point to the importance of selecting the irradiation conditions to simulate synergistic ion effects on cavity formation and evolution in plasma-facing tungsten-base materials.

Crack self-healing by high-temperature annealing of a 90W7Ni3Fe tungsten heavy alloy

WNiFe tungsten heavy alloys (WHAs) are leading candidates for fusion plasma-facing components, since they are toughened by embedding brittle W-particles in a ductile phase honeycomb matrix. We show that fatigue cracks in a 90W7Ni3Fe WHA, up to ≈ 1.3 mm long, are completely healed by vacuum annealing at 1300 °C for 24 h, as confirmed by microstructural observations and locating the healed crack initiation site, by both re-loading and re-precracking. Annealing also improves WHA toughness. Healing is not due to previously postulated mechanisms. We hypothesize that it derives from internal annealing stresses that extrude the ductile phase into adjacent cracks.

Ion–surface interactions in plasma-facing material design

A multi-scale simulation framework for ion–solid interactions in plasma-exposed materials provides crucial insight into advancing fusion energy and space electric propulsion. Leveraging binary-collision approximation (BCA) simulations, the framework uniquely predicts sputter yields and analyzes material transport within volumetrically complex materials. This approach, grounded in the validated BCA code TRI3DYN, addresses key limitations in existing models by accurately capturing ion–solid interaction physics. A case study is presented, highlighting the framework’s ability to replicate experimental sputter yield results, underscoring its reliability and potential for designing durable materials in harsh plasma environments. Insights into sputtering transport phenomenology mark a significant advancement in material optimization for improved resilience in plasma-facing applications.

Improved structure and enhanced properties of detonation sprayed W coating via hot isostatic pressing

Tungsten has garnered significant attention for its superior performance as a plasma-facing material in tokamak devices. However, its hardness and high ductile-to-brittle transition temperature present challenges for its application in plasma-facing components. To address these issues, this study employs the detonation spraying (DS) technique to prepare tungsten coatings on a TZM alloy substrate and further optimizes their properties through hot isostatic pressing (HIP) treatment. Advanced characterization techniques were utilized to conduct a detailed microstructural and compositional analysis of the coatings. The results indicated that the HIP treatment significantly reduced the porosity of the coatings, improving their density and overall quality. Additionally, HIP treatment led to a decrease in internal stress, a reduction in dislocations, and the grain growth within the coatings. The study also discussed how the HIP process improved the microstructure of the DS tungsten coatings through these mechanisms, thereby enhancing the coatings' thermodynamic properties. Mechanical property tests revealed that the coatings treated with HIP exhibited higher bonding strength and lower hardness, attributed to the increase in grain size and the reduction in dislocation density within the coatings. Thermal property tests demonstrated that the HIP-treated coatings possessed higher thermal diffusivity and specific heat capacity in the temperature range of 100 to 500 °C. Furthermore, the performance of the coatings under transient high heat flux conditions was significantly improved after the HIP treatment, showing better resistance to the thermal shock and cracking. These findings are of great importance for understanding the performance enhancement of DS tungsten coatings in nuclear fusion reactor applications via HIP, especially as candidate materials for the first wall of tokamak devices.

Finite Element Analysis and Experimental Verification of Thermal Fatigue of W-PFM with Stacked Structure

As the prime candidate for plasma-facing materials (PFM), the response of tungsten (W) to thermal shock loads is an important research topic for future fusion devices. Under heat loads, the surface of tungsten plasma-facing materials (W-PFM) can experience thermal damage, including brittle cracking and fatigue cracks. Therefore, exploring solutions for thermal damage of W-PFM remains one of the current research focuses. We propose a novel approach to mitigate thermal radiation damage in PFM, namely, the stacked structure W-PFM. The surface thermal stress distribution of the stacked structure W-PFM under heat loads was simulated and analyzed by the finite element method. As the foil thickness decreases, both the peak thermal stresses in the normal direction (ND) and rolling direction (RD) decrease. When the thickness decreases to a certain value, the peak thermal stress in the RD decreases to about 1384 MPa and no longer decreases; while the peak thermal stress in the ND approaches 0 MPa and can be neglected. In the range of approximately 5–100 mm, the accumulated equivalent plastic strain decreases sharply as the thickness decreases; in other thickness ranges, it decreases slowly. Thermal fatigue experiments were conducted on the stacked structure W composed of W foils with different thicknesses and bulk W using an electron beam facility. The samples were applied with a power density of 30 MW/m2 for 10,000 and 20,000 pulses. The cracks on the surface of the stacked structure W extended along the ND direction, while on the surface of bulk W, besides the main crack in the ND direction, a crack network also formed. The experimental results were consistent with finite element simulations. When the pulse number was 10,000, as the thickness of the W foil decreased, the number and width of the cracks on the surface of the stacked structure W decreased. Only four small cracks were present on the surface of stacked structure W (0.05 mm). When the pulse number increased to 20,000, the plastic deformation and number of cracks on the surface of all samples increased. However, the stacked structure W (0.05 mm) only added one small crack and had the smallest surface roughness (Ra = 1.536 μm). Quantitative analysis of the fatigue cracks showed that the stacked structure W-PFM (0.05 mm) exhibited superior thermal fatigue performance.

Electronically Temperature-Dependent Interplay between He and Trivacancy in Tungsten Plasma-Facing Materials.

Both microvoids and helium (He) impurities are widely present in tungsten (W) plasma-facing materials (PFMs), where the interaction between microvoids and He atoms has led to the intriguing development of microvoids. In this paper, we comprehensively investigated the interaction between He atoms and trivacancy (V3), a fundamental microvoid in W-PFMs, at the level of tight-binding theory. Our study showed that He atoms can catalyze the decomposition of the original V3 or facilitate its transformation into another V3 variant. We propose that a He atom near the V3 defect induces significant changes in the distribution of d-electron charges within the W atoms lining the inner wall of the V3 defect, making the W atom nearest to this He atom cationic and the other W atoms anionic. The attractive interaction between them promotes the decomposition and deformation of V3. As electronic excitation increases, the ionization of W atoms on the V3 wall gradually intensifies, thereby enhancing the cationic characteristics of the W atoms closest to the He atom. This process also prompts other W atoms to shift from anions to cations, leading to a transition in the electrostatic interactions between them from attraction to repulsion. This transformation, driven by electronic excitation, plays a significant inhibitory role in the decomposition and deformation of V3.

Study of Nanostructured Layer Growth on Plasma-Facing Materials Irradiated with Helium Plasma

Tungsten and stainless steel samples were irradiated with stationary helium plasma in the plasma linear multicusp plasma device. The surface of the material is modified under the influence of helium plasma with the formation of nanostructures and microstructures on the surface. The fluence of helium ions equal to 8 × 1027 ions/m2 was achieved on the tungsten sample. Depending on the helium ion fluence, fuzzlike layers, loops, and bubbles of 20- to 500-nm scale were formed on the tungsten surface. The fuzz layer thickness depends on the duration of plasma irradiation in a wide range of fluence. Saturation of the growth of the thickness of the tungsten fuzz layer was observed at a fluence of more than 8 × 1026 ions/m2. The growth of microstructures and nanostructures on the surface of stainless steel irradiated with helium plasma was observed. The growth of nanostructured layers is explained by a theoretical model considering the dynamics of adatoms under the influence of plasma.

First-principles study of helium solution and diffusion in tungsten borides

The behavior of helium (He) in plasma facing materials (PFMs) is crucial for the performance of PFMs under neutron irradiation in a fusion environment. First-principles density functional theory calculations have been performed to model the solution and diffusion of He in six stoichiometric ground-state tungsten borides (WxBy), including W2B, WB, WB2, W2B5, WB3 and WB4. The solution energy of He interstitial in W borides linearly decreases with increasing B content, except for the deviation in WB and WB3. The mechanical and electronic contributions to He solubility in WxBy have been discussed. The calculated diffusion energetics indicates that He transport in W2B5, WB3 and WB4 is much more difficult than in W2B, WB and WB2. Subsequently, we investigate the influence of vacancies on the He behavior. Except for WB3, He is more strongly bound in W vacancies rather than B vacancies and the He binding energy to a W vacancy generally decreases with increasing B content in WxBy. These findings can be used to predict the potential influence of He on the properties of different W borides in the fusion environment.