Future Reactors Research Articles

The role of ion-scale micro-turbulence in pedestal width of the DIII-D wide-pedestal QH mode

Abstract The low-edge rotation, intrinsically ELM-free, and improved confinement wide-pedestal quiescent H-mode (QH-mode), discovered in DIII-D tokamak, has pedestal widths exceeding the EPED-KBM model scaling typically by at least 25%. Ion-scale (k_y ρ_s<1) microturbulence and its role in setting the pedestal structure is investigated using the radially local δf gyrokinetic code CGYRO. The electromagnetic trapped electron mode (TEM) is unstable at the pedestal top, while plasma beta (β_e) is ~60% below the kinetic-ballooning mode (KBM) onset threshold and the electron temperature gradient (ETG) mode is found to be unstable in the peak gradient region. Nonlinear simulation reveals that the ion-scale turbulence could produce electron energy flux consistent with the flux inferred from power balance at the pedestal top, with a reasonable variation of the local E×B shearing rate; and the local neoclassical transport from NEO is dominant over the simulated turbulent transport in the ion energy flux channel. The simulated ion-scale turbulence produces much lower electron energy flux than inferred from experiment in the pedestal peak gradient region. A correction to the EPED-KBM pedestal width scaling is obtained based on the two-dimensional scan of pedestal top plasma beta (β_e) and normalized electron density and temperature scale lengths, a⁄L_(n_e ) , a⁄L_(T_e ) using CGYRO linear simulations. Mode transitions among TEM, micro-tearing mode (MTM), ion-temperature gradient (ITG) mode and KBM, are observed in the 2D scan at the pedestal top. A fixed normalized growth rate for these drift-type modes is taken to determine the pedestal width scaling, which shows good consistency with the QH experimental database on pedestal heights and widths. The onset of KBM instabilities and the local ExB shear suppression criterion set the lower and upper limit for the pedestal width of standard QH-mode, wide-pedestal QH-mode and type-I ELMy H mode. A potentially higher and wider pedestal is expected from the new scaling of pedestal width. This work presents an improved understanding of the ion-scale micro-turbulence of wide-pedestal QH-mode and sheds light on a promising scenario for future reactors, including ITER and beyond.

Effects of plasma nonuniformity on zero frequency zonal structure generation by drift Alfvén wave instabilities in toroidal plasmas

Abstract The effects of plasma nonuniformity on zero frequency zonal structure (ZFZS) excitation by drift Alfvén wave (DAW) instabilities in toroidal plasmas are investigated using nonlinear gyrokinetic theory. The governing equations describing nonlinear interactions among ZFZSs and DAWs are derived, with the contribution of DAW self-beating and radial modulation accounted for on the same footing, and the physics picture contributing to both channels clarified. The obtained equations are then used to derive the nonlinear dispersion relation, which is then applied to investigate ZFZS generation in several scenarios. In particular, it is found that the condition for zonal flow excitation by the kinetic ballooning mode (KBM) could be sensitive to plasma parameters and a more detailed investigation is needed to understand KBM nonlinear saturation, crucial for bulk plasma transport in future reactors.

ICRF resonance cones in the low-density scrape-off-layer of ASDEX Upgrade

Abstract The ICRF slow wave is a potential carrier for parallel RF electric fields known to cause unwanted plasma-wall interactions in magnetic confinement fusion experiments.
In nowadays machines the slow wave is usually confined to the far scrape-off layer or the limiter shadow, but conditions in future experiments and reactors may allow the slow wave to be propagative in a larger region.

Simulations with RAPLICASOL for various geometries show that the ICRF slow waves appear as the so-called resonance cones characterized by large localized electric fields.
The resonance cones emerge from the points along the plasma-antenna interface where (in the cold plasma approximation) the radio frequency electric field diverges.
We demonstrate that in the parameter range of interest, the propagation of the resonance cones in a plasma with a density gradient is defined by a simple geometric model, using the local plasma density and the frequency as input parameters.
In the context of experiments at ASDEX Upgrade, simulations illustrate that the resonance cones can emerge from a single tile of the ICRF antenna limiter.
Experiments at the Ion-cyclotron System Hardware Test ARrangement (ISHTAR) were conducted to test the detection principle in a simple environment.
In agreement with simulations and with predicted characteristics which depend on operation parameters, the resonance cones are excited by an RF antenna and propagate through the relatively homogeneous plasma in ISHTAR. 
The cones are detected at a distance from the antenna using two probes scanning through the plasma.
In ASDEX Upgrade, a single tile of an antenna limiter was modified to launch RF power into a specially tailored low-density scrape-off layer.
Probes at the midplane manipulator were then used to detect the wave electric fields at a distance from the RF source.
The detected RF signals show that the signal maxima are located close to the lower hybrid resonance density and are highest when the source and the probes are connected along magnetic field lines.
These observations agrees with the model for resonance cones from the simulations.

Simultaneous reduction of tungsten and rotation in the core region induced by RMP

Abstract Tungsten (W) impurity control is critical for plasma performance and a priority for ITER. The simultaneous reduction of W and rotation in the core region induced by resonant magnetic perturbations (RMP) has been found and understood in EAST. A positive feedback loop between the W and rotation is first proposed, resulting in core W accumulation and high rotation even in low-torque plasma before the RMP application. This cycle can be reversed by the edge rotation braking induced by RMP, causing a significant simultaneous reduction of W concentration and rotation. These new mechanisms are based on several repeatable experiments and confirmed by the modeling results from TGYRO and NTVTOK. It provides a new understanding of the RMP effects on W and rotation and can be used for W and rotation control in future reactors.

ELM-free H-mode phase and decoupling of peeling–ballooning stability boundary in the MAST Upgrade tokamak

Abstract A linear magnetohydrodynamic (MHD) peeling–ballooning stability analysis of the edge-localized mode (ELM)-free phase of a MAST Upgrade (MAST-U) H-mode plasma is presented. In contrast to other similar discharges, #47018 is found to have a significantly higher and wider pedestal during its ELM-free H-mode phase that lasts for approximately 80 ms; this is made possible by the reduced core MHD mode activity on the q = 2 surface. During this period, there is sustained decoupling of peeling and ballooning branches of the stability boundary on J–α space, opening an access channel to the second stability regime with higher peaks in pedestal current density J N , ped and pressure gradient (α). Such decoupling of the stability boundary has not previously been observed in MAST-U H-modes, and if such a condition can be readily reproduced, it opens a wide range of opportunities for MAST-U to explore low-collisionality peeling-limited pedestal regimes as well as advanced scenarios such as quiescent H-modes that are relevant to future reactors such as STEP and ITER.

Comprehensive Screening of Plasma-Facing Materials for Nuclear Fusion

Plasma-facing materials (PFMs) represent one of the most significant challenges for the design of future nuclear fusion reactors. Inside the reactor, the divertor will experience the harshest material environment: intense bombardment of neutrons and plasma particles coupled with extremely large and intermittent heat fluxes. The material designated to cover this role in ITER is tungsten. While no other materials have shown the potential to match the properties of tungsten, many drawbacks associated with its application remain, including cracking and erosion induced by a low recrystallization temperature combined with a high ductile-brittle transition temperature and neutron-initiated embrittlement; surface morphology changes (helium bubbles and fuzz layer) due to plasma-tungsten interaction with subsequent risk of spontaneous material melting and delamination; and low oxidation resistance in the case of air contamination. Exploring alternatives to tungsten requires the design of a multivariable optimization problem. This work aims to produce a structured and comprehensive material screening of PFM candidates based on known inorganic materials. The method applied in this study to identify the most promising PFM candidates combines peer-reviewed data present in the PAULING FILE database and first-principles density-function theory calculations focusing on two key PFM defects—namely, the sputtering of surface atoms and the incorporation of interstitial hydrogen, respectively characterized by the surface binding energy and the interstitial formation energy. The crystal structures and their related properties, extracted from the PAULING FILE, are ranked according to the heat-balance equation of a PFM subjected to the heat loads in the divertor region of an ITER-like tokamak. The materials satisfying the requirements are critically compared with the state-of-the-art in plasma-facing materials research and in studies of refractory materials exposed to high temperatures, plasma, and neutron bombardment. This comparison assesses their thermo-mechanical properties under such conditions, identifying a subset of promising materials for first-principles electronic structure calculations. Most previously known and extensively studied PFMs, such as tungsten, molybdenum, and carbon-based materials, are captured by this screening process, confirming its reliability. Additionally, less familiar refractory materials suggest performance that calls for further investigations. Published by the American Physical Society 2024

Radial electric field driven by vertical neutral beamlet injection under future fusion reactor conditions

Abstract The fast ions and electrons generated by neutral beam injection (NBI) can induce charge separation, resulting in radial electric fields. Employing beamlet injection of small cross-section may effectively generate radial electric field, and adjusting beamlet parameters allows active control over their distribution. A new NBI injection geometry system has been developed in the NEOE code to explore the potential for vertical beamlets injection in future fusion reactor scenarios. Both high-field side and low-field side beamlet vertical injections can establish radial electric fields. In scenarios dominated by collision effects, the direction of the radial electric field is influenced by the toroidal angle of the beamlet. Adjusting the poloidal angle can alter the location of the electric field shear. Injecting particles into the trapped region using broader banana orbits can establish electric fields within the plasma core. Alternatively, injecting particles into the passing region can yield higher electric fields. Under future reactor conditions, conservative estimates of the electric field shear may even surpass critical velocities, potentially contributing to instability suppression.

Reduction of neoclassical bulk-ion transport to avoid helium-ash retention in stellarator reactors

Abstract The reduction of neoclassical energy transport in stellarators has traditionally focussed on optimizing magnetic fields for small values of ‘effective helical ripple’ — ϵ eff , the geometric factor associated with electron 1 / ν transport—and relying on the radial electric field, E r , needed to maintain ambipolarity in the plasma, to simultaneously diminish ion energy losses to a tolerable level. As one must generally expect E r < 0 , such a strategy has a drawback for reactor operation, however, as negative values of E r tend to hinder the exhaust of helium ash, and this will become intolerable if it results in excessive fuel dilution. Theoretically, one can show that the neoclassical transport of low-Z impurities depends critically on the ratio L 11 e / L 11 i , where the L 11 σ are the neoclassical particle diffusion coefficients of the bulk-plasma electrons ( σ = e ) and ions ( σ = i ). Increasing the value of this ratio is shown here to counteract impurity retention, but maintaining good confinement of the bulk species requires this to be achieved by decreasing L 11 i rather than increasing L 11 e , implying the need for more than just minimization of ϵ eff in reactor design. To assess the prospects of such an endeavor, a predictive 1-D transport code is used here to determine the range of L 11 e / L 11 i values which arises in simulations of conventionally optimized reactor candidates. This range is found to be considerable, and includes examples with L 11 e / L 11 i values large enough to provide neoclassical temperature screening of the helium ash and even to flip the sign of the neoclassical convective velocity from inward- to outward-directed. More intriguingly, the strong reduction of L 11 i in such cases can also lead to the appearance of E r > 0 (a so-called ‘electron root’) within the core of plasmas having central densities as large as 2 × 10 20 m−3. Having E r > 0 in the plasma core produces the ideal situation in which all thermodynamic forces aid in the exhaust of helium ash, although this benefit is tempered by the small values of L 11 z which accompany an electron root (where σ = z denotes helium ash). Means are discussed for improving the results presented here, so that avoiding helium-ash retention can be explicitly targeted in future reactor optimizations.

Recent progress in the development of liquid metal plasma facing components for magnetic fusion devices

One of the most critical challenges for future fusion reactors is to develop longevity plasma-facing components (PFCs) exposed to extremely high heat and neutron loads. As opposed to those employing solid metals, PFCs with flowing liquid metals (LM) have shown self-healing, heat removal and good impurity control capabilities, all essential to fusion devices. Recently, significant progress in LM-PFC development has been reported globally, with data from several magnetic fusion devices. These studies reveal that LM-PFCs can endure extreme heat fluxes while maintaining plasma compatibility. New design concepts have been proposed and numerically analyzed, advancing models for liquid PFCs in future reactors. Despite existing technical challenges, these developments suggest that LM-PFCs hold promise for future fusion applications.

Repair of heat load damaged plasma–facing material using the wire-based laser metal deposition process

Due to its unique properties tungsten is a promising candidate as plasma-facing-material (PFM) in future nuclear fusion reactors. Tungsten features an exceptionally high melting point, high thermal conductivity, low tritium inventory and comparatively low erosion rate under plasma loading [1]. But given the extreme loads on the PFM during operation of a fusion reactor, the lifetime of plasma-facing components (PFC)s is limited. Currently, it is planned to replace damaged PFCs when they reach the end of their service life. However, the lifetime of PFCs could be increased by in situ repair using additive manufacturing technology (AM) in the form of direct-energy-deposition (DED). The wire–based laser metal deposition process (LMD-w) meets several necessary conditions for operation in the vessel and could be used for performing such in situ repairs.It was investigated if the LMD-w process is able to heal thermal induced surface cracks and roughening by remelting the substrate during deposition of tungsten. For this purpose, tungsten samples of 12 × 12 × 5mm3, which later served as substrate plates for the LMD-w experiments, were treated with combined steady-state and transient thermal loads in the electron beam facility JUDITH 2. These samples were brazed to a copper cooling structure and exposed to 105 thermal shocks of 0.5 ms duration and an intensity of Labs = 0.55 GW m−2 (FHF = 12 MW s0,5 m−2) at a base temperature of Tbase = 700 °C. This way, edge localized mode (ELM) like thermal load damage was induced on the tungsten samples. On these samples, different LMD-w and laser remelting process strategies were performed. Subsequently, these samples were analyzed, and it was examined that the healing of the pre-damaged substrate material was successful. In parallel, the laser remelting process was modeled in a thermal transient finite element method(FEM) simulation in order to gain an insight into the temperatures prevailing in the material during the process.

Collisional-Radiative modeling of unresolved transition array spectra near 200 Å from W17+-W25+ emissions for diagnostics of ITER edge plasma

Tungsten (W) is one of the major impurities in ITER and future DEMO reactors. However,diagnosing the ion density, temperature, and spatial distribution of tungsten ions in intermediate charge states, such as W8+-W27+, is difficult because of the lack of spectral line data. In this study, we observed a tungsten Unresolved Transition Array (UTA) spectrum, which has a pseudo-continuum structure, around W20+ in the Extreme Ultraviolet wavelength region in the Large Helical Device (LHD). We conducted Collisional-Radiative (CR) modeling for W17+-W25+. Two pseudo-continuum peaks corresponding to 5 s-5p transition and transition between doubly excited states of 5 s2-5s5p, are strongly emitted around 200 Å for each ion. The synthesized spectrum of W17+-W25+ ions reproduced the observed LHD spectrum near 200 Å. From the electron temperature dependence of the UTA spectral shape, the UTA shifted toward longer wavelength region with several pseudo-continuum peaks appearing for ions in lower charge states, with decreasing electron temperature from 0.4 keV to 0.2 keV. This result qualitatively explained the observed time evolution of the spectrum.

Experimental research on chemisorption energy storage performance for industrial waste heat recovery and conversion

This study investigates a solid chemisorption energy storage system utilizing a multi-component chloride salt composite adsorbent with a mass ratio of NH4Cl, CaCl2, MnCl2, and expanded natural graphite treated with sulfuric acid (ENG-TSA). The system features a shell-and-tube heat exchanger and analyzes refrigeration and heat storage performance under varying industrial low-temperature waste heat (60–110 °C) conditions, with condensation and evaporation temperatures of 0–25 °C and discharging temperatures of 40–60 °C. The results indicate that the multi-salt composite adsorbent reduces hysteresis between adsorption and desorption stages. The highest coefficient of performance (COP) and specific cooling power (SCP) were 0.461 and 328 W/kg, respectively, at an evaporation temperature of 25 °C and a charging temperature of 110 °C. Excessively long cycle times negatively affect system efficiency, with optimal performance achieved at a 40-min cycle. The findings highlight the system's effectiveness for industrial low-grade waste heat recovery, providing a solid foundation for future reactor design and material selection.

Sustainability of shear stress conditioning in endothelial colony-forming cells compared to human aortic endothelial cells to underline suitability for tissue-engineered vascular grafts

The endothelialization of cardiovascular implants is supposed to improve the long-term patency of these implants. In addition, in previous studies, it has been shown, that the conditioning of endothelial cells by dynamic cultivation leads to the expression of an anti-thrombogenic phenotype. For the creation of a tissue-engineered vascular graft (TEVG), these two strategies were combined to achieve optimal hemocompatibility. In a clinical setup, this would require the transfer of the already endothelialized construct from the conditioning bioreactor to the patient. Therefore, the reversibility of the dynamic conditioning of the endothelial cells with arterial-like high shear stress (20 dyn/cm2) was investigated to define the timeframe (tested in a range of up to 24 h) for the perseverance of dynamically induced phenotypical changes. Two types of endothelial cells were compared: endothelial colony-forming cells (ECFCs) and human aortic endothelial cells (HAECs). The results showed that ECFCs respond far more sensitively and rapidly to flow than HAECs. The resulting cell alignment and increased protein expression of KLF-2, Notch-4, Thrombomodulin, Tie2 and eNOS monomer was paralleled by increased eNOS and unaltered KLF-2 mRNA levels even under stopped-flow conditions. VCAM-1 mRNA and protein expression was downregulated under flow and did not recover under stopped flow. From these time kinetic results, we concluded, that the maximum time gap between the TEVG cultivated with autologous ECFCs in future reactor cultivations and the transfer to the potential TEVG recipient should be limited to ∼6 h.

Enhancing predictive monitoring of ethylene oxychlorination reactor states through spatiotemporal coupling analysis

The production of polyvinyl chloride (PVC) encounters challenges stemming from the temporal and spatial coupling characteristics inherent in the fixed bed ethylene oxychlorination process. Consequently, the implementation of enhanced safety measures and risk reduction strategies becomes imperative. This study introduces a pioneering methodology leveraging a spectral temporal graph neural network. By leveraging reactor temperature data, spatial variable decoupling facilitated by the Fourier transform, and a self-attentive mechanism within graph neural networks, the proposed approach adeptly forecasts future reactor states. The model's seamless alignment with the physical knowledge of reaction processes, validated through the adjacency matrix and hotspot region identification, underscores its efficacy in ensuring process safety and mitigating operational risks in PVC production. Empirical findings further validate the effectiveness of the approach, with predictions demonstrating an error margin of less than 0.5°C in forecasting future reactor temperatures.

Direct optimization of neoclassical ion transport in stellarator reactors

Abstract We directly optimize stellarator neoclassical ion transport while holding neoclassical electron transport at a moderate level, creating a scenario favorable for impurity expulsion and retaining good ion confinement. Traditional neoclassical stellarator optimization has focused on minimizing ϵ eff , the geometric factor that characterizes the amount of radial transport due to particles in the 1 / ν regime. Under expected reactor-relevant conditions, core electrons will be in the 1 / ν regime and core fuel ions will be in the ν regime. Traditional optimizations thus minimize electron transport and rely on the radial electric field (Er ) that develops to confine the ions. This often results in an inward-pointing Er that drives high-Z impurities into the core, which may be troublesome in future reactors. In this work, we increase the ratio of the thermal transport coefficients L 11 e / L 11 i , which previous research has shown can create an outward-pointing Er . This effect is very beneficial for impurity expulsion. We obtain self-consistent density, temperature, and Er profiles at reactor-relevant conditions for an optimized equilibrium. This equilibrium is expected to enjoy significantly improved impurity transport properties.

Overview of results from the 2023 DIII-D negative triangularity campaign

Abstract Negative triangularity (NT) is a potentially transformative configuration for tokamak-based fusion energy with its high-performance core, edge localized mode (ELM)-free edge, and low-field-side divertors that could readily scale to an integrated reactor solution. Previous NT work on the TCV and DIII-D tokamaks motivated the installation of graphite-tile armor on the low-field-side lower outer wall of DIII-D. A dedicated multiple-week experimental campaign was conducted to qualify the NT scenario for future reactors. During the DIII-D NT campaign, high confinement ( H 98 y , 2 ≳ 1), high current ( q 95 < 3), and high normalized pressure plasmas ( β N > 2.5) were simultaneously attained in strongly NT-shaped discharges with average triangularity δ avg = −0.5 that were stably controlled. Experiments covered a wide range of DIII-D operational space (plasma current, toroidal field, electron density and pressure) and did not trigger an ELM in a single discharge as long as sufficiently strong NT was maintained; in contrast, to other high-performance ELM-suppression scenarios that have narrower operating windows. These strong NT plasmas had a lower outer divertor X-point shape and maintained a non-ELMing edge with an electron temperature pedestal, exceeding that of typical L-mode plasmas. Also, the following was achieved during the campaign: high normalized density ( n e / n GW of at least 1.7), particle confinement comparable to energy confinement with Z eff ∼ 2 , a detached divertor without impurity seeding, and a mantle radiation scenario using extrinsic impurities. These results are promising for a NT fusion pilot plant but further questions on confinement extrapolation and core-edge integration remain, which motivate future NT studies on DIII-D and beyond.

Disruption runaway electron generation and mitigation in the Spherical Tokamak for Energy Production (STEP)

Abstract Generation of Runaway Electrons (REs) during plasma disruptions is of great concern for ITER and future reactors based on the tokamak concept. The current STEP (Spherical Tokamak for Energy Production) concept design features a plasma current higher than 20 MA, thus expected to be in a regime of very high avalanche multiplication. This is confirmed by modelling runaway electron generation in unmitigated disruptions using the code DREAM, with hot-tail generation found to be the dominant primary generation mechanism and avalanche confirmed to be extremely high. Varying assumptions for the prescribed thermal quench phase (duration, final electron temperature) as well as the wall time, the plasma-wall distance, and shaping effects, all STEP unmitigated disruptions are found to generate large runaway electron beams (from 10 MA up to full conversion). The possibility of RE avoidance is first studied by idealized, i.e. radially uniform, impurity injection of a mixture of argon and deuterium, with ad-hoc particle transport arising from the stochasticity of the magnetic field during the thermal quench (TQ). Unfortunately, no such injection scenario allows runaways to be avoided while respecting the other constraints of disruption mitigation. STEP disruption mitigation system (DMS) has then been tested with DREAM, by modelling 2-stage Shattered Pellet Injections consisting of pure D2 followed by Ar+D2. Such a scheme is found to reduce the generation of REs by the hot-tail mechanism, reducing the final RE current to about 13 MA, but isn't sufficient by itself. Options for mitigating such a RE beam (benign termination scheme), as well as estimations of required RE losses during the current quench (from a potential passive RE mitigation coil) will also be briefly discussed.

Light vector bosons and the weak mixing angle in the light of future germanium-based reactor CEνNS experiments

In this work, the sensitivity of future germanium-based reactor neutrino experiments to the weak mixing angle sin2θW, and to the presence of new light vector bosons is investigated. By taking into account key experimental features with their uncertainties and the application of a data-driven and state-of-the-art reactor antineutrino spectrum, the impact of detection threshold and experimental exposure is assessed in detail for an experiment relying on germanium semiconductor detectors. With the established analysis framework, the precision on the Weinberg angle, and capability of probing the parameter space of a universally coupled mediator model, as well as a U(1)B−L-symmetric model are quantified. Our investigation finds the next-generation of germanium-based reactor neutrino experiments in good shape to determine the Weinberg angle sin2θW with < 10% precision using the low-energetic neutrino channel of CEνNS. In addition, the current limits on new light vector bosons determined by reactor experiments can be lowered by about an order of magnitude via the combination of both CEνNS and EνeS. Consequently, our findings provide strong phenomenological support for future experimental endeavours close to a reactor site.

Analysis of a plasma reactor performance for direct nitrogen fixation by use of three-dimensional simulations and experiments

This study utilizes state-of-the-art in-situ measurements and advanced three-dimensional simulations to investigate the operation of a microwave plasma reactor for nitrogen fixation at a high subatmospheric pressure under various power inputs. It is shown that the system should be treated as a warm plasma due to negligible differences in vibrational and rotational temperatures and that it is required to account for chemical non-equilibrium. Complex flow field and significant temperature and concentration gradients are observed, emphasizing the need for three-dimensional simulations to accurately describe the interactions between mass, heat, and momentum transport and chemical reactions in non-equilibrium conditions. The innermost plasma regions reach temperatures close to 6000 K, while wall temperatures remain low due to presence of a low-temperature swirling flow field around the hot core. Analysis reveals that the swirling flow field around the hot core provides intrinsic partial quenching in regions with large temperature gradients which counteract the return to chemical equilibrium and enrich the flow surrounding the hot core. The inner flow exhibits a recirculation zone, stagnation point, and a transition into an increasingly parabolic flow profile closer to the reactor outlet. The agreement between the numerical modeling and the measurements is strong. In-situ Raman spectroscopy measurements confirm the calculated temperature field inside the reactor. Thermocouple measurements on the reactor wall are also in good agreement with the simulation. Additionally, Fourier-transform infrared spectroscopy measurements agree very well with the simulated amount of nitrogen oxide exiting the reactor at different microwave power inputs. The understanding and modeling capabilities are expected to support the development of future reactor designs that can facilitate higher yields.

Improving interfacial microstructure and mechanical properties of ODS-W/Cu joints via anodization treatment and spark plasma sintering

The bonding properties at the interface between the first wall oxide dispersion strengthened tungsten (ODS-W) and the heat sink copper (Cu) are vital for their application in future nuclear fusion reactors. However, the immiscibility and significant differences in the coefficient of thermal expansion between ODS-W and Cu pose considerable challenges for bonding these two materials. This study utilizes anodization and hydrogen reduction processes to form a nanoporous structure on the surface of ODS-W, thereby achieving effective bonding between ODS-W and Cu by spark plasma sintering (SPS). The effects of the anodization parameters on the surface characteristics of ODS-W, and the influence of the bonding temperature on the interfacial microstructure and mechanical properties of the ODS-W/Cu joints were investigated. The results demonstrate that under an anodization condition of 50 V for 40 min, a nanoporous structure with an average pore size of about 90 nm can forms on the surfaces of ODS-W. This significantly increases the surface roughness, thereby enhancing the interfacial bonding of ODS-W with Cu during SPS. As the bonding temperature increases, the interfacial microstructure changes from linear to serrated shape, effectively suppressing crack propagation and forming a stronger mechanical interlock between ODS-W and Cu. This significantly enhances the mechanical properties of the joints. Typically, at a bonding temperature of 1000 °C, the tensile strength of the anodized ODS-W/Cu joints reaches 245.2 MPa, which is 157.9 % higher than that of directly bonded joints. Additionally, the tensile fracture primarily occurs within the Cu substrate, transitioning from brittle to ductile fracture modes.