Magnetic Fusion Reactor Research Articles

Tokamak edge-SOL turbulence in H-mode conditions simulated with a global, electromagnetic, transcollisional drift-fluid model

Abstract The design of commercially feasible magnetic confinement fusion reactors strongly relies on the reduced turbulent transport in the plasma edge during operation in the high confinement mode (H-mode). We present first global turbulence simulations of the ASDEX Upgrade tokamak edge and scrape-off layer in ITER baseline H-mode conditions. Reasonable agreement with the experiment is obtained for outboard mid-plane measurements of plasma density, electron and ion temperature, as well as the radial electric field. The radial heat transport is underpredicted by roughly 1/3. These results were obtained with the GRILLIX code implementing a transcollisional, electromagnetic, global drift-fluid plasma model, coupled to diffusive neutrals. The transcollisional extensions include neoclassical corrections for the ion viscosity, as well as either a Landau-fluid or free-streaming limited model for the parallel heat conduction. Electromagnetic fluctuations are found to play a critical role in H-mode conditions. We investigate the structure of the significant E × B flow shear, finding both neoclassical components as well as zonal flows. But unlike in L-mode, geodesic acoustic modes are not observed. The turbulence mode structure is mostly that of drift-Alfvén waves. However, in the upper part of the pedestal, it is very weak and overshadowed by neoclassical transport. At the pedestal foot, on the other hand, we find instead the (electromagnetic) kinetic ballooning mode, most clearly just inside the separatrix. Our results pave the way towards predictive simulations of fusion reactors.

Growth of α-Al2O3 layer involving abnormal oxides in FeCrAl alloy tube fabricated by WEDM process and electrical insulating performance in fusion reactor blanket

The tubular specimens of FeCrAl alloy APMT (Fe-22Cr-5Al-3Mo) were fabricated by a wire electrical discharge machining (WEDM) process. The oxidation characteristics of the machined surface were investigated by means of the oxidation tests performed in air atmosphere at the temperatures of 1273 K and 1373 K up to 100 h. The abnormal oxides such as CuAl2O4 and ZnAl2O4 were grew on the recast layer which was locally remained on the machined surface even after the mechanical polishing procedure. The α-Al2O3 layer involving the abnormal oxides was formed on the machined surface in the oxidation treatment at 1273 K for 10 h. The apparent electrical conductivity of the layer involving the abnormal oxides was larger than that without the abnormal oxides. Thick Al-rich oxide layer was formed under the abnormal oxides when the oxidation treatment was performed at 1373 K. The apparent electrical conductivity of the oxide layer became lower due to the presence of the thick Al-rich oxide. This oxide layer formed at 1373 K could function as an electrical insulating layer to reduce the magnetohydrodynamic pressure drop in liquid breeder blanket systems of magnetic confinement fusion reactors.

Measurement of the Gamma-Ray-to-Neutron Branching Ratio for the Deuterium-Tritium Reaction in Magnetic Confinement Fusion Plasmas.

At present, magnetic confinement fusion devices rely solely on absolute neutron counting as a direct way of measuring fusion power. Absolute counting of deuterium-tritium gamma rays could provide the secondary neutron-independent technique required for the validation of scientific results and as a licensing tool for future power plants. However, this approach necessitates an accurate determination of the gamma-ray-to-neutron branching ratio. The gamma-ray-to-neutron branching ratio for the deuterium-tritium reaction ^{3}H(^{2}H,γ)^{5}He/^{3}H(^{2}H,n)^{4}He was determined in magnetic confinement fusion plasmas at the Joint European Torus in predominantly deuterium beam heated plasmas. The branching ratio was found to be equal to (2.4±0.5)×10^{-5} over the deuterium energy range of (80±20) keV. This accurate determination of the deuterium-tritium branching ratio paves the way for a direct and neutron-independent measurement of fusion power in magnetic confinement fusion reactors, based on the absolute counting of deuterium-tritium gamma rays.

Investigation of MHD flow balancing in electrically conductive manifolds of liquid metal blankets

In the liquid metal (LM) blanket design for magnetic confinement fusion reactors, manifolds including main and branch ducts are commonly used to direct fluids to various locations of the blanket to cool its components. The primary objective of this structural configuration is to achieve flow balancing within each branch duct. However, magnetohydrodynamic effects will cause uneven flow distribution, consequently, uncontrollable overheating phenomena in certain blanket segments may be caused by this flow distribution. To assess its impact, the influences of external magnetic fields, inlet velocities and coupling wall thicknesses on the flow partitioning in conductive manifolds are investigated in this study. The results indicate that the flow distribution among the branch ducts becomes less uniform as the external magnetic field strength increases, while varying the inlet velocity has minimal impact. In addition, increasing the coupling wall thickness leads to more even flow distribution when the rectangular branch ducts are perpendicular with the magnetic field direction. Based on above physical understanding, three controlling methods for the flow partitioning in conductive manifolds are proposed to achieve flow balancing in each branch duct and then validated through numerical simulations. The results indicate that flow balancing can be achieved by adjusting the coupling wall thickness when the rectangular branch ducts are aligned with the magnetic field direction, however, it is challenging when they are perpendicular. Furthermore, the flow balancing can be realized by a symmetrical design of the manifold with circular branch pipes and it is independent of external magnetic fields and inlet velocities. These findings will enhance the physical understanding of flow distribution patterns in conductive manifolds under strong magnetic fields, which is important for the flow control and design of manifolds in LM blankets.

High pressure phase transition and its influence on structural, mechanical properties and elastic anisotropy of [formula omitted]: An ab-initio study

Lithium carbide (Li2C2), a promising tritium breeding material with potential applications in magnetic confinement-based fusion reactor blankets, has drawn significant attention due to its recent utilization in lithium-ion batteries. In the current study, we employed first-principles calculations with van der Waals (vdW) force correction to explore the structural and mechanical properties of Immm and Pnma phases of Li2C2 subjected to external hydrostatic pressure. It is demonstrated that inclusion of vdW effects accurately predicts Immm→Pnma phase transition at 16 GPa, aligning closely with experimental transition pressure. In contrast, PBE-GGA calculations underestimate the same by about 21%. Current paper is the first comprehensive study of pressure evolution of mechanical properties and strength determining parameters of Li2C2. Interestingly, this inherently brittle material exhibits ductile characteristics beyond 7.5 GPa. The paper also offers a quantitative comparison of its strength parameters with other prominent tritium breeding materials. Finally, the strong directional dependence due to the presence of C2 dimer is investigated by analyzing the pressure variation of elastic anisotropy. The study has its relevance in assessing structural integrity of Li2C2 during inadvertent buildup of tritium and helium that would generate stress in the solid breeder.

An improved analysis method for assessing the nuclear-heating impact on the stability of toroidal field magnets in fusion reactors

An improved analysis method for assessing the nuclear-heating impact on the stability of toroidal field magnets in fusion reactors

Impact of Reverse Buoyancy on Heat and Mass Transfer in MHD Flow of PbLi Within a Poloidal Duct in a Typical DCLL Liquid Breeder Blanket

The typical dual-coolant lead-lithium (PbLi) design of a liquid breeder blanket in a magnetic confinement fusion reactor involves the utilization of PbLi as the working fluid to effectively remove neutron heat. However, the nonuniform heating of neutrons with a significant radial gradient induces a buoyancy effect, resulting in the formation of vortexes ices within the downward flow duct. These vortexes have an adverse impact on the heat and mass transfer characteristics of the magnetohydrodynamic (MHD) flow of PbLi. The simulations in this work employed a MHD buoyant mixed-convection solver to resolve the characteristics of PbLi flow and a one-way coupled Lagrangian method to analyze the qualitative characteristics of tritium transport in PbLi flow. The results indicate that buoyant reverse flow can create vortexes that contain hot spots in the PbLi fluid, which can significantly impede heat transport. Additionally, the vortex causes tritium recirculation in the flow field and retention, resulting in adverse effects on tritium transport.

One-dimensional simulation and validation of divertor detachment induced through nitrogen seeding on HL-2A

Divertor detachment is a promising method to solve the power exhaust problem in tokamak devices or even in future magnetic fusion reactors. In this work, a detachment experiment (HL-2A shot #38008) with mixed gas seeding (60% nitrogen and 40% deuterium) is simulated using the SD1D module in BOUT++. In the process from attachment to detachment, the target electron temperature and the target ion saturation current in simulations are found to be consistent with the experimental results measured by Langmuir probes on the target plate. In order to understand the underlying detachment mechanism on HL-2A, this work analyses the role of different particle species in the cases with different seeding rates. It shows that the plasma density varies little and the density of neutrals ( D and D2 ) slightly decreases with the increase of nitrogen seeding rate, such that plasma–neutral interactions cannot effectively reduce plasma energy and plasma momentum in the divertor. The case with a high seeding rate predicts that increasing seeding rate cannot reach a target temperature lower than 2.5eV , which is the required temperature for strong plasma–molecule interactions. Thus, the plasma–molecule interactions may not be important in the divertor during nitrogen seeding. This work also studies the parallel forces including (1) force due to the parallel electric field, (2) friction forces, (3) ion- and electron-thermal forces, and (4) collisional reactions (e.g. charge exchange recombination and ionisation). It is found that the friction force between nitrogen ions and other particle species (primarily D+ ) is the dominant force towards the target, while ion- and electron-thermal forces are the dominant force pushing nitrogen ions back to upstream. Parallel forces determine the parallel distribution of nitrogen impurities, and therefore decide the region of nitrogen radiation.

Deuterium reclamation from C-Si codeposits using thermo-oxidation

SiC exhibits a remarkable resistance to neutron irradiation damage and, being a low-Z material, is seen as a potential material candidate for plasma-facing components in magnetic confinement fusion reactors. The current work investigates the reclamation of deuterium from C-Si codeposits produced by sputter-deposition at temperatures from 300 K to 700 K using thermo-oxidation at 350 °C and 400 °C (623 or 673 K). While initial D content was found to be close to that of similarly-produced pure carbon codeposits, hydrogen removal from the investigated C-Si codeposits was significantly reduced comparatively. This reduced removal behaviour was found to follow closely with that observed for B-C codeposits from the DIII-D tokamak. This suggests that hydrogen removal via thermo-oxidation, where a non-carbon inclusion is known to form a stable oxide, is unlikely to be effective.

Thermo-hydraulic performance of tungsten flat tile type divertor target with two types of novel engineered cooling channel wall

The divertor is one of the key components in magnetic confinement fusion reactors. Improving the heat transfer performance of the divertor has great importance for promoting its service capacity. This work proposed a new kind of cooling surface configuration for a flat-type divertor design, which is based on the concept of synergetic heat transfer enhancement via combination of a sinusoidal undulation structure with a micro-nano structured surface (copper-mesh brazed coating). The combination surface configuration is preliminarily obtained by numerical optimization based on the calibrated model of subcooled flow boiling heat transfer. Then, experimental mockups were fabricated and tests with high heat flux (HHF) conditions were conducted to examine the effects of the new design. Both the numerical and experimental results confirm that the incident critical heat flux is increased from 15 MW/m2 of the micro-nano structured flat surface case to over 22 MW/m2 of the combination case under the conditions of relatively low flow velocity of 3 m/s and low flow resistance increase of 7.3 %.

Study of irradiation-induced vacancy-type defects in tungsten using positron annihilation spectroscopy

Tungsten has been considered as a potential candidate material for the application of plasma-facing components in magnetic confinement fusion reactors. Fusion neutrons introduce severe displacement damage into the first-wall materials along with a high amount of helium which is detrimental to the desirable material properties. Ion beam implantation combined with positron annihilation spectroscopy is a well-adopted method to investigate the irradiation damage in materials without any induced radioactivity. In this study, two separate tungsten samples were individually irradiated with 9.1 MeV Au5+ ions to a dose of 2.5 × 1014 ions/cm2 and 130 keV He ions to a dose of 1 × 1016 ions/cm2 with the intention to investigate the evolution of vacancy-type defects in the absence and presence of helium. The irradiated samples were subjected to isochronal annealing and characterized using positron annihilation Doppler broadening spectroscopy. The evolution of vacancy-type defects during different stages of annealing was investigated using defect-sensitive positron S-parameter. An increase in S-parameter due to the presence of vacancy-type defects was observed upon irradiation in both samples. The S-parameter variations with respect to the temperature showed different stages of vacancy annealing corresponding to stage III from 473 K to 673 K where the migration and clustering of mono-vacancies occurs, stage IV from 673 K to 973 K due to the growth of nano-voids and stage V from 973 K to 1273 K due to the dissociation of nano-voids. Complete recovery of irradiation-induced defects occurred in both samples at 1273 K. Both the samples showed similar vacancy annealing stages and no signature on the formation and growth of He-vacancy clusters or He bubbles was observed in the present study.

HTS conductor coil by in-situ winding technology for large-scale high-field magnet

High temperature superconducting (HTS) conductors have become important candidates to be used in large-scale high-field magnets, owing to its high critical current density as well as good mechanical properties. At present, there are various forms of HTS conductors proposed. In this work, we reported the fabrication of the REBCO HTS coil using in-situ winding technology. The six-layer coil with a diameter of 410 mm achieved a total current of 2562 A, and generated a central magnetic field of 7.8 mT at 77 K. The stress, deformation, and defect caused during conventional winding procedure could be significantly decreased using in-situ winding technology, indicating it has great potential for the large-scale high-field magnets in next-generation fusion reactors.

Atomic collisional data for neutral beam modeling in fusion plasmas

The injection of energetic neutral particles into the plasma of magnetic confinement fusion reactors is a widely-accepted method for heating such plasmas; various types of neutral beam are also used for diagnostic purposes. Accurate atomic data are required to properly model beam penetration into the plasma and to interpret photoemission spectra from both the beam particles themselves (e.g. beam emission spectroscopy) and from plasma impurities with which they interact (e.g. charge exchange recombination spectroscopy). This paper reviews and compares theoretical methods for calculating ionization, excitation and charge exchange cross sections applied to several important processes relevant to neutral hydrogen beams, including H + Be4+ and H + H+. In particular, a new cross section for the proton-impact ionization of H (1s) is recommended which is significantly larger than that previously accepted at fusion-relevant energies. Coefficients for an empirical fit function to this cross section and to that of the first excited states of H are provided and uncertainties estimated. The propagation of uncertainties in this cross section in modeling codes under JET-like conditions has been studied and the newly-recommended values determined to have a significant effect on the predicted beam attenuation. In addition to accurate calculations of collisional atomic data, the use of these data in codes modeling beam penetration and photoemission for fusion-relevant plasma density and temperature profiles is discussed. In particular, the discrepancies in the modeling of impurities are reported. The present paper originates from a Coordinated Research Project (CRP) on the topic of fundamental atomic data for neutral beam modeling that the International Atomic Energy Agency (IAEA) ran from 2017 to 2022; this project brought together ten research groups in the fields of fusion plasma modeling and collisional cross section calculations. Data calculated during the CRP is summarized in an and is available online in the IAEA’s atomic database, CollisionDB.

Construction Strategy for the Future K-DEMO Superconductor Test Facility

The 16 T superconducting magnet for the next generation fusion reactor, called K-DEMO would be a great engineering challenge with current technology. The project on the test facility, named SUCCEX (Super Conducting Conductor Experiment) is prerequisite for the development of K-DEMO superconducting magnet. It would generate an external magnetic field of 16 T for developing and testing K-DEMO superconductors. Also, this project will give a guideline to fabricate future superconducting conductors for the next generation nuclear fusion reactor construction. It will be an essential facility for the production of superconducting magnets for K-DEMO. In 2021, the Korean government has approved this project and the project is in its early stages of construction. The construction strategy and the current progress are discussed in this paper.

Modeling of subcooled flow boiling for hypervapotron in divertor of fusion reactor based on RPI model

The divertor is the key in-vessel component to remove the high heat flux and impurities in magnetic confinement fusion reactor. Hypervapotron (HV) has been developed as one of the candidates for cooling structure of divertor, owing to its excellent heat transfer performance in subcooled flow boiling. To address the design and development requirements of divertor, a suitable model for single-phase and two-phase flow of HV are developed. Four combination models of boiling properties expressed by different empirical correlations are evaluated. The results indicate that the LC-UN (Lemmert-Chawla, Unal) model shows a good agreement with the experimental results of Lim et al. (2022), with average deviation of 8.15%. Moreover, for the first time, the model successfully predicts the onset of fully developed nucleate boiling (OFDB) point of HV. Based on the LC-UN model, the detailed heat transfer and flow characteristics of HV are revealed. It shows that the boiling process is independent between each slot due to the fin barrier, and the bubbles are condensed by the subcooled main stream after leaving the wall. The “double vortex” in HV intensifies the interaction between vapor and liquid phase, preventing the aggregation of bubbles, allowing it has excellent heat transfer performances.

Study on the effect of neutron shielding design on tritium breeding based on water-cooled ceramic blanket for CFETR

The Chinese Fusion Engineering Test Reactor (CFETR) is a magnetic confinement fusion reactor independently designed and developed by China, which is based on the ITER design experience. As one of the candidates for the CFETR, the water cooled ceramic blanket (WCCB) mainly carries out many important tasks, such as tritium breeding and radiation shielding. The high tritium breeding ability is one of the most significant goals of the blanket. For the design of the CFETR, in addition to meeting the requirements of tritium breeding, the design of the blanket must also consider meeting relevant shielding limits. Due to the limited space of the blanket, the improvement in tritium breeding space will inevitably lead to the reduction in neutron shielding space behind the breeding area, and vice versa. Therefore, under the premise of meeting the requirements of neutron shielding, increasing the tritium breeding space as much as possible is the focus of research. In this work, a three-dimensional neutronics model containing the WCCB blanket is developed, and the neutronics performance is calculated based on 1.5 GW fusion power. A set of nuclear analyses are carried out by the MCNP code, including analysis of the neutron wall load, tritium breeding ratio (TBR), and fast neutron fluence of the TF coil. It is found that the shielding space of certain blanket modules could be optimized. After the shielding optimization, the global TBR increased from 1.168 to 1.186, an increase of 0.018 TBR. The current research has important guiding significance for the future design and optimization of the WCCB for the CFETR.

Development of pebble-based extruded carbon rods for extreme plasma heat flux environments

This work presents first experiments toward the development of continuously renewable (extrudable) pebble-based carbon rods for use as plasma-facing components in extreme steady-state plasma flux environments. The primary envisioned application of this work is a first wall that can survive long-term in future magnetic fusion power reactors while also improving recovery of the reactor fuel (tritium and deuterium atoms). Bench tests applying extreme steady-state front-surface heat loads of up to 50 MW/m2 are presented. Continuous pebble rod front-surface recession and intact pebble recovery are successfully demonstrated, at a rate of order 0.2 cm/s. Numerical simulations of the pebble rod front-surface recession are able to match observations reasonably well, indicating that the recession mechanism can be understood as occurring due to pebble thermal expansion and resulting shock and cracking of the inter-pebble binder. Tests of the pebble rod extrusion demonstrate that friction between the rods and the stainless steel extrusion channel is tolerably low (<50 N for the expected channel length) over a wide range of temperatures. Front-surface outgassing rates below 1000 Torr L/s/m2 are achieved, believed to be sufficiently low for use in magnetic fusion reactors. Initial parametric scans over pebble rod size and binder fraction to vary front-surface recession rates are presented.

Impact Assessment of Radiative Heat Transport in ARC-Class Reactor FLiBe Liquid Immersion Blanket

Molten salt systems have become of growing interest in the energy industry due to a wide range of applications (concentrated solar power systems, energy storage, Generation IV fission reactors, and high magnetic field fusion reactors). Because of the high temperature that characterizes such materials, radiative heat transfer (RHT) may become a nonnegligible heat transfer mechanism in molten salt components. In this paper, an investigation of FLiBe RHT has been conducted, with a focus on Affordable, Robust, Compact (ARC)–class fusion reactors, a preconceptual design proposed by Commonwealth Fusion Systems and the Plasma Science and Fusion Center. This class of reactors largely employs FLiBe molten salt due to its thermal and neutronic properties. The reactor is characterized by high temperatures, and its 0.5-m-thick liquid immersion blanket is a component where RHT contribution to the temperature distribution is yet to be evaluated. Therefore, this study is the first work that quantifies the contribution of RHT in ARC-class reactor FLiBe systems. FLiBe optical property spectral-banding assessment is carried out, and the impact of RHT in FLiBe systems is assessed in operational ARC-class scenarios through computational fluid dynamics models by taking advantage of COMSOL® Multiphysics. Heat transfer, thermal-dependent properties, and buoyancy effects are considered in a comparison between scenarios with and without RHT modeling. The flow field in the tank is unaffected by RHT effects, even when considering buoyancy effects. The external layer of the vacuum vessel shows an average decrease in the temperature of 5.4 K and an average decrease in temperature on the surface in contact with the FLiBe tank of 8.1 K. Results indicate that for ARC-class reactors, RHT phenomena are negligible (<1% increase in heat transfer) in operational conditions.

Effects of thickness and Ti substitutions on the anti-disproportionation behavior of ZrCo alloy films

ZrCo alloy has been considered as a tritium storage material to replace uranium for large-scale hydrogen isotope storage and delivery in tritium plant of the magnetic confinement nuclear fusion reactor. However, the disproportionation of ZrCo greatly obstructed its aspects to engineering application. A multifaceted understanding about the intrinsic nature of disproportionation reaction, and proposing anti-disproportionation strategies from material design have always been the bottleneck and hotspot in the tritium science and technology field. Herein, we studied the anti-disproportionation properties of Zr(Ti)Co alloy with thin film structure, and the effects of film thickness and Ti substitutions on the anti-disproportionation properties of the films were obtained. Firstly, a series of ZrCo films with different deposited time were prepared by magnetron sputtering instrument. The disproportionation ratio of the 2 h deposited ZrCo film (1.641μm) is 23.9%, and attenuated with the increase of film thickness. It indicates that the anti-disproportionation properties of the film structure is significantly better than that of ZrCo particle (68%). The microtopography analysis shows that the micro stress in the film structure is beneficial to hinder the disproportionation process. However, the hydrogenation reaction is also obviously inhibited in the ZrCo film. In addition, Ti doping can further improve its anti-disproportionation performance (10.3–7.3% for the 2 h deposited Zr1-xTixCo film) without significantly reducing the hydrogen absorption capacities of the ZrCo flims. The anti-disproportionation performance is closely related to the weakened interaction between Zr and H atoms, after Ti doping into the ZrCo flim, leading to the impaired driving force for disproportionation reaction. The study in this work can provide an important reference for the anti-disproportionation modification of ZrCo alloys.

Refractive index measurement of hydrogen isotopologue mixture and applicability for homogeneity of hydrogen solid at cryogenic temperature in fusion fuel system

Deuterium (D)-Tritium (T) nuclear fusion reaction has potential as an energy source in the future. In both magnetic confinement and inertial confinement fusion reactors, solid D–T will generally be supplied as fusion fuel. The efficiency of the nuclear fusion reaction depends on the quality of solid D–T fuel, which is related to the composition, homogeneity, helium-3 (3He) content, and so on. However, there is no technique for in-situ examination of solid D–T fuel. In this study, we consider a simple and precise method for the characterization of solid hydrogen isotopologues at cryogenic temperature using refractive index measurement, and evaluate the distribution of hydrogen isotopologue composition and homogeneity. To evaluate without the effect of tritium decay, the homogeneity of the hydrogen (H2)-deuterium (D2) mixture is measured at first. By the in-situ refractive index measurement at cryogenic temperature, the homogeneity of solid H2–D2 mixture is roughly quantified. The phase diagram of the H2–D2 mixture shows a solid solution type. D2-rich crystal first appears from the liquid phase as a primary crystal. The composition of D2 in liquid phase ias homogeneous, whereas it reduces by obeying the liquidus line in the phase diagram with the crystallization. On the other hand, the composition of the H2–D2 mixture in solid phase is inhomogeneous because the mobility of H2 and D2 in solid phase was too slow to be homogeneous and solid. The compositions of H2–D2 mixture in liquid and solid phases could be evaluated by the in-situ refractive index measurement in time. Consequently, the refractive index measurement shows great potential as an inspection method of solid D–T fuel in fusion reactors.