Xiaoyu Wang

Among the different breeding blanket concepts proposed by various countries for DEMO design, China ultimately decided to develop and test the Helium-Cooled Ceramic Breeder Test Blanket Module (HCCB TBM) in ITER, so as to align with the national strategy for magnetic confinement fusion energy development. During the preliminary design phase, some design updates were undertaken by the CN HCCB TBM. taking into account its engineering performance and manufacturing feasibility, However, considering certain engineering and technical issues, a new ITER baseline has been under development since early February 2023, which proposed a new ITER operation strategy to initiate the nuclear phase as soon as possible and reduce some operational parameters at initial stages. Regarding the ITER 2024 new baseline scenario, further design optimization and engineering analysis were carried out for the CN HCCB TBM. The thermal-hydraulic analysis results show that, the maximum operating temperatures of structural and functional materials have been significantly reduced for the 2024 baseline scenario, especially for the tritium breeder Li4SiO4 pebble bed, whose temperature decreased by 20.6%. By adding electric heaters inside the tritium breeder zone, the maximum temperature of Li4SiO4 can be raised to 821°C, and the average temperature raised o 604°C, essentially meeting the tritium release temperature requirements. Based on the temperature distributions calculated under three operating conditions, a system-level transient tritium transport analysis was further performed. The results indicated that adding electric heaters inside the tritium breeder region is sufficient to meet the tritium balance requirements, without the need to further add electric heaters in the neutron multiplier region. A preliminary structural analysis was also conducted, and the results indicated that both the primary and secondary stresses were well below the limits of the structural material, which ensures the integrity of the overall structure.

Tritium self-sufficiency is a critical prerequisite for future fusion reactors. The tritium breeding blanket, as the component responsible for in-vessel tritium generation, requires coordinated neutronic and engineering optimization in order to maximize its achievable tritium breeding ratio (TBR). In this work, a high-fidelity 22.5° toroidal sector neutronics model of the China Fusion Engineering Test Reactor (CFETR) equipped with a Helium-Cooled Ceramic Breeder (HCCB) blanket was established. On this basis, we developed a Multi-physics Coupling Intelligent Neutronic Optimization code (MCINO), a two-stage neutronics optimization framework that combines global exploration by simulated annealing with subsequent local refinement. The objective was to maximize the global TBR by optimizing the radial distribution of breeder (Li ₄ SiO ₄ ) and neutron multiplier (Be) zones. The optimized design increased the global TBR to approximately 1.193, corresponding to an 8.36% improvement over the initial configuration. The improvement is associated with a more effective radial allocation of breeding and multiplying materials, which enhances neutron moderation, multiplication, and use for tritium production. The optimization workflow was designed to reduce the number of expensive high-fidelity transport recalculations, thereby improving computational efficiency relative to direct brute-force search. Finally, the engineering feasibility of the optimized design was checked through three-dimensional thermal–hydraulic verification, which confirmed that the representative modules remained within their prescribed operating limits. The present work provides an efficient and physically transparent framework for integrated blanket neutronics design and optimization.

Lithium-containing ceramics were significant tritium breeders for the fusion blanket concept, for which tritium release performance and mechanical properties serve as the core indicators for evaluating their performance as tritium breeders. The Li₄Si_0.8Ti_0.2O₄ material was designed as an advanced tritium breeder and fabricated into ceramic pebbles via the freeze-drying method. The tritium release properties of the Li₄Si_0.8Ti_0.2O₄ sample pebbles were investigated via temperature-programmed desorption (TPD). The mechanical properties of the same batch of tritium breeder pebbles were analyzed comparatively, specifically examining the change in their compressive strength before and after irradiation. The sample pebbles irradiated with different neutron doses show different tritium release characteristics, and the tritium release temperature was about 293–553 °C. This was due to the H₂-tritium isotope exchange reaction, and radiation with different neutron doses will lead to different release temperatures of tritium. The mechanical properties of the Li₄Si_0.8Ti_0.2O₄ ceramic pebbles decreased significantly after irradiation. The main reason was that the accumulation of lattice defects and helium bubbles produced by high-energy neutron irradiation leads to internal cracks and helium embrittlement in the material. These results indicate that Li₄Si_0.8Ti_0.2O₄ solid solution may be considered a potential candidate for tritium breeder materials.

In a tritium-breeding blanket of a fusion reaction, helium, used as a tritium-purging gas, will purge the tritium breeder pebble beds to extract the tritium in blanket. The purge gas flow characteristics will affect the tritium extraction efficiency. The effect of the fixed wall on the pebble packing structures and purge gas flow characteristics was investigated by combining the discrete element method (DEM) and computational fluid dynamics (CFD) method. The results indicate that the fixed wall leads to a regular packing of the pebbles adjacent to the fixed wall in association with drastic fluctuations in the porosity of the pebble bed, which can affect the purge gas flow behaviors. Further analyses of helium flow behaviors show that the helium pressure in the pebble bed decreases in a linear manner along the flow direction, whereas the pressure drop gradient of helium increases gradually with an increase in the packing factor. The reduction in porosity in the pebble bed leads to a notable escalation in helium flow velocity. Concerning the direction perpendicular to the helium gas flow, the evolution of the cut-plane averaged velocity of helium is similar to that of the porosity, except in the region immediately adjacent to the wall. The pressure drop and flow characteristics obtained in this study can serve as input for the thermohydraulic analysis of the tritium blowing systems in the tritium-breeding blanket of a fusion reactor.

The flow characteristics of purge gas helium in the pebble bed of the tritium breeding blanket are important in analyzing the tritium purging process and optimizing the design of the solid breeder blanket. Therefore, the flow characteristics of helium gas in randomly packed pebble beds are investigated experimentally with a focus on the analysis of the pressure loss distribution. The results show that gas velocity, bed dimension, and pebble diameters have an obvious influence on the helium flow characteristics in pebble beds. With the increase in the inlet helium gas velocity, the pressure drop gradient of helium in the pebble bed gradually increased. With increases in the pebble bed dimension, the pressure drop gradient of helium in the pebble bed gradually increased. With the increase in the pebble diameter, the pressure drop gradient gradually decreased. In addition, the effect of temperature on the pressure drop of helium in the pebble bed was also preliminarily investigated. The pressure drop gradient of the helium through the pebble bed obviously increased with the increase in the helium and the bed temperature. The experimentally obtained pressure loss characteristics can be used for the validation of the simulation of a blanket pebble bed and as input parameters in the thermo-hydraulic analysis of solid-tritium breeder blankets.

The packing characteristics of a pebble bed are essential to understand the heat- and mass-transfer processes occurring within a granular system. Therefore, the packing characteristics of rectangular prismatic pebble beds randomly packed with mono-sized pebbles are analyzed. In terms of the average and local packing fraction distribution, coordination number, and radial distribution function, the effects of the rectangular tube aspect ratio and cross-sectional area on the packing properties of pebble beds are explored in depth. The findings indicate that the packing structures of the rectangular pebble bed exhibit noticeable fixed-wall effects. The average packing fraction and coordination number gradually decline as the rectangular tube aspect ratio rises. Close to the fixed wall, a noticeable wall effect can be seen in the distribution of axial and local packing fractions and the pebble center distribution. The wall effect has an increasing effect on the axial and local packing fraction distributions in rectangular tubes with increasing aspect ratios. Additionally, the average packing fraction and the average coordination number also increase as the cross-sectional area increases, indicating a gradual weakening of the wall effect as the cross-sectional area increases. Furthermore, the cross-sectional area and aspect ratio of the rectangular tubes affect the RDF values of the rectangular pebble beds but have no impact on the RDF features. The findings reported in this paper will be helpful for designing and optimizing pebble beds in the breeding blanket of fusion reactors.

The first wall of the helium-cooled ceramic breeder test blanket module (TBM) will be subjected to charge exchange neutral from the mixed deuterium (D) and tritium (T) deuterium–tritium (D–T) mixture plasma irradiation. To understand the isotope effects on mixed plasma-driven permeation (PDP) through the TBM wall, which is typically made by reduced activation martensitic/ferritic steel (RAFM), mixed D and protium (H) PDP experiments for a Chinese RAFM CLF-1 have been performed. Permeation fluxes of H and D through ∼mm thick CLF-1 have been recorded at various sample temperatures and particle incident energies. The co-permeation of H and D is found to be diffusion-limited in the bulk and obeys classical mass effects of diffusion species. The H/D ratio of the permeation fluxes is found to be stable when changing sample temperature and is insensitive to ion incident energy. In addition, the mixed H can provide more channels for D recombination to reduce the D permeation.

Pre-depositing a Pd film is crucial for accurately acquiring the hydrogen permeability of metal materials, as it permits the production of ultra-pure hydrogen. However, the microstructure of Pd film and its effect on the hydrogen isotope permeation behavior of substrate materials have been neglected. In this study, Pd films were deposited on China Low-activation Ferritic (CLF−1) steel by magnetron sputtering. The effect of sputtering pressure on the microstructure and deuterium permeation behavior of Pd films at temperatures of 550−650 °C is presented. SEM results demonstrated that the films had a columnar crystal structure with a thickness of 0.6 ± 0.2 μm. The gas-driven permeation results revealed that the deuterium ion current intensity of the coated CLF−1 sample was at least three times lower than that of uncoated CLF−1 steel, which was influenced by the combined effect of oxygen and surface cracks. Oxygen could not be excluded from the films at a sputtering pressure of 10⁻³ Pa order of magnitude. It was also found that the films cracked during deuterium permeation experiments, which affected the deuterium permeation behavior. Films with large surface coverage and small grain sizes exhibited better cracking resistance. Our study provides promising insights into the hydrogen permeability of Pd films.

The hydrogen permeation behavior of novel AlCrFeTiNb, AlCrMoNbZr and AlCrFeMoTi high-entropy alloy (HEA) coatings were investigated. The hydrogen permeability of HEA coatings prepared by magnetron sputtering technology were tested using gas-driven deuterium permeation and electrochemical hydrogen permeation methods. The gas-driven permeation results show that the deuterium permeation resistance of the AlCrFeTiNb coating is the worst because of the unstable structure at a high temperature. Scanning electron microscope (SEM) and X-ray Diffraction (XRD) analysis proved a loose surface morphology of the AlCrFeTiNb coating and demonstrated the formation of iron-based oxides after deuterium permeation experiments. A high content of iron in HEA coating is disadvantageous for improving the hydrogen permeability. Differently, electrochemical hydrogen permeation reveals that the AlCrMoNbZr coating could resist hydrogen permeation better in a corrosive environment (0.2 mol/L KOH solution). The AlCrFeMoTi coating was peeled off after an electrochemical hydrogen permeation test due to the poor corrosion resistance. The hydrogen behavior of HEA coatings was discussed in detail. Our study provides a promising thought on hydrogen permeation of HEA coatings.