Tritium Breeding Research Articles

The scoping, design, and plasma physics optimization of the Eos neutron source stellarator

Abstract On the path to a fusion pilot plant, Thea Energy plans to build Eos, a sub-breakeven, deuterium-deuterium, beam-target fusion, stellarator neutron source facility for producing tritium and other valuable radioisotopes. In this paper, a set of 1D plasma physics models are coupled and used to design the operating point of the facility and predict performance. At this foundational stage of the design, analytic and approximate models are sufficient to capture the leading-order effects, and fast enough to run in the inner loop of an optimizer. Higher-fidelity analyses will follow. Models of 1D profile-dependent neutral beam stopping, ion beam slowing down, beam-target fusion, electron-ion classical heat transfer, energy confinement (ISS04), beam pressure, beam heating of ions and electrons, beam-beam fusion fraction, and neutral beam injection and gyrotron heating electrical efficiencies are included. A numerical optimizer is used to determine the minimum required facility electric power to generate tritium at a given rate. A potentially advantageous regime is described in which modern precisely-quasisymmetric stellarators, new high-temperature superconductors, ITER-derived neutral beam injection, and new high-frequency gyrotrons enable a suitible target plasma with hot electrons, cold ions, peaked density and temperature profiles, and high beam-injected ion density. It appears possible at this time for a facility with a medium-scale and medium-strength stellarator whose required facility electric power is less than 40 MW to produce 2.5 × 10 17 neutrons s−1 for the production of radioisotopes. With the addition of a tritium breeding blanket, such a facility could produce 0.2 grams d−1 or 70 grams yr−1 of tritium.

Impact of trapping on tritium self-sufficiency and tritium inventories in fusion power plant fuel cycles

Abstract The dynamic analysis of fusion power plant (FPP) fuel cycles highlights the challenge of achieving tritium self-sufficiency in future FPPs. While state-of-the-art fuel cycle models offer valuable insights into the necessary design parameters for attaining tritium self-sufficiency, none of these models currently consider the impact of tritium trapping within fuel cycle components. However, detailed analysis of individual components reveals that substantial amounts of tritium can be trapped within the first wall, divertors, and breeding blanket systems, suggesting that tritium trapping may significantly influence the FPP ability to achieve self-sufficiency. The compounded effects of additional tritium traps generated by irradiation effects and component replacements further exacerbate this challenge. The novelty of this work is the integration of an explicit, physics-based model for tritium trapping, evolution of damage-induced traps, and component replacements into a dynamic, system-level model of a fuel cycle. The results show an increase of a factor 103 - 104 of tritium inventory in the first wall and vacuum vessel of an ARC-class FPP when accounting for the aforementioned phenomena. This, coupled with the replacement of components subject to significant tritium trapping, slows down fuel cycle dynamics, resulting in an extended tritium doubling time (50% increase), higher start-up inventory (30% increase), and higher required tritium breeding ratio (2-5%) compared to a scenario without tritium trapping.

Advancing tritium self-sufficiency in fusion power plants: insights from the BABY experiment

Abstract In the pursuit of fusion power, achieving tritium self-sufficiency stands as a pivotal challenge.
Tritium breeding within molten salts is a critical aspect of next-generation fusion reactors, yet experimental measurements of \gls{tbr} have remained elusive.
Here we present the results of the \gls{baby} experiment, which represents a pioneering effort in tritium research by utilizing high-energy (\SI{14}{\mega\electronvolt}) neutron irradiation of molten salts, a departure from conventional low-energy neutron approaches. 
Using a small-scale (\SI{100}{\milli\litre}) molten salt tritium breeding setup, we not only simulated, but also directly measured a \gls{tbr} \textcolor{red}{(\num{3.57e-4})}. 
This innovative approach provides crucial experimental validation, offering insights unattainable through simulation alone.
Moreover, our findings reveal a surprising outcome: tritium was predominantly collected as HT, contrary to the expected TF. 
This underscores the complexity of tritium behavior in molten salts, highlighting the need for further investigation.
This work lays the foundation for a more sophisticated experimental setup, including increasing the volume of the breeder, enhancing neutron detection, and refining tritium collection systems.
Such improvements are crucial for advancing our understanding of fusion reactor feasibility and paving the way for future experiments.

Impact of MHD mixed convection on heat and mass transfer in a typical DCLL breeder blanket

Abstract A typical dual-coolant lead lithium (DCLL) liquid blanket employs eutectic liquid lithium lead alloy (PbLi) as coolant and tritium breeder for Tokamak nuclear fusion reactor, and is responsible for transporting heat and tritium. The blanket is subjected to both strong magnetic field and strong radially non-uniform neutron volume heating. The flow of liquid metal is influenced by the combined effects of magnetohydrodynamics (MHD), buoyancy, and inertial effects, resulting in a complex MHD mixed convection phenomenon. This study employs direct numerical simulation (DNS) based on finite volume method to analyze the flow, heat transfer, and mass transfer within blanket unit. The results demonstrate that mixed convection has a significant impact on the MHD pressure drop and flow characteristics. The complex MHD mixed convection gives rise to a considerable number of vortices, particularly in regions of the turn and the downward poloidal duct. The formation of low-speed zones and jets within the blanket has the potential to significantly impact heat and mass transfer, leading to the accumulation of heat and tritium. On the other hand, the markedly elevated tritium production rate is a significant contributing factor to the obviously elevated tritium concentration in the vicinity of the first wall.

Influence of full-cycle heat treatment on microstructure and mechanical properties of CLF-1 steel

Based on the manufacturing requirements of tritium breeding blankets for fusion reactors, it is valuable to study the microstructure evolution and mechanical properties of CLF-1 steel, a candidate structural material, during the full-cycle heat treatment from profile to component. The full-cycle heat treatment includes isothermal annealing and multiple quenching and tempering treatments. Results indicate that multiple quenching and tempering treatments refine the size of prior austenite grains and reduce dislocation density and the width of lath martensite. With the increase in the number of quenching and tempering treatments, the CLF-1 steels exhibit a slight decrease in tensile strength while the impact toughness noticeably improves. Research has found that the effect of isothermal annealing treatment includes an increase in the size of the prior austenite grains and precipitates, as well as a change in the matrix from martensite to ferrite. As a result, the tensile strength and impact energy decrease sharply, while the ductility improves after isothermal annealing treatment. Subsequent quenching and tempering treatment can recover the microstructure and mechanical properties.

Physics and technology maturity level required for the K-DEMO design points

The conceptual design study for the Korean fusion demonstration reactor (K-DEMO) has been conducted, placing an emphasis on the imperative to reduce the reactor's dimension as a means to cost minimization. The design space of K-DEMO was delineated based on maturity level of physics and technology. Advanced technology features such as a use of a tungsten carbide (WC) shield, a tritium breeding blanket concept of He cooled lithium lead (HCLL), the maximum allowable magnetic field at the TF coil, Bmax = 16 T with a Nb3Sn superconducting material, etc. were adopted. With specified design criteria, including a net electric power ≥ 300 MW, a fusion gain, Q > 20.0, a neutron wall loading < 2.0 MW/m2, an indicator of divertor power handling capability (ratio of power to divertor to major radius), Pdiv/R0 < 25 MW/m, and the capability for steady-state operation, a design space of the K-DEMO was established based on the energy confinement scaling law of IPB98[y,2] under physics level of Greenwald density fraction, ne/nG < 1.2, normalized plasma beta (ratio of plasma pressure to magnetic pressure normalized by plasma current divided by the product of minor radius and toroidal magnetic field), βN < 3.0, confinement enhancement factor, H < 1.3, and a direct cost ≤ 7.5 B$. After an exploration of system parameters, prospective design points for K-DEMO were identified, characterized by a major radius, R0 ∼ 6.8 m, an aspect ratio, A = 3.1, a toroidal magnetic field at plasma center, BT ≥ 6.5 T, and a fusion power, Pfusion ∼ 1,500 MW. When β-independent energy confinement scaling law was applied, the design points were accessible with smaller ne/nG, βN, H, Pfusion, and larger BT. From a sensitivity analysis of the minimum major radius to the input parameters, strong sensitivities to Greenwald density fraction, ne/nG, normalized plasma beta, βN, confinement enhancement factor, H, edge safety factor, qedge, and elongation, κ were found. Additionally, the operational envelope in physics and technology parameters was established with system parameters associated with the design points.

Novel high temperature tritium blanket designs for confined spaces in spherical tokamak fusion reactors

Tritium self-sufficiency is one of the fundamental challenges for commercially viable deuterium–tritium nuclear fusion power stations. The combination of key high temperature radiation shielding materials that possess dense, high neutron absorption cross-section, and moderation properties, and tritium breeding materials could involve interesting design spaces for the central column challenge in spherical tokamaks. Potential tungsten alloys can be used for two functions: radiation shielding and structural material, providing a new design space window for spherical tokamak central column breeding space. In this paper, we present two novel high temperature concepts for the inboard side of the breeder blanket in a confined space, such as a spherical tokamak. A tungsten–rhenium–hafnium-carbide lithium-based design was found to offer the best TBR given a parameter optimisation based on shielding and thermal requirements. A silicon-carbide lead-lithium breeder design was also investigated. The highest TBR was found to be 0.135 in a 3D neutronics calculation with a W-24.5Re-2HfC (structural and shielding, wt%), Li (90% Li-6 enriched breeder), and tungsten pentaboride (W2B5) (shielding) option. Although this TBR is lower than unity, it will contribute to the reactor’s global TBR.

Conceptual design study of the tritium breeding unit for verifying the long-term performance of the breeding blanket

Fusion reactors produce high-energy neutrons through the Deuterium-Tritium (DT) reaction. The long-term performance verification and structural integrity assessment of the breeding blanket under continuous plasma operation are essential for the development of Demonstration fusion power reactor (DEMO) breeding blankets. The Korea Institute of Fusion Energy (KFE) has been planning the construction of a test facility called the Integrated Breeding Test Facility (IBTF) that uses a linear accelerator as the neutron source. A solid beryllium target has been considered to generate fusion-like neutrons. A conceptual design of the Tritium Breeding Unit (TBU), which is a key component of the breeding blanket, has been carried out. This study evaluated tritium production under fusion-like conditions using factors such as the position and thickness of the graphite block, the thickness of the first wall, and the radial length of the breeding zone. The results showed that increasing graphite block thickness and optimizing the dimensions of the TBU enhanced tritium production. However, these changes affected the structural integrity of the TBU. The side wall of the TBU experienced membrane plus bending stress, which exceeded the Level C and D acceptance criteria under the assumption of In-Box LOCA. Measures were investigated to ensure structural integrity while improving tritium production of the TBU.

A novel preparation of porous Li2TiO3 pebbles with a distinctive structure

Optimization of the pore structure stands out as an effective approach for bolstering the tritium release performance of lithium-based tritium breeding ceramics, particularly when employing a high open porosity configuration. In this study, Li2TiO3 fibers were synthesized via the hydrothermal method, with the assistance of TiO2-B nanowires acting as templates. Subsequently, porous Li2TiO3 pebbles with varying microstructures were fabricated through a combination of the wet method and a sintering process. Sintering process plays a crucial role in determining the microscopic morphology. An increase in temperature resulted in a suppression of the growth of fibrous grains and the transformation of fibrous grains into quasi-spherical grains. Moreover, a two-step sintering process was proposed as a means of improving the crush load (15.9 N vs. 22.0 N) of Li2TiO3 pebbles while simultaneously preserving their high porosity (24.91%). The Li2TiO3 pebbles, produced by the two-step sintering process, exhibited a distinctive micromorphology characterized by interconnected particles forming short rod-shaped microstructures. This distinctive micromorphology contributes to the maintenance of a high degree of open porosity (10.52%).

Optimization of Tokamak Blanket Design to Minimize Neutron Flux on Magnets

Fusion energy, a promising solution to global energy challenges, replicates the same processes that give our sun its energy, notably through deuterium-tritium (D-T) fusion reactions that produce significant energy. This study explores the mechanics and challenges of various fusion reactions, emphasizing the pivotal role of lithium breeder blankets in producing tritium, essential for sustaining D-T fusion in tokamak reactors. The helium-cooled lithium lead (LiPb) blanket design was simulated to optimize tritium breeding and neutron flux management. Results indicated a tritium breeding ratio (TBR) of 1.15, surpassing the self-sufficiency target of 1.1, with further improvements through increased lithium content and blanket thickness. Effective neutron shielding ensured safe operational limits for reactor components. These findings demonstrate the feasibility of achieving self-sustaining fusion reactions, essential for the viability of fusion power as a sustainable energy source. Future research will focus on advanced materials, refined simulations, and enhanced cooling technologies to further optimize fusion reactor designs.

Managing the heat: In-Vessel Components.

The Spherical Tokamak for Energy Production (STEP) programme aims to deliver a first-of-a-kind fusion prototype powerplant (SPP). The SPP plasma places extreme heat, particle and structural loads onto the plasma-facing components (PFCs) of the divertor, limiters and inboard and outboard sections of the first wall. The PFCs must manage the heat and particle loads and wider powerplant requirements relating to safety, net power generation, tritium breeding and plant availability. To enable STEP PFC concepts to be identified that satisfy these wide-ranging requirements, an iterative design ('Decide & Iterate') methodology has been used to synchronize a prioritized set of decisions, within the fast-paced, iterative, whole plant concept design schedule. This paper details the 'Decide and Iterate' methodology and explains how it has enabled the identification of the SPP PFC concepts. These include innovative PFC solutions such as a helium-cooled discrete and panel limiter design to increase tritium breeding while providing sufficient coverage and enabling individual limiter replacement; the integration of the outboard first wall with the breeding zone to enhance fuel self-sufficiency and power generation; and the use of heavy water (D2O) within the inboard first wall and divertor PFCs to increase tritium breeding within the outboard breeding zone. This article is part of the theme issue 'Delivering Fusion Energy - The Spherical Tokamak for Energy Production (STEP)'.

MANTA: a negative-triangularity NASEM-compliant fusion pilot plant

Abstract The MANTA (Modular Adjustable Negative Triangularity ARC-class) design study investigated how negative-triangularity (NT) may be leveraged in a compact, fusion pilot plant (FPP) to take a ‘power-handling first’ approach. The result is a pulsed, radiative, ELM-free tokamak that satisfies and exceeds the FPP requirements described in the 2021 National Academies of Sciences, Engineering, and Medicine (NASEM) report ‘Bringing Fusion to the U.S. Grid’ (2021 Bringing Fusion to the U.S. Grid). A self-consistent integrated modeling workflow predicts a fusion power of 450 MW and a plasma gain of 11.5 with only 23.5 MW of power to the scrape-off layer (SOL). This low P SOL together with impurity seeding and high density at the separatrix results in a peak heat flux of just 2.8 MW m−2. MANTA’s high aspect ratio provides space for a large central solenoid (CS), resulting in ∼15 minute inductive pulses. In spite of the high B fields on the CS and the other REBCO-based magnets, the electromagnetic stresses remain below structural and critical current density limits. Iterative optimization of neutron shielding and tritium breeding blanket yield tritium self-sufficiency with a breeding ratio of 1.15, a blanket power multiplication factor of 1.11, toroidal field coil lifetimes of 3100 ± 400 MW · yr, and poloidal field coil lifetimes of at least 890 ± 40 MW · yr. Following balance of plant modeling, MANTA is projected to generate 90 MW of net electricity at an electricity gain factor of ∼ 2.4 . Systems-level economic analysis estimates an overnight cost of US$3.4 billion, meeting the NASEM FPP requirement that this first-of-a-kind be less than US$5 billion. The toroidal field coil cost and replacement time are the most critical upfront and lifetime cost drivers, respectively.

Blistering Behavior of Beryllium and Beryllium Alloy under High-Dose Helium Ion Irradiation.

Beryllium (Be) has been selected as the solid neutron multiplier material for a tritium breeding blanket module in ITER, which is also the primary option of the Chinese TBM program. But the irradiation swelling of beryllium is severe under high temperature, high irradiation damage and high doses of transmutation-induced helium. Advanced neutron multipliers with high stability at high temperature are desired for the demonstration power plant (DEMO) reactors and the China Fusion Engineering Test Reactor (CFETR). Beryllium alloys mainly composed of Be12M (M is W or Ti) phase were fabricated by HIP, which has a high melting point and high beryllium content. Beryllium and beryllide (Be12Ti and Be12W) samples were irradiated by helium ion with 30 keV and 1 × 1018 cm-2 at RT. The microstructures of Be, Be12Ti and Be12W samples were analyzed by SEM and TEM before and after ion irradiation. The average size of the first blistering on the surface of Be-W alloy is about 0.8 μm, and that of secondary blistering is about 79 nm. The surface blistering rates of the beryllium and beryllide samples were also compared. These results may provide a preliminary experimental basis for evaluating the irradiation swelling resistance of beryllium alloy.

Pre-conceptual design and proof of principle assessments of self-cooled Toroidally symmetric lead-lithium (TSLL) blanket

A new self-cooled liquid metal blanket concept called TSLL (Toroidally Symmetric Lead-Lithium) blanket is proposed and assessed, including analysis for magnetohydrodynamic (MHD) flows, structural analysis, and heat transfer and neutronics assessments using the ARC reactor with demountable magnets designed by the Commonwealth Fusion Systems (CFS) as a testbed. The proposed blanket utilizes lead-lithium (PbLi) alloy as breeder/coolant and reduced activation ferritic/martensitic (RAFM) steel as structural material. A special feature of the new concept is the toroidally symmetric flow in the blanket integrated first wall and the breeding zone to reduce the MHD pressure drop, while using anchor links to strengthen the first wall construction. Provided analysis suggests acceptable MHD pressure drop, required mechanical integrity and high tritium breeding ratio of ∼ 1.64. As a result of these assessments, the new blanket concept can be recommended for more detailed studies as a promising blanket candidate for implementation in future fusion devices.

Comparative Study on the Reduction of Tritium Breeding Ratio Caused by Inventory Changes of a Solid-State Tritium Breeding Blanket in a Fusion Demonstration Reactor Using MCNP and FISPACT-II

The self-sufficient supply of tritium in the deuterium-tritium nuclear fusion demonstration reactor (DEMO) is a fundamental design requirement. But, it is hindered by depletion of tritium breeding materials resulting in reduction of tritium breeding ratio (TBR) less than the initial design value especially in the solid-state tritium breeding blanket (TBB) of the DEMO. Unlike the liquid tritium breeding blanket of DEMO, compensation measures of the depleted breeding material in the solid-state TBB will be its substitution depending on the reduction rate of TBR. To estimate the replacement period of the solid-state TBB, it is required to estimate the reduction rate of TBR according to the operation conditions of the DEMO and the physical configuration of a solid-state TBB. In this study, the representative simulation codes, MCNP and FISPACT-II, are used for assessment of the reduction rate of TBR with the benchmarking model which is modified from the one poloidal segment of the TBB in the Korean-DEMO. After 3 full power-year operations with the neutron irradiation with the benchmarking model, the TBR simulated by MCNP is reduced to 96.84% of the initially calculated TBR, but the TBR calculated by FISPACT-II is reduced to 90.57% of the initially calculated TBR.

Assessment of hydrogen production potential of APEX fusion blanket via cobalt-chlorine and copper-chlorine cycles

In this paper, the neutronic and hydrogen production performances of the advanced power extraction fusion reactor (APEX) have been investigated by using TRISO thorium fuel. With this study, TRISO-coated nuclear fuel was first used in the APEX fusion reactor. In this calculation, firstly, the neutronic performances for 10 % ThC + 90 % FLiBe and various 6Li enrichment ratios have been computed with MCNP. Secondly, the potential of hydrogen production of the hydrogen unit, which contains cobalt-chlorine cycle and copper-chlorine cycle, integrated with the advanced power extraction fusion reactor has been performed. The tritium breeding ratio has changed between 1.09 and 1.15 among 7.5 % and 90 % 6Li enrichment ratios for 10 % ThC + 90 % FliBe. The energy multiplication factors have been obtained as 1.2727 and 1.3209, respectively, among 7.5 % and 90 % 6Li enrichment ratios. The thermal power, its ratio, and the hydrogen production amount have been calculated depends on the energy multiplication factor for various 6Li enrichment ratios and each considered cycle. The results showed that hydrogen production quantity increases with a rising 6Li enrichment ratio for both cycles. The highest produced hydrogen amount was obtained as 12.74 kg/s via the Cu-Cl cycle, whereas hydrogen production with the Co-Cl cycle has been computed as 6.79 kg/s in the cases of 10 % ThC fuel ratio and 90 % 6Li enrichment ratio.

DEM study on the force chain evolution of biaxial compression of pebble bed

Li4SiO4 and Li2TiO3 are granular materials in the pebble bed of fusion reactors, and their physical parameters and mechanical properties directly affect the working status and structural design of the pebble bed. The force chain can be employed to intuitively describe the mechanical properties of the pebble bed in the particle aggregate. Based on the discrete element method, we conducted a numerical study on the evolution characteristics of the force chain of Li4SiO4 and Li2TiO3 pebble beds under biaxial compression and explored the friction coefficient and pebble bed effect of the aspect ratio on the force chain distribution. The results revealed that as the particle friction coefficient increased, the force chain tended to be isotropically distributed, indicating that the friction mechanism was more conducive to a uniform distribution of the force chain. During compression, the average coordination number of the pebble bed increased with the friction coefficient. The Li2TiO3 particles had larger gravity than the Li4SiO4 pebble bed therefore, the Li2TiO3 pebble bed accounted for more force chains in the vertical direction. When the aspect ratio of the pebble bed was less than 0.5 or greater than 2.5, the distribution of force chains exhibited strong anisotropy. Conversely, when the pebble bed aspect ratio ranged between 0.5 and 2.5, the distribution of force chains tended towards an isotropic trend. Moreover, the number of force chains in each direction varied with changes in the aspect ratio of the pebble bed, characterized by a high concentration at the edges and a lower concentration in the middle. The results can provide an in-depth understanding of the force chain distribution and evolution characteristics in the pebble bed and provide a theoretical basis for designing and analyzing tritium breeding pebble beds.

A comprehensive study on the phase composition, mechanical properties and stability of Li4SiO4-Li2ZrO3 biphasic ceramics

Lithium orthosilicate (Li4SiO4) is regarded as a candidate for tritium breeding in fusion reactors. In this study, the Li4SiO4-Li2ZrO3 biphasic was developed to improve the sinterability, density, and mechanical properties of Li4SiO4. The Li4SiO4-xLi2ZrO3 (x = 0.5, 1) composite powders were prepared using solid-state reaction via in-situ method. Li6Zr2O7 existed in the ceramic powders at low calcination temperatures. When the calcination temperature was increased to 900 °C, Li6Zr2O7 was transformed into Li2ZrO3 due to the decomposition of Li6Zr2O7 at high temperatures. The TEM observation confirmed that the powders consisted of Li4SiO4 and Li2ZrO3. The Li4SiO4 and Li4SiO4-xLi2ZrO3 (x = 0.5, 1) pebbles were fabricated by the sol-gel method. The measurement results showed that the pebbles had a narrow size distribution and fine sphericity. The density of Li4SiO4-Li2ZrO3 pebbles reached 96.01% of the theoretical density when it was sintered at 1000 °C for 4 h. Compared with the Li4SiO4, the grain size of ceramic pebbles was significantly reduced. Owing to decreased grain size, 139.0 N crush load for the pebbles and 92.4 MPa bending strength for the sintered bodies were achieved. Besides, the ceramics with the Li4SiO4 to Li2ZrO3 ratio of 2: 1 exhibited preferable mechanical properties. Further, to investigate the chemical stability of biphasic ceramics, the structure and mechanical properties were examined under high temperatures and continuous inert gas purging, simulating the working condition of fusion reactors. It was shown that there was almost no change in phase composition and grain size after purging for 60 h at 650 °C. For the mechanical properties, the crush load was decreased initially due to the cracking in the surface region of the ceramic and then increased because the cracking was recovered.

Chemical characterization of lithium based ceramics utilizing charged particle activation and ion beam techniques

Non-destructive methodologies using activation analysis and ion beam analysis techniques were optimized for the chemical characterization of ceramic materials, lithium titanate and lithium niobate, which have application in tritium breeding blanket. The analyses were carried out as a part of chemical quality control exercise. The atomic ratios of Li/Ti, Li/Nb were quantified by charged particle activation analysis using 13 MeV proton beam from variable energy cyclotron facility and particle induced gamma ray emission/Rutherford backscattering spectrometry using 3 MeV/2 MeV proton beam from 3MV tandem accelerator facility. The results of these different analytical methods are in good agreement, which established the applicability of these activation analysis and ion beam techniques for the chemical characterization of these ceramic materials.

Comparison between tritium breeders with pebble bed and Gyroid block configurations using numerical investigation: Flow behavior and heat transfer performance

This study proposed using a Gyroid structure, a type of triply periodic minimal surface (TPMS), to replace conventional pebble bed configurations and address drawbacks such as inherently low packing fractions, extensive stress concentrations, high flow resistance, and low thermal conductivity. The feasibility of the Gyroid block configuration was evaluated by comparing it with a pebble bed configuration regarding the flow behaviors and heat transfer performance. The flow behavior of the porous block configuration was superior to that of the pebble bed configuration. At the same inlet velocity of 0.1–0.2 m/s, the porous block configuration had a ∼56 % higher average flow velocity and only half the pressure drop. The convective heat transfer coefficient and Nusselt numbers in the porous block configuration were 56–84 % and 2.6–3.1 times higher, respectively, than those in the pebble bed configuration. The effective thermal conductivity of the porous block configuration increased by ∼18 % at an inlet velocity of 0.15 m/s and a temperature of 773–1173 K. This study preliminarily verified the feasibility of the Gyroid structures for tritium breeders in engineering applications and is anticipated to provide new opportunities for the design of tritium breeders with higher lithium densities, remarkable flow behavior, and outstanding thermal performance.