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Efficient tritium extraction from liquid lead-lithium (PbLi) breeding blankets is critical for fusion reactor technology development. In this study, numerical simulation methods based on a fundamental gas-liquid contactor (GLC) model were employed to optimize boundary conditions and external structures, ensuring stability of liquid PbLi level during operation. Computational results show that the improved GLC achieves a tritium extraction efficiency of 19.18%. Further enhancement was realized by increasing GLC mesh density and installing a gas-guiding device at the internal helium inlet, significantly improving gas-liquid contact within the GLC. Final results indicate that GLC’s tritium extraction efficiency increased to 25.73%, suggesting that the gas guiding device introduction effectively enhanced gas-liquid contact and tritium extraction efficiency.

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ∼400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α ′), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

Fusion reactor materials for the first wall and blanket must have high strength, be radiation tolerant and be reduced activation (low post-use radioactivity), which has resulted in reduced activation ferritic/martensitic (RAFM) steels. The current steels suffer irradiation-induced hardening and embrittlement and are not adequate for planned commercial fusion reactors. Producing high strength, ductility and toughness is difficult, because inhibiting deformation to produce strength also reduces the amount of work hardening available, and thereby ductility. Here we solve this dichotomy to introduce a high strength and high ductility RAFM steel, produced by a modified thermomechanical process route. A unique multiscale microstructure is developed, comprising nanoscale and microscale ferrite, tempered martensite containing fine subgrains and a high density of nanoscale precipitates. High strength is attributed to the fine grain and subgrain and a higher proportion of metal carbides, while the high ductility results from a high mobile dislocation density in the ferrite, subgrain formation in the tempered martensite, and the bimodal microstructure, which improves ductility without impairing strength. Reduced-activation ferritic-martensitic (RAFM) steels are promising fusion reactor materials having high radiation tolerance and low post-use radioactivity, but achieving high strength, ductility and toughness is difficult. The authors demonstrate a modified thermomechanical process to produce RAFM steel with a multiscale microstructure which endows both high strength and high ductility.

High-performance structural materials (HPSMs) are needed for the successful and safe design of fission and fusion reactors. Their operation is associated with unprecedented fluxes of high-energy neutrons and thermomechanical loadings. In fission reactors, HPSMs are used, e.g., for fuel claddings, core internal structural components and reactor pressure vessels. Even stronger requirements are expected for fourth-generation supercritical water fission reactors, with a particular focus on the HPSM’s corrosion resistance. The first wall and blanket structural materials in fusion reactors are subjected not only to high energy neutron irradiation, but also to strong mechanical, heat and electromagnetic loadings. This paper presents a historical and state-of-the-art summary focused on the properties and application potential of irradiation-resistant alloys predominantly strengthened by an oxide dispersion. These alloys are categorized according to their matrix as ferritic, ferritic–martensitic and austenitic. Low void swelling, high-temperature He embrittlement, thermal and irradiation hardening and creep are typical phenomena most usually studied in ferritic and ferritic martensitic oxide dispersion strengthened (ODS) alloys. In contrast, austenitic ODS alloys exhibit an increased corrosion and oxidation resistance and a higher creep resistance at elevated temperatures. This is why the advantages and drawbacks of each matrix-type ODS are discussed in this paper.

The transition to turbulence in conduits is among the longest-standing problems in fluid mechanics. Challenges in producing or saving energy hinge on understanding promotion or suppression of turbulence. While a global picture based on an intrinsically 3-D subcritical mechanism is emerging for 3-D turbulence, subcritical turbulence is yet to even be observed when flows approach two dimensions, e.g. under intense rotation or magnetic fields. Here, stability analysis and direct numerical simulations demonstrate a subcritical quasi-two-dimensional (quasi-2-D) transition from laminar flow to turbulence, via a radically different 2-D mechanism to the 3-D case, driven by nonlinear Tollmien–Schlichting waves. This alternative scenario calls for a new line of thought on the transition to turbulence and should inspire new strategies to control transition in rotating devices and nuclear fusion reactor blankets.

Fusion energy is a promising, safe, and reliable green energy solution to the increasing energy demand. However, there are several materials challenges that need to be overcome to increase the technical readiness to a level that enables a fusion pilot plant on the grid. This focus issue aims to identify and address a set of such key impediments for realizing deuterium-tritium (D–T) fusion power in a tokamak reactor and highlight the most recent progress on those research frontiers. The main emphasis of this collection is on materials development challenges resulting from helium irradiation, neutron-induced degradation, thermomechanical loading, and the corrosive environment faced by the divertor and first-wall materials, commonly known as plasma-facing components, and blanket systems for tokamak fusion reactors.

The exploitation of fusion energy in tokamak reactors relies on efficient and reliable tritium management. The tritium needed to sustain the deuterium–tritium fusion reaction is produced in the Li-based blanket surrounding the plasma chamber, and, therefore, the effective extraction and purification of the tritium bred in the Li-blankets is needed to guarantee the tritium self-sufficiency of future fusion plants. This work introduces a new technology for the extraction of tritium from the Pb–Li eutectic alloy used in liquid blankets. Process units based on the concept of Membrane Gas–Liquid Contactor (MGLC) have been studied for the extraction of tritium from the Pb–Li in the Water Cooled Lithium Lead blankets of the DEMO reactor. MGLC units have been preliminarily designed and then compared in terms of the permeation areas and sizes with the tritium extraction technologies presently under study, namely the Permeator Against Vacuum (PAV) and the Gas–Liquid Contactors (GLCs). The results of this study show that the DEMO WCLL tritium extraction systems using MGLC require smaller permeation areas and quicker permeation kinetics than those based on PAV (Permeator Against Vacuum) devices. Accordingly, the MGLC extraction unit exhibits volumes smaller than those of both PAV and GLC.

High flux components in fusion reactor are designed to transfer high thermal loads from the core plasma under steady-state operation and emergency conditions. In order to improve the heat transfer ability, hypervapotron structure is proposed to promote the heat transfer performance by optimizing the thermal hydraulic design. In this paper, three-dimensional numerical simulation was carried out using CFD methods for the improved solution of the first wall. The results show that the heat transfer performance of the triangular fin structure with tilt angle of 30° is superior to 0 °, 15°and 45°. At a typical flow rate of 2 m/s, the heat transfer efficiency of the triangular fin with tilt angle of 30° improved by a maximum of 118.1% compared with that of the smooth flow channel. In addition, the heat transfer performance of the triangular fin structure is better than other fin structure, and its maximum temperature drops by 136.3K compared to the smooth flow channel. The design solution may provide a viable new method for efficient cooling of future fusion reactor components with high heat flux.

The operation of fusion reactors is based on the reaction that occurs when two heavy hydrogen isotopes, deuterium and tritium, combine to form helium and a neutron with an energy of 14.1 MeV D + T → He + n. For this reaction to occur, it is necessary to produce tritium in the facility itself, as tritium is not common in nature. The generation of tritium in the facility is a key function of the breeder blanket. During the operation of a D–T fusion reactor, high-energy tritium is generated as a result of the 6Li(n,α)T reaction in a lithium-containing ceramic material in the breeder blanket. Lithium metatitanate Li2TiO3 is proposed as one of the promising materials for use in the solid breeder blanket of the DEMO reactor. Several concepts for test blanket modules based on lithium ceramics are being developed for testing at the ITER reactor. Lithium metatitanate Li2TiO3 has good tritium release parameters, as well as good thermal and thermomechanical characteristics. The most important property of lithium ceramics Li2TiO3 is its ability to withstand exposure to long-term high-energy radiation at high temperatures and across large temperature gradients. Its inherent thermal stability and chemical inertness are significant advantages in terms of safety concerns. This study was a continuation of research regarding tritium and helium release from lithium metatitanate Li2TiO3 with 96% 6Li during irradiation at the WWR-K research reactor using the vacuum extraction method. As a result of the analysis of experiments regarding the irradiation of lithium metatitanate in vacuum conditions, it has been established that, during irradiation, peak releases of helium from closed pores of the ceramics are observed, which open during the first 7 days of irradiation. The authors assumed that the reasons samples crack are temperature gradients over the ceramic sample, resulting from the internal heating of pebbles under the conditions of their vacuum evacuation, and contact with the bottom of the evacuated capsule. The temperature dependence of the effective diffusion coefficient of tritium in ceramics at the end of irradiation and the parameters of helium effusion were also determined.