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MHD R&D Activities for Liquid Metal Blankets
According to the most recently revised European design strategy for DEMO breeding blankets, mature concepts have been identified that require a reduced technological extrapolation towards DEMO and will be tested in ITER. In order to optimize and finalize the design of test blanket modules, a number of issues have to be better understood that are related to the magnetohydrodynamic (MHD) interactions of the liquid breeder with the strong magnetic field that confines the fusion plasma. The aim of the present paper is to describe the state of the art of the study of MHD effects coupled with other physical phenomena, such as tritium transport, corrosion and heat transfer. Both numerical and experimental approaches are discussed, as well as future requirements to achieve a reliable prediction of these processes in liquid metal blankets.
Journal Article
That's my blanket, baby!
Bella has had her special blanket, named Blanket, since she was a baby. She and Blanket did everything together-painting, singing, and playing in mud puddles. Now Bella has a new baby brother who has his own brand-new blanket. But he prefers Bella's old, muddy, smelly blanket! What's a big sister to do?
Book
Development of water-cooled cylindrical blanket in JA DEMO
The concept of the tritium breeding blanket for Japan’s DEMOnstration fusion reactor (JA DEMO) has been developed with pressure tightness against in-box loss-of-coolant accidents based on a water-cooled solid breeder concept. The cooling conditions are designed on the pressurized-water reactor water conditions which are the coolant temperature of 290 °C–325 °C and the operating pressure of 15.5 MPa, respectively. The point of the blanket design is to reduce the amount of structural material in casing as well as to ensure its pressure tightness. This is because a decrease in the amount of structural material improves tritium breeding ratio (TBR). A cylindrical structure, a thin wall casing structure which ensures pressure tightness and could increase TBR, is feasible. However, a relatively larger useless space is expected between modules when the cylindrical blanket modules are arranged in a vacuum vessel, which could decrease TBR. Therefore, the cylindrical blanket modules are to be in a close-packed arrangement to reduce useless space. The Be12Ti block (which shows a minor swelling compared to Be) is selected to achieve the target TBR as the net Be density is equivalent to the case of the Be pebble. The use of Be12Ti blocks can reduce or remove the cooling piping inside the module as the Be12Ti block has a higher thermal conductivity than pebbles. As a result of the neutronics, finite element method, and computational fluid dynamics analyses, it was found that the target TBR value can be achieved in cylindrical structure blanket that ensure pressure tightness.
Journal Article
The poky little puppy and the patchwork blanket
The poky little puppy loves his blanket so much that he takes it outside, even though he's not supposed to. But too much rough play soon leaves his blanket in shreds. What will he do now?
Book
Neutronics experiment of tritium breeding in supercritical CO2 cooled lithium-lead blanket mock-up by D–T neutron irradiation
The tritium breeding capacity of fusion blankets is a critical factor in achieving tritium self-sufficiency in fusion reactors. Current designs for tritium production in blankets are based on neutronics simulations, whose accuracy requires experimental validation. This study focuses on the experimental validation of the neutronic design reliability of the supercritical CO2-cooled lithium-lead (COOL) blanket developed for the China fusion engineering test reactor (CFETR). For this purpose, an experimental mock-up was designed and fabricated to replicate the key neutronics characteristics of COOL blanket. The mock-up was irradiated using D–T neutron generator and validated using multiple techniques: tritium production rate (TPR) online monitoring with a miniature back-to-back lithium glass scintillator detector, TPR integral measurement using Li2CO3 pellets analyzed by liquid scintillation counting (LSC), and neutron flux measurement with activation foils (Au, Zr). To enhance accuracy, corrections were applied for lithium-lead (PbLi) segregation based on element analysis and for neutron source intensity based on in-situ depth profiling of tritium in the tritide target. The results show excellent agreement between experiments and simulations, with calculation-to-experimental (C/E) value ranging from 0.96 to 1.11 for TPR measured by lithium-glass scintillator detectors, 0.90–1.09 for TPR by Li2CO3 pellets, and 0.84–1.13 for reaction rates measured by activation foils.
Journal Article
A code-to-code benchmark for magneto-convection in a horizontal duct
Liquid metals and magnetic fields are used in many technical applications such as metallurgy, crystal growth and nuclear fusion reactors. When an electrically conducting fluid moves in a magnetic environment, electric currents and electromagnetic forces are generated that affect velocity and pressure losses in the flow. These magnetohydrodynamic (MHD) interactions have to be investigated to optimize the engineering processes. The characteristics of MHD flows depend on the geometrical configuration, the strength of the applied magnetic field, the electrical properties of fluid and structural materials and the thermal conditions. In the so-called blankets for fusion reactors, where liquid metals are used to breed the plasma fuel component tritium and to extract the generated heat, magneto-convective flows play a crucial role in determining heat and mass transfer. Therefore, the availability of numerical codes to simulate this type of flow is mandatory and their validation is a necessary step to guarantee the reliability of the results. For that reason, a benchmark problem has been defined to simulate liquid metal flows in a horizontal rectangular duct heated from below and exposed to a non-uniform magnetic field. Results obtained by five research groups using different codes are compared.
Journal Article
The Irradiation Effects in Ferritic, Ferritic–Martensitic and Austenitic Oxide Dispersion Strengthened Alloys: A Review
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.
Journal Article