Abstract: The ITER Divertor is aimed at exhausting the major part of the plasma thermal power including alpha power and at minimizing the helium and impurity content in the plasma. It consists of assembly of 54 divertor cassettes. Each divertor cassette has three plasma-facing components (PFCs) and one cassette body that accommodates these PFCs. The PFCs consist of the Inner and Outer Vertical targets and the Dome. The Vertical Targets directly intercept the magnetic field lines and are designed to withstand heat fluxes as high as 20 MW/m2. After the successful completion of the qualification phase, three Parties (Japan, Europe and Russia) are qualified and they are in charge of procurement of the first Divertor set. In the paper, the detail component design and the manufacturing as well as the integration of the system are reported.
Abstract: Refractory materials, in particular tungsten base materials are considered as primary candidates for structural high heat load applications in future nuclear fusion power plants. Promising helium-cooled divertor design outlines make use of their high heat conductivity and strength. The upper operating temperature limit is mainly defined by the onset of recrystallization but also by loss of creep strength. The lower operating temperature range is restricted by the use of steel parts for the in- and outlets as well as for the back-bone. Therefore, the most critical issue of tungsten materials in connection with structural divertor applications is the ductile-to-brittle transition. Another problem consists in the fact that especially refractory alloys show a strong correlation between microstructure and their manufacturing history. Since physical and mechanical properties are influenced by the underlying microstructure, refractory alloys can behave quite different, even if their chemical composition is the same. Therefore, creep and thermal conductivity have been investigated using typical commercial tungsten materials. Moreover, the fracture behavior of different tungsten based semi-finished products was characterized by standard Charpy tests which have been performed up to 1100 °C in vacuum. Due to their fabrication history (powder mixing, pressing, sintering, rolling, forging, or swaging) these materials have specific microstructures which lead different fracture modes. The influence of the microstructure characteristics like grain size, anisotropy, texture, or chemical composition has been studied.
Abstract: V-4Cr-4Ti alloy is an attractive candidate low activation structural material for nuclear fusion reactors. The properties of V-4Cr-4Ti during its operation strongly depend on initial microstructure, especially size and density of Ti-rich precipitates and their interaction with dislocations. This paper focuses on the effects of the precipitate states and other microstructural variables on mechanical properties of V-4Cr-4Ti alloys and the potential methods to improve the properties by microstructural control, based on Transmission Electron Microscopy.
Abstract: We developed a new method to quantify the He atoms in the SiC/SiC Composites Irradiation Behaviour in Fusion Reactor Environment Conditions. We focus on the cavities measurements of the energy shift of the He K-edge of the electron energy loss spectrum. The method is based on mapping the density of He atoms inside the measured bubbles. Combining with the number density and the average diameter of the cavities, we are able to distinguish and estimate the amount of He atoms in the cavities and in the matrix, respectively. We found that small bubbles contain much higher density of He atoms than the larger ones. At irradiation temperature above 1000oC, all the He atoms are trapped inside the bubbles. On the contrary, it is not the case at the lower irradiation temperature.
Abstract: Ferritic/Martensitic steels, with chromium contents ranging between 9 and 12%, were introduced into fusion material programs due to their better creep resistance and excellent thermal and nuclear properties compared to austenitic stainless steels. Reduced activation ferritic/martensitic (RAFM) steels are considered promising candidates for the test blanket modules of the future International Thermonuclear Experimental Reactor (ITER), being EUROFER steel is the EU reference material. It is a 9 % Cr RAFM steel which exhibits a tempered martensitic microstructure and presently allows operation up to 550 ⁰C. This paper shows the work carried out to develop at a pilot plant scale a Reduced Activation Ferritic/Martensitic (RAFM) steel, Asturfer ®, with chemical composition and mechanical properties very close to EUROFER steel.
Abstract: Liquid lithium lead (LiPb) eutectic is considered as one of the promising candidates of tritium breeder materials for fusion reactors. Series experiments on compatibility of LiPb with candidate structural materials such as CLAM steel and SiC f/SiC composites have been done in DRAGON serious experimental devices in FDS team such as DRAGON-RTand stirred pot device in NIFS at 500 oC and 600oC, respectively. The weight loss of CLAM specimens exposed in flowing LiPb with the velocity of 0.17m/s increased with temperature, and the morphology and composition of the corroded surfaces were done by SEM observation and EDX analysis. The coating specimens including Al 2O3 and FeAl/Al2O3 coatings prepared on the CLAM specimens were also exposed in the DRAGON-RT device, the results revealed that there was no obvious thinning observed on the outer surface of the protective coating. Preliminary analysis of SiC f/SiC composites specimens indicated that the mullite coating with plasma spray method on the SiC f/SiC composites specimen corroded in the high temperature LiPb, but no obvious corrosion attack was observed on the specimen surface, while the matrix and fiber of reaction-sintered composites showed slightly corrosion attack after exposure in static LiPb at 800°C for 200 hrs. Further corrosion experiment will be carried out in the near future.
Abstract: Metal hydrides have high hydrogen atom density, which is equivalent to that of liquid water. Fast neutrons are efficiently moderated by hydrogen in metal hydrides. Metal hydrides have been studied for their potential application as nuclear materials in fast reactors (FRs). Two types of the utilizations of metal hydride in FRs are discussed in this paper. One is the utilization for transmutation target of long-lived nuclear wastes. Hydride fuel containing 237Np, 241Am and 243Am has been studied as a candidate transmutation target to reduce the radioactivity of long-lived nuclides included in reprocessed nuclear wastes.
An application of the hafnium hydride has been investigated as neutron absorber in FRs. The core design has been performed to examine its characteristics and to evaluate the cost reduction effect. Demonstration of fabrication of hydride pin has been done with hydride pellets and stainless steel cladding. Coating technique of inner cladding surface has been also developed to reduce the permeation of hydrogen through stainless steel cladding. Physical and chemical properties of the pellet have been measured for designing the hydride pin. The integrity of the pellets at high temperature has been tested and their compatibility with sodium has also been tested. Irradiation test of hydrides has been performed in the fast experimental reactor, JOYO, at Japan Atomic Energy Association (JAEA).
Abstract: Graphite is used as a moderator for high temperature gas-cooled reactor (HTGR) because the microstructure ofwithin the graphite can store defect energy (Wigner energy) induced by high-doses radiation. However, typically, there are temperature spikes associated with the random releases of enormous amount of defect energy. The irregular events of temperature spikes subsequently heat up HTGR and risk the HTGR experiencing beyond the designed-limit temperature limitsof HTGR. To further investigate the graphite-microstructure evolution subject to radiation damage, we develop a progressive method. We quantify the structural integrity of nuclear-grade graphite using High-resolution Transmission Electron Microscopy (HRTEM). In this study, the nuclear-grade graphite was first radiated by ion implantation with 3Mev C2+ to simulate the material in a very-high-temperature-gas-cooled-reactor (VHTGR)-core environment. The temperatures were controlled in the ranges of 500~1000oC. After the artificial irradiation, HRTEM analyses were proceeded to quantify the microstructure evolution and we calculated the stored energy within the microstructure of the radiated nuclear-grade graphite via Fast Fourier Transform of HRTEM images. Our results show that, at 600 oC and 3 dpa, the microstructure can store energy up to 136 (cal/g), which can heat the HTGR up to 345 oC.