| Fusion energy produced by fusion reaction is an important potential way to solve the energy problem of human. Recent researches on fusion reactor have made significant progress, and the resulting material problem has become a realistic problem because that the materials in the reactor will face harsh working environment. Tungsten (W) is considered to be the potential candidate for plasma facing materials like first wall in future fusion reactors owning to a series of advantages including high melting point, low thermal expansion coefficient, high thermal conductivity, low tritium inventory and high energy threshold for sputtering, etc. However, serious problems that limit the use of W still exist,such as low-temperature brittleness, recrystallization embrittlement. Defects induced by irradiation often lead to embrittlement of the material,thus shortening the service life of components. Therefore, it will be of great significance to reduce brittleness and alleviate irradiation demage of W through designing composition/structure/organization.In this paper, the effect of the activation pretreatment of the original W powder on the sintering performance was studied, providing a new way to improve the sintering densification of W and W-based alloys; rare-earth oxide dispersion strengthened W-based composites (W-Pr2O3?W-CeO2 and W-Nd2O3) and carbide dispersion strengthened W-based composites (W-TiC and W-ZrC) were fabricated by performing the wet chemical method and spark plasma sintering (SPS) technique, and the effects of the addition of rare earth oxides and carbide second phase particles on the microstructure and properties of the composites were studied; in addition, the irradiation damage effect of deuterium,helium and iron ion on as-synthesized W-based composites were investigated. The results are as follows:(1) W powder was subjected to chemical activation pretreatment by immersion into an aqueous solution of hydrofluoric acid and ammonium fluoride,and then was reinforced with an ultrasonic wave at room temperature to obtain uniform distribution of defects on the surface of W powder, increasing the specific surface area of powder. The activated W sintered sample showed high relative density and was more densified compared with initial W after sintering at 1800 ? and 2200 ? under the same sintering process.Therefore, to have the same final density, initial specimens must be sintered at a higher temperature. The W-1.5wt%TiC composites after activation treatment of sintered powder achieve a relative density 7% higher than that of the original powder and more compact organization after surface activation pretreatment for W and TiC povwders.(2) Rare-earth oxide dispersion strengthened W-based composites with fine grain and high relative density were successfully synthesized via wet chemical method and SPS technique, doped with highly uniform oxide (Pr2O3, CeO2 and Nd2O3) at the tungsten grain boundaries as well as within the grains. The relative density of the W-1wt%Pr2O3 sample was approximately 98.3%, while the relative density of the W-1wt%Nd2O3 and W-1wt%CeO2 were 96.5% and 95.9%, respectively. Second phase oxide particles were effective in hindering the grain growth of tungsten in the sintering process. The average grain size of rare-earth oxide doped composites was about 4 ?m,while pure tungsten grew to be 10 ?m. The thermal conductivity, of composites decreased with increasing temperatures. However, the thermal conductivity of W-1wt%Pr2O3 is above 150 W/m·K at room temperature, and the value of W-1wt%Nd2O3 and W-1wt%CeO2 also above 140 W/m.K.(3) Carbide doped composite powders with core-coating structure were prepared by wet chemical process, and W-lwt%TiC composites with fine grain and high relative density were fabricated by SPS. The average grain size was approximately 4?m, and the W-1wt%TiC sample has the relative density of 98.6%, higher than the W-lwt%ZrC of 97.2% and higher microhardness. The thermal diffusivities and thermal conductivities of W-1wt%TiC composites decreased with increasing temperatures. The thermal conductivity of W-1wt%TiC is 127 W/m·K at room temperature, but the thermal conductivity of W-1wt%ZrC is only 110 W/m·K.(4) Irradiation of 5 keV D+ to low fluence of 1×1020D+/m2 at room temperature, W-1wt%Pr2O3, W-1wt%La2O3, W-1wt%TiC and W-1wt%ZrC showed no damage.micro structure like pure W. However, different from pure W, when fluence increased to 1×1021D+/m2,there were still no obvious defects in the four composites, indicating that the addition of rare-earth oxide and carbide particles can improve the resistance to D+ ion irradiation of W. After irradiation with 5×1021D+/m2, all samples were annealed isothermally for varying temperatures, resulting in dislocation loop number volume density lower. A complete removal of dislocation structures occurred at 800 ?. When irradiation of 5 keV He+ achieve to the fluence of 5×10/He+/m2 at room temperature, lots of helium bubbles were noticeably formed in pure W. But there were still no helium bubbles forming in the four composites irradiated with 5×1021He+/m2 at 600 ? ,demonstrating that doping with rare-earth oxide and carbide particles can improve the resistance to He+ ion irradiation of W materials for fusion reactor.(5) With increasing irradiation fluences of deuterium, retained amount of deuterium became larger. However, the deuterium retention of the four composites were significantly lower than that of pure W, proving that the addition of rare-earth oxide and carbides decreased the deuterium retention of W materials. Pre-irradiation using 5 keV He+to fluence of 1 ×1021He+/m2 resulted in an increase of deuterium retention. The four samples did not have additional desorption peaks as compared with samples without He+pre-irradiation, meaning that He+ pre-irradiation did not introduce additional kinds of traps. There was a clear saturation for the deuterium trapping by the defects produced by He+ pre-irradiation when the fluence increased to 1×1021D+/m2.(6) Irradiation of 1 MeV Fe+ to the fluence of 1dpa at room temperature,dislocation networks formed with many small dislocation loops embedded within the networks in W-1wt%Pr2O3, W-1wt%La2O3 and W-1wt%ZrC. But no networks formed in W-1wt%TiC,indicating that W-lwt%TiC has a better resistance to Fe+ irradiation damage. Average dislocation loop size increased and loop density decreased with increasing annealing temperature.(7) The high temperature peaks (or shoulders) in TDS spectra occurred between 700 K and 900 K for Fe+ pre-irradiation samples and did not appear for without Fe+ pre-irradiation samples. With increasing irradiation dose, the position of high temperature peaks (or shoulders) became more obvious, implying that these peaks were caused through the release of deuterium trapped by the defects produced by Fe+ pre-irradiation.Compared with the samples without Fe+ pre-irradiation, deuterium retention of four composites increased after Fe+ pre-irradiation, and retained amount of deuterium was also increased with the increase of Fe+ irradiation dose. |
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