| Nuclear energy has been developed and utilized over 60 years from its humble beginnings in the first nuclear reaction at the University of Chicago in 1942. Today, as annually reported by the International Atomic Energy Agency(IAEA) in 2014, there are over 434 operational nuclear reactors producing a total capacity of 371.7 GW and supplying approximately 16% of the world’s total power generating capacity. With the gradual consumption of fossil fuel energy, the world will then face energy scarcity. As a clean low carbon source of energy, nuclear energy can help to meet the increasing demands for electricity energy, achieve the energy security and sustainable development objectives. Therefore, the development of nuclear technology will get much higher priority in the future.The investigation on the material adapting to nuclear reactor environment has been stimulated by the development of nuclear energy. The mechanical properties and corrosion resistance in the high temperature water environment are two key factors to assess the feasibility of the materials to be used in the reactors. Zirconium and its alloys have been chosen as the fuel claddings, pressurize pipe and structure materials in the nuclear reactor primarily for their very low thermal neutron absorption cross section, good mechanical property and corrosion resistance. Under a certain harsh environment of nuclear reactor, zirconium alloys also exhibit a good resistance to radiation damage from neutron "ux, electron beam and !-ray. While nuclear reactor is running, the structure material of a zirconium alloy will face neutron and multiple types of ion beam irradiations, which will induce the structure damage in the zirconium alloy. The study of radiation effects of zirconium alloy is important and attracts much attentions.Nowadays, China has successfully become a dominant country in the field of nuclear energy after 30 years rapid development since its first nuclear reactor passed technical appraisement in 1984, thereby increasing the needs of zirconium alloy and the urgency of developing high performance zirconium alloy with self-owned intellectual property. N18 is one of zirconium alloys that were developed in China. Results of outof-pile tests confirmed that N18 zirconium alloy has better corrosion resistance and creep resistance than that of Zircaloy-4, which was attributed to the more densely disposed Zr(Fe,Cr)2 precipitate and niobium solutes in "-Zr. However, research is limited can be found in the literature on the studies of radiation effects, helium bubble evolution and hydrogenation behavior of N18 zirconium alloy. Therefore, a detailed researches will be explored in this dissertation to evaluate the radiation resistance, helium bubble evolution and hydrogenation behavior of the N18 zirconium alloy. The main research and innovations of this dissertation are as follows:(1) The behavior of the precipitate in the N18 zirconium alloy under a 2 MeV proton irradiation was studied by transmission electron microscopy(TEM). The planview TEM specimens were prepared by a focused ion beam(FIB) lift-out method. It is con#rmed that the Zr(Fe,Cr,Nb)2 precipitate with the hexagonal close packed(hcp) structure is the major secondary phase particle(SPP) in this zirconium alloy. At irradiation temperature of 360 oC, the Zr(Fe,Cr,Nb)2 precipitate starts to be partially amorphized when the zirconium alloy was irradiated to 3.9 dpa, and was completely amorphized at 8.2 dpa. The mechanism for the Zr(Fe,Cr,Nb)2 precipitate amorphisation under proton irradiation is attributed to the Fe diffusion from the Zr(Fe,Cr,Nb)2 phase to the zirconium alloy matrix. The current research indicates that the N18 zirconium alloy has lower irradiation resistance than that of Zircaloy-4, but the results still need to be further veri#ed by its irradiation resistance with respect to neutron irradiation.(2) The behavior of the Zr(Fe,Cr,Nb)2 precipitate in the N18 zirconium alloy under a 500 keV Ne ion beam irradiation and post heating were studied by in-situ TEM. It is found that the Zr(Fe,Cr,Nb)2 precipitate with the hexagonal close packed(hcp) structure is the major secondary phase particle(SPP) in this zirconium alloy. At the temperature of 310 oC, the precipitate starts to be amorphized after having been irradiated to 0.5 displacement per atom(dpa). The critical irradiation dose for the precipitate to be completely amorphized is 1.0 dpa. The amorphized Zr(Fe,Cr,Nb)2 precipitate starts to be recrystallized when it was heated up to 600 oC. The temperature for the recrystallization of the amorphized Zr(Fe,Cr,Nb)2 precipitate is determined to be between 450 oC and 600 oC. Instead of forming large single grain crystals, some nano crystals that have the same crystal structure to that of the precipitate before being amorphized were formed during the recrystallization.(3) The formation of helium bubbles is considered to be detrimental to the mechanical properties of the nuclear materials. The growth behaviors of helium bubbles in a helium ion implanted N18 zirconium alloy with respect to the helium fluence and subsequently annealing procedure were investigated by in-situ TEM study. In the as-implanted sample, the measured size distributions of the helium bubbles are consistent with the computer simulated helium concentrations. Moreover, the average size of the helium bubbles increases with the increase of the irradiation temperatures and the helium fluence. The in-situ heating study performed in a TEM indicates that the average size of the helium bubbles increase slowly with the temperature below 650 oC and then increase quickly above 650 oC. The growth mechanism of the helium bubbles in the alloy is suggested based on the study.(4) The hydrogenation behavior of the N18 zirconium alloy was in-situ studied in a FIB system in which, the zirconium alloy takes hydrogen generated from decomposition of an organometallic precursor used for platinum deposition by the gallium ion beam irradiation. The formed zirconium hydride that has a fcc structure was observed to grow as preferentially oriented platelets in the zirconium alloy matrix. The mechanism of the formation of the zirconium hydride is correlated to the stress relaxation of the alloy, as the hydrogenation of the alloy occurred only when the alloy foils have been thinned below a critical thickness below which the foils start to be bent. The dehydrogenation of an individual hydride platelet was in-situ investigated in a TEM on the microscopic scale. Our results on the stability of zirconium hydrides with various hydrogen phases are contrary to those obtained on a macroscopic scale, where it is difficult to discriminate between the stability of zirconium hydride and multiple kinds of material properties and defects of the zirconium matrix. The experiment demonstrates a new way for in-situ study of hydrogenation behaviors of other hydrogen storage alloys, and also provides a general framework for studying the structure properties of individual nanosize precipitate in a metal or alloy without the influence of the matrix. |
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