Irradiation Growth Modeling of Zirconium in Nuclear Reactors

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Irradiation Growth Modeling of Zirconium in Nuclear Reactors
Choi, Sang Il
Kim, Ji Hyun
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Graduate school of UNIST
Zirconium and zirconium alloys are extensively used as nuclear core material due to their excellent ability to withstand reactor conditions including corrosion and neutron irradiation. Despite their excellent properties as structure material, radiation-induced phenomena (growth, hardening, creep, and amorphization) are still main concerns at structure materials. Among them, irradiation growth is main concern in zirconium base alloy. In nuclear reactor condition, main design criteria of zirconium alloys is corrosion. Therefore history of zirconium alloy development has been done to develop it which has better corrosion resistant properties. Representative zirconium alloy in nuclear society such as Zircaloy-1, Zircaloy-2, Zircaloy-4 and Zirlo are materials which show high corrosion resistant property. So far, irradiation growth was not main concern of nuclear plant safety criteria because of corrosion phenomenon. Therefore irradiation growth research is still insufficient compare with corrosion research filed. Nowadays, however, commercial reactor life extension is needed and research reactor is received more neutron flux than commercial thing. Therefore, safety analysis of irradiated materials is become important. Irradiation growth combine with irradiation hardening could induce materials failure. Therefore this is the timing to analysis of irradiated materials. So far, to recognize environment and material parameter effects on growth, extensive experiments have been done and some important experiments were completed. In the extensive research, to figure out the individual sink effects (grain boundary and dislocation loop and line, texture) on the growth mechanism, irradiation growth of single crystal experiments had been done. After single crystal growth data were analyzed, polycrystalline and zircaloy were also researched systemically. From these experiments, major parameter effect on irradiation growth was conformed. After growth experiment, specimen was analyzed by microstructure morphology such as sink density and size. From microstructure change, researcher could analyze irradiation growth mechanism. In 1980s, a lot of research had been done irradiated materials. Eventually, in rate of 1980s, research about irradiation growth, microstructure and growth theory are published. However, recently, microstructure morphology change and fundamental parameter are updated and anisotropy diffusion difference concept was developed. Former research does not contain these developed. Recently theoretical modeling, which contain new developed concept, has been done in single modeling. However irradiation growth of polycrystalline and zircaloy do not developed. Therefore this research object is modeling of irradiation growth from single crystal to zircaloy. To modeling of irradiation growth, in the literature study, microstructure change morphology and growth experiment were systemically reviewed by organization from single crystal to zircaloy. Also theoretical analysis, which conducted various authors are reviewed. At lastly, defect rate theory, which could explain microstructure change fundamentally, is reviewed for theoretical modeling. In the results, based on the defect rete theory, improved theoretical modeling are suggested and growth results are presented from single to zircaloy. Before to growth modeling in research reactor, growth modeling are conducted at commercial reactor condition because there are more extensive research results are exist in commercial reactor condition. These results also conducted systemically form single to zircaloy for represent individual sink effect on growth. Individual parameters effects on irradiation growth are analyze by growth modeling results. Many metallurgical conditions eventually change the material parameters hence texture, grain size, dislocation density effect on growth are analyzed and also experimental conditions effect on growth are reviewed. In case of single crystal and cold worked polycrystalline, the growth result show well agreement with experimental result. For the fundamental understanding of these case, the behavior of the defect flux and sink strength behavior are analyzed. From this analysis, it was clear that the growth modeling are well established. However, in case of annealed polycrystalline case, the results was not well matched with experimental results because in case of polycrystalline case, the sinks behavior are more complex than single and cold worked case. Unfortunately, irradiation growth modeling of annealed polycrystalline was disagreement of experiment result. However, from the detail analysis in result and discussion, it was clear that limitation of the modeling and improvement direction. Therefore, in the future work, assumption of defect rate equation will be extended to cluster and grain boundary sink strength modeling will be conducted. Also irradiation growth equation will be modified to contain axis shortening. Diffusion coefficient is most important parameter in the modeling result. Therefore, defect induced diffusion coefficient change also will be treated. This specific work will be done with MD simulation.
Department of Nuclear Engineering
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