摘要: 玻璃固化体是用来固化放射性废物的硅酸盐。作为放射性废物处理的第一道工程屏障，它的耐辐照性能尤其引人注目。本工作使用硼硅酸盐玻璃模拟玻璃固化体材料，用不同剂量的γ射线辐照硼硅酸盐玻璃模拟天然放射性对固化体的辐照损伤。通过测量和分析辐照后硼硅酸盐玻璃的吸收光谱，证实了辐照后在玻璃中产生了E'缺陷，非桥氧空位色心、过氧自由基以及过桥氧联接等缺陷。此外，还得出了不同微观结构随吸收剂量的演化关系。对于辐照后产生的缺陷，它们的浓度都随吸收剂量的增加而增大。同时，发现玻璃在辐照后其吸收光谱的带隙随着剂量的增加而逐渐变窄；而当吸收剂量大于等于10⁵ Gy时，玻璃的带隙则达到饱和值。


Vitrification is one kind of silicates which is used for immobilization of high-level waste (HLW). As the first engineered barrier of HLW disposition, its anti-irradiation characteristic is particularly noticeable. Vitrification is replaced by borosilicate glass to investigate radiation effect, and the irradiation damage generated by natural radioactivity in vitrification is simulated by different doses gamma rays on borosilicate glass. By measuring and analyzing the absorption spectrum of irradiated borosilicate glasses, it is confirmed that E'defect, non-bridging oxygen hole center, peroxy dangling bond and bridge oxygen link, etc. are induced in borosilicate glass after irradiation. Furthermore, the relations between the defects and absorbed doses are shown. For the concentrations of these defects induced by irradiation, all of them increase with absorbed dose. Meanwhile, absorption band gap in borosilicate glass after irradiation decreases with absorbed dose and the band gap becomes saturated when absorbed dose is equal to or greater than 10⁵ Gy.