核电厂典型中低放射性废物等离子体熔融处理试验研究

Investigation of Thermal Plasma Melting of Typical Intermediate and Low Level Radioactive Wastes from Nuclear Power Plant

选取国内压水堆核电厂两种典型的中低放射性废物开展等离子体熔融中试试验,在1250℃温度条件下熔融100 min,熔制成完全玻璃态的固化体,两种玻璃固化体的XRD衍射谱均呈现为典型的非晶态谱;并选取了非放射性的Co_2O_3、SrCO_3、CsCl作为放射性核素示踪剂,模拟放射性核素^(137)Cs、^(90)Sr、^(58)Co、^(60)Co在核电站放射性废物等离子熔融处理过程中的包容情况;经检测,玻璃固化体物理性能、抗浸出性能以及机械性能满足高放玻璃固化体要求,且机械性能优于水泥固化体标准;最后对后续试验进行了展望,并提出了需要进一步解决的问题。

In this pilot scale testing research, two typical intermediate and low level radioactive wastes from the nuclear power plant were vitrified with thermal plasma. Both samples were melted at 1250℃ for 100min, and the XRD patterns of both glass waste-forms presented as the typical amorphous state. Co2O3, SrCO3, CsCl had been added to the original wastes as the radionuclide tracer for simulating the containment effect of radionuclide in the process of vitrification. The physical, leaching and mechanical performance of resultant glass waste-forms were tested, showing the comparable requirement with the high level radioactive waste glasses and better mechanical properties than cement waste-form. Finally, we discussed the prospect of the current work and proposed issues to be resolved.