退役核设施不锈钢中氚的分析  

Analysis of Tritium in Stainless Steel of Decommissioned Nuclear Facilities

作　　者：文炜[1] 孙宇[1] 邱永梅[1] 但贵萍[1] 

机构地区：[1]中国工程物理研究院核物理与化学研究所,四川绵阳621900

出　　处：《同位素》2012年第2期78-81,共4页Journal of Isotopes

摘　　要：在自制的氚污染不锈钢的高温热解析制样系统中对氚污染不锈钢样品进行解析分析,并对影响不锈钢样品中氚回收率的解析温度、解析时间、催化氧化效率、冷凝收集效率等条件进行了优化研究。结果表明,当载气(空气)流速为60～80L/h,解析温度和催化氧化温度为700℃,冷凝温度为-20℃,解析时间为2h时,氚污染不锈钢样品高温制样系统的冷凝收集效率为99%,催化氧化效率为98.6%,不锈钢样品中氚的解析率>95%。以上结果表明,该不锈钢高温热解析制样系统及其解析条件能够满足退役核设施不锈钢样品中氚的分析要求。A thermal vacuum analytical system was established for the determination of tritium in stainless steel samples.The releas temperature,time,catalyst efficiency and condensed recovery were described in detailed.The results indicated that the condensed recovery,catalyst efficiency,and released coefficient were 99%,98.6%,and more than 95% respectively,when the stainless steel samples were heated for 2 hours at 700 ℃ under the following conditions： the condenser temperature at-20 ℃ and the flow velocity of the compressed air at 60-80 L/h.This system could meet the need for the determination of tritium in stainless steels samples during decommissioning.

关 键 词：退役核设施 不锈钢 氚 热解析 

分 类 号：TL943[核科学技术—辐射防护及环境保护] TL944