摘要
熔盐堆运行及乏燃料后处理过程中产生的氟化物熔盐核废物,由于氟元素在玻璃体中较低的溶解度和对玻璃网络的破坏性,导致其固化处理一直是放射性废物管理领域的难题之一。为有效解决氟化物熔盐核废物固化处理的难题,本文利用H_(2)C_(2)O_(4)作为脱氟剂,开展了H_(2)C_(2)O_(4)对模拟氟化物熔盐废物的脱氟实验研究。利用综合热分析、物相分析及化学成分分析手段,确定了H_(2)C_(2)O_(4)与氟化物熔盐在热处理过程中的脱氟反应,通过研究热处理温度与H_(2)C_(2)O_(4)掺入量对脱氟率的影响,确定了最优脱氟工艺参数。结果表明,H_(2)C_(2)O_(4)与F的摩尔比为2、热处理温度为300℃时,脱氟率可达93%。脱氟过程中H_(2)C_(2)O_(4)与氟化物生成HF气体,其中的金属阳离子形成草酸盐,并在500℃转化为碳酸盐。对脱氟后废物进行硼硅酸盐玻璃固化处理,废物负载量(质量分数)为25%时,玻璃固化体化学稳定性优异。
The molten salt reactor,utilizing molten fluoride salt as fuel solvent and coolant,is one of the generation-Ⅳnuclear reactors for advanced nuclear energy.The generated fluoride nuclear waste during the operation of molten salt reactor and reprocessing of spent fuel should be immobilized in a stable matrix before disposal.At present,vitrification is still the only technology in the world that could industrially immobilize high-level waste.However,such fluoride waste cannot be directly vitrified into borosilicate glass since it contains a large amount of F,which could lead to oversaturation in the glass.In order to provide an effective solution for the safe treatment of molten fluoride salt waste,H_(2)C_(2)O_(4)was used as a defluorination agent to defluorinate the simulated salt waste mainly consisting of alkali fluorides.The mixed sample of H_(2)C_(2)O_(4)and simulated fluoride salt waste was characterized with TG-DSC-MS and XRD to analyze the endothermic and exothermic behaviors,released gases,and phase transitions at elevated temperatures.Afterward,the effects of thermal treatment temperature and the molar ratio of H_(2)C_(2)O_(4)to F on the fluorine removal efficiency were investigated,and the optimal process parameters of defluorination were finally determined.The results show that H_(2)C_(2)O_(4)could react with alkali fluoride salt to release HF gas besides being decomposed into H2O,CO and CO2at temperatures between 100℃and 300℃,while the alkali fluorides turned to alkali oxalates,which could decompose to alkali carbonates at temperatures of 500℃.The fluorine removal efficiency would reach up to 93%when the molar ratio of H_(2)C_(2)O_(4)to F was 2 and the thermal treatment temperature was 300℃.The defluorinated waste thus obtained was then immobilized in a borosilicate glass waste form at about 1200°C with a waste loading of 25%,and the normalized elemental(B,Li,Na,K,Cs,Sr,and Ce)releases of the waste glass conducted with 7-day product consistency test were lower than 2.0 g/m^(2),showcasing acceptable durability for nuclear waste glass.The above results indicate that the proposed approach which defluorination with H_(2)C_(2)O_(4)in the first step at temperatures of below 300°C and vitrifying the remaining waste into a borosilicate glass in the second step,would provide a practical way to safely treat the molten fluoride salt nuclear waste.
作者
董要港
贾子强
徐凯
DONG Yaogang;JIA Ziqiang;XU Kai(State Key Laboratory of Silicate Materials for Architectures,Wuhan University of Technology,Wuhan 430070,China)
出处
《原子能科学技术》
EI
CAS
CSCD
北大核心
2023年第3期478-484,共7页
Atomic Energy Science and Technology
基金
国家自然科学基金面上项目(21876138)
国家重点研发计划(2018YFB1900203)。
关键词
熔盐堆
核废物
氟化物
脱氟
molten salt reactor
nuclear waste
fluoride salt
defluorination
