Electrochemical decontamination of irradiated nuclear graphite from corrosion and fission products using molten salt
Electrochemical decontamination of irradiated nuclear graphite from corrosion and fission products using molten salt
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DOI：
10.1039/d1ee00332a
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发表时间：
2021-09-03
影响因子：
32.5
通讯作者：
Sharrad, Clint A.
中科院分区：
文献类型：
--
作者：
Grebennikova, Tatiana;Jones, Abbie N.;Sharrad, Clint A.
辐照石墨废物管理是世界各地核电站退役的主要挑战之一，在英国、法国和俄罗斯尤为明显，这些国家超过85个反应堆使用石墨作为慢化剂或堆芯材料。到目前为止，在辐照石墨中观察到的相当大的体积(全球30万吨)和不均匀的污染表明，这种材料是活性较高的废物，主要完好无损地存在于多个反应堆地点，等待在尚未建成的深层或近地表地质处置设施(GDF)中处置。这项研究提出了一种适用于辐照石墨材料的新的非破坏性处理方法，目的是去除腐蚀和裂变产物中的污染物，从而降低废物的类别，并实现不需要管理储存和处置的加速处置，作为GDF中的中级废物。将一种新的高温熔盐介质中的电化学去污方法应用于辐照的UK Magnox级石墨，固定在工作电极上，浸泡在723K的LiClKCl共晶中，通过选择合适的电流绝对值和正负电流转换次数，实现了对石墨中放射性核素污染(Co-60、Ba-133、Cs-137、Eu-154)的实质性去除。在没有显著降解石墨材料的情况下，Co-60的总初始活性降低了80%(
Irradiated graphite waste management is one of the major challenges of nuclear power-plant decommissioning throughout the world and significantly in the UK, France and Russia where over 85 reactors employed graphite as moderator or core material. To date, the considerable volume (>300 000 tonnes worldwide) and heterogeneity of contamination observed in irradiated graphite designates this material as higher activity waste, which resides predominately intact, at multiple reactor sites awaiting disposal in a deep or near-surface Geological Disposal Facility (GDF) yet to be built. This study presents a novel and non-destructive method for the treatment suitable for irradiated graphite materials, with the aim to remove contamination from corrosion and fission products and therefore downgrade the category of waste and enable accelerated disposal which does not require managed storage and disposal as intermediate level waste in a GDF. A novel electrochemical decontamination approach in a high-temperature molten salt medium was applied to irradiated UK Magnox grade graphite, fixed on the working electrode immersed in LiCl-KCl eutectic at 723 K. By selecting the appropriate value of the absolute magnitude of current and the number of transitions between positive and negative current, substantial removal of radionuclide contamination (Co-60, Ba-133, Cs-137, Eu-154) from the graphite was achieved. Up to 80% reduction of total initial activity for Co-60 was achieved without significant degradation of the graphite material (