Boric acid-containing B-10 is used in a nuclear reactor as a coolant and absorbs thermal neutrons generated during nuclear fission in the primary circuit. Boron-containing coolant water waste is generated from maintenance, floor drain, decontamination, and reactor letdown flows. There are two options for aqueous solution waste of boric acid. One is recycling and discharge through filtration, ion exchange, and reverse osmosis. The other is immobilization after evaporation and crystallization processes. The dry powder of boric acid waste liquid can be immobilized by cement, polymer, etc. Before the mid-1990s, concentrated boric acid waste was solidified with a cement matrix. To overcome the disadvantage of low waste loading of cement waste form, a method of solidifying with paraffin was adopted. However, paraffin solids were insufficient to be disposed of as final waste. Paraffin is a kind of soft solidified material and has low compressive strength and poor leaching resistance. As a result, it was decided as an unsuitable form for disposal. In KOREA, paraffin waste form was adopted for boric acid waste treatment in the 1990s. A large amount of paraffin waste forms about 20,000 drums (200 l drum) were generated to treat boric acid waste and were stored in nuclear power sites without disposal. In this study, we want to obtain high-purity boric acid waste by oxidizing and decomposing solid paraffin waste form through a boric acid catalytic reaction. In this reaction, paraffin is separated in the form of various by-products, which can then be treated through a liquid waste treatment device or an exhaust gas treatment device. The proper temperature for sample decomposition during the catalytic reaction was set through TGA analysis. Compositions of by-products and residues generated at each stage of the reaction could be analyzed to determine the state during the reaction. Finally, the boric acid waste powder was perfectly separated from paraffin waste form with disposable products through this pyrolysis process.

Radioactive carbon, C-14, can be generated by the neutron capture reaction of O-17 during the nuclear power plant operation. Since C-14 is classified as an intermediate level waste radionuclide, it is required that an effective separation process for C-14. C-14 is mainly absorbed on activated carbon in the air cleanup system. Therefore, the main generation source of C-14 during the nuclear power plant decommissioning is spent activated carbon. KAERI has been developing the treatment of spent activated carbon. In this process, C-14 can be desorbed as a gaseous oxide form from the spent activated carbon at high-temperature vacuum conditions. This radioactive carbon dioxide can be captured into alkaline earth metal incorporated glass and can be transformed into carbonate form. However, the carbonate (e.g. CaCO3 and SrCO3) is dispersive. When the radioactive carbonates are disposed into a geological repository, they should be immobilized to remove future uncertainty. This study examined the stabilization/immobilization of the radioactive carbonates by the cement hydration process. Cement wasteform incorporated with calcium carbonate and strontium carbonate was produced under various waste loading (e.g. 20wt%, 40wt%, and 60wt% of CaCO3 and SrCO3, respectively). Then we evaluated mechanical and chemical durability by measuring compressive strength and leachability according to standard test methods specified in the waste acceptance criteria of the Gyeongju low and intermediate level waste repository (WAC-SIL-2022-1). Also, microstructure and thermal characteristics were investigated by SEM-EDS and TGA analysis.

Legacy waste from the decommissioned A-1 nuclear power plant in the Slovak Republic is scheduled for immobilisation within a tailored alkali borosilicate glass formulation, as part of ongoing site cleanup. The aqueous durability and characterisation of a simulant glass wasteform for Chrompik III legacy waste, was investigated, including dissolution experiments up to 112 days (90°C, ASTM Type 1 water). The wasteform was an amorphous, light green glassy product, with no observed phase separation or crystalline inclusions. Aqueous leach testing revealed a suitably durable product over the timescale investigated, comparing positively to other simulant nuclear waste glasses and vitreous products tested under similar conditions. Iron and titanium rich precipitates were observed to form at the surface of monolithic samples during leaching, with the formation of an alkali deficient alteration layer behind these at later ages. Overall this glass appears to perform well, and in line with expectations for this chemistry, although longer-term testing would be required to predict overall durability. This work will contribute to developing confidence in the disposability of vitrified Chrompik legacy wastes.

본 연구에서는 전해환원공정에서 발생하는 폐용융염에서 LiCl을 재활용하기 위해 핵종제거 물질로 제올라이트를 사용할 때, 발생하는 폐제올라이트와 여기에 흡착된 유리염을 고감용으로 고화하는 경우의 고화체특성을 살폈다. 주종 핵종인 Cs의 침출속도는 붕규산유리보다는 석회유리로 고화한 경우, SAP과의 반응비와 유리의 첨가량을 변화시켜도 그 값은 1/10 정도로 낮았으며 그 범위는 0.1에서 0.01g/m2d이었다. 한편으로 Sr의 침출속도는 유리의 종류와 첨가량변화에 크게 지배를 받지 않으며 Cs보다 훨씬 낮은 0.001에서 0.0001g/m2d이었다. 그리고 압축강도는 유리의 함량이 증가할수록 감소하였고, 열팽창율은 어떤 온도에서 도 유리를 30% 함유한 것이 가장 적게 나타났다. 한편으로 이 고화체들의 용융온도는 약 1,100℃로서 유리의 함량이 증가하면 약간씩 높아졌다.