Various disposal methods for spent nuclear fuels (SNFs) are being researched, and one of these methods involves separating high heat-generating nuclear isotopes such as Strontium-90 (90Sr) and Cesium-137 (137Cs) for deep disposal. These isotopes has relatively short half-lives and substantial decay energies. Especially, 90Sr undergoes decay through Yttrium-90 to Zirconium-90, emitting intense heat with beta radiation. Therefore, the removal of these high heat-generating isotopes will significantly contribute to reducing disposal site area. To remove 90Sr from SNFs, molten salt was utilized in KAERI. During this process, it was discovered that 90Sr dissolves in the molten salt in the form of SrCl2 and/or Sr4OCl6. Afterwards, it is crucial to recover 90Sr in the form of oxide from the salt to create immobilized forms for disposal. This can be achieved by reactive distillation with K2CO3. However, the amount of 90Sr within the SNFs is only 0.121wt%, and even if all the 90Sr in the SNFs were to leach into the molten salt, the quantity of 90Sr in the molten slat would still be very small. Therefore, adding K2CO3 to the molten salt for reactive distillation could result in significant possibilities of side reactions occurring. In this study, a two-step process was employed to mitigate the side reactions: the 1st step involves evaporating the all molten salts and the 2nd step includes adding K2CO3 to make oxides through solid-solid reaction. Eutectic LiCl-KCl, which is the most commonly used salt, was employed. The eutectic LiCl-KCl with SrCl2 was heated at 850°C for 2 h to evaporate the salts under a vacuum (> 0.02 torr). However, after examining the distillation product before the solid-solid reaction, it was observed that SrCl2 reacted with KCl in the salt, resulting in the formation of KSr2Cl5. It means that salts containing KCl are not suitable candidates for reactive distillation aimed at producing immobilized forms. As an alternative, MgCl2 could be a highly promising candidate because it is inert to SrCl2 and according to a recent study in KAERI, MgCl2 exhibited the most efficient separation of Sr among various salts. Therefore, we plan to proceed with the two-step reactive distillation using MgCl2 for the future work.

현재, 국내에서는 가공식품인 식용유지에 대한 잔류농약 허용기준이 설정되어 있지 않아 잔류농약은 식용유 품질평 가의 사각지대라 할 수 있다. 본 연구에서는 식용유지에서 가열증류법을 이용하여 68종의 농약을 대상으로 추출 및 정 제법을 최적화하여 GC-MS/MS 분석법을 확립하였다. 가열 증류법은 가열온도 및 시간의 영향을 받았으며 이동상의 역 할을 하는 질소의 유량과 용출용매의 종류에는 큰 영향을 받지 않는 것으로 나타났다. 본 연구에서 잔류농약의 결정 계수(R2)는 0.99 이상으로 나타났고, 정량한계(LOQ)는 0.01- 0.02 mg/L이었으며, 대두유를 이용하여 0.01, 0.02, 0.1, 0.5 mg/L 수준으로 회수율 실험 결과 평균 회수율(n=5)은 66.1- 120.0%이었고 상대표준편차는 ±10%이하로 나타났다. 또한 실험실내 일간정밀도는 11%이하로 조사되어, 식품의약품안 전평가원의 ‘식품등 시험법 마련 표준절차에 관한 가이드라 인(2016)’에 적합한 수준임을 확인하였다. 따라서 본 연구에 서 개발한 시험법은 식용유지의 잔류농약 안전관리를 위한 시험법으로 활용이 가능할 것으로 기대된다.

A phosphorylation (phosphate precipitation) technology of metal chlorides is considering as a proper treatment method for recovering the fission products in a spent molten salt. In KAERI’s previous precipitation tests, the powder of lithium phosphate (Li3PO4) as a precipitation agent reacted with metal chlorides in a simulated LiCl-KCl molten salt. The reaction of metal chlorides containing actinides such as uranium and rare earths with lithium phosphate in a molten salt was known as solidliquid reaction. In order to increase the precipitation reaction rate the powder of lithium phosphate dispersed by stirring thoroughly in a molten salt. As one of the recovery methods of the metal phosphates precipitated on the bottom of the molten salt vessel cutting method at the lower part of the salt ingot is considered. On the other hand, a vacuum distillation method of all the molten salt containing the metal phosphates precipitates was proposed as another recovering method. In recent study, a new method for collecting the phosphorylation reaction products into a small recovering vessel was investigated resulting in some test data by using the lithium phosphate ingot in a molten salt containing uranium and three rare earth elements (Nd, Ce, and La). The phosphorylation experiments using lithium phosphate ingots carried out to collect the metal phosphate precipitates and the test result of this new method was feasible. However, the reaction rate of test using lithium phosphate ingot is very slower than that of test using lithium phosphate powder. In this presentation, the precipitation reactor design used for phosphorylation reaction shows that the amount of molten salt transferred to the distillation unit will reduce by collecting all of the metal phosphates that will be generated using lithium phosphate powder into a small recovering vessel.

This study presents a rapid and quantitative sequential separation method for H-3 and C-14 isotopes with distillation apparatus in environmental samples released from nuclear facilities. After adding 200 mg of granulated potassium permanganate and 500 mg of sodium hydroxide in 100 mL of sample solution, the sample solution was heated until approximately 10 mL of distillate, and the distillate fraction was removed. The sample solution was heated again until a minimum 10 mL of additional distillate was collected. 10 mL of distillate was transferred to the LSC vail and the measurement sample for H-3 was made by adding 10 mL of Ultima Gold LLT to the LSC vial. After adding 2.5 g of potassium persulfate, 2 mL of 1M silver nitrate and 15 mL of concentrated nitric acid to the remained sample solution, the sample solution was heated for 90 minutes and C-14 isotopes were adsorbed into 10 mL of Carbo-Sorb solution in glass vial. The measurement sample for C-14 was made by adding 10 mL of Permafluor to the C-14 fraction in glass vial. The purified H-3 and C-14 samples were measured by the liquid scintillation counter after quenching correction. The average recoveries of H-3 and C-14 with CRM were measured to be 96% and 85%, respectively. The sequential separation method for H-3 and C-14 investigated in this study was applied to activated charcoal filter produced from nuclear power plants after validating the reliability by result of proficiency test (KOLAS-KRISS, PT-2021-51).

The purpose of this study was to effectively purify U-contaminated soil-washing effluent using a precipitation/distillation process, reuse the purified water, and self-dispose of the generated solid. The U ions in the effluent were easily removed as sediments by neutralization, and the metal sediments and suspended soils were flocculated–precipitated by polyacrylamide (PAM). The precipitate generated through the flocculation–precipitation process was completely separated into solid–liquid phases by membrane filtration (pore size < 45 μm), and Ca2+ and Mg2+ ions remaining in the effluent were removed by distillation. Even if neutralized or distilled effluent was reused for soil washing, soil decontamination performance was maintained. PAM, an organic component of the filter cake, was successfully removed by thermal decomposition without loss of metal deposits including U. The uranium concentration of the residual solids after distillation is confirmed to be less than 1 Bq·g−1, so it is expected that the self-disposal of the residual solids is possible. Therefore, the treatment method of U-contaminated soil-washing effluent using the precipitation/distillation process presented in this study can be used to effectively treat the washing waste of U-contaminated soil and self-dispose of the generated solids.