Ion exchange resins are commonly employed in the treatment of liquid radioactive waste generated in nuclear power plants (NPP). The ion exchange resin used in NPP is a mixed-bed ion exchange resin known as IRN-150, which is of nuclear grade. This resin is a mixture of cation exchange resin and anion exchange resin. The cation exchange resin removes cationic radionuclides such as Cs and Co, while anion exchange resin handles anions (e.g., H14CO3 -), effectively purifying the liquid waste. Spent ion exchange resins (spent resin) containing C-14 are classified as low and intermediate level radioactive waste, and their radioactivity needs to be reduced as it exceeds the disposal limit regulated by law. Therefore, the microwave technology for the removal of C-14 from spent resin has been investigated. Previous studies have successfully developed a method for the effective removal of C-14 during the resin treatment process. However, it was observed that, in this process, functional groups in the resin were also removed, resulting in the generation of off-gases containing trimethylamine. These off-gases can dissolve in water from process, increasing its pH, which can subsequently hinder the recovery of C-14. In this study, we investigated the high-purity recovery of C-14 by adjusting the moisture content within the reactor following microwave treatment. Mock spent resins, consisting of 100 g of resin with HCO3 - ion-exchanged and 0, 25, or 50 g of deionized water, were subjected to microwave treatment for 40 or 60 minutes. Subsequently, the C-14 desorption efficiency of the mock spent resins was evaluated using an acid stripping process with H3PO4 solution. The functional group status of the mock spent resins was analyzed using 15N NMR spectroscopy. The results showed that the mock spent resins exhibited efficient C-14 recovery without significant functional group degradation. The highest C-14 desorption efficiency was achieved when 25 g of deionized water was used during microwave treatment.

본 연구는 전기투석과 용매추출을 융합한 희유금속 회수 공정에서 분리막과 음이온교환막의 개질을 통해 유기상 과 수상에 대한 분리막의 낮은 젖음성 및 AEM을 통한 수소이온 투과로 인한 금속이온의 회수 효율 감소를 개선하였다. 구체 적으로, 분리막 표면 중 한면은 polydopamine (PDA) 통한 친수성 개질, 다른 면은 SiO2 또는 graphene oxide를 통한 친유성 개질을 함으로써 분리막의 젖음성을 개선하였다. 또한, 음이온교환막의 표면을 polyethyleneimine, PDA, poly(vinylidene fluoride) 등을 이용, 개질해 수분 흡수(Water uptake) 감소 및 기공구조 변화를 통해 수소이온 수송을 억제해 수소이온 투과를 억 제할 수 있다. 개질된 막 표면 형상과 화학적 특성 및 조성은 주사전자현미경과 푸리에변환 적외선 분광법을 통해 확인되었 고, 이를 구리 이온 회수 시스템에 적용해 향상된 추출 및 탈거 효율과 수소이온 수송 억제능을 확인하였다.

수산화리튬(LiOH)에 대한 수요는 현재의 대안들에 비해 환경에 대한 효율성과 안전성 때문에 매년 증가하고 있 다. 리튬은 다른 염분과 염수 호수에서 발견될 수 있으며, 나중에 합성되어 다양한 용도로 LiOH를 생성한다. 리튬 이온을 분 리 및 회수하기 위해 다양한 방법이 사용되며, 그 중 가장 일반적인 방법은 전기투석법(ED)이다. ED는 이온을 한쪽에서 다 른 쪽으로 밀어내는 구동력으로서 그 층의 전위차에 작용하는 멤브레인 기반 분리 기술이다. ED의 이온교환막(IEM)은 유체 역학적 부피에 따라 상이한 이온의 선택성이 달라지기 때문에 공정을 효율적으로 만든다. 본 총설에서는 리튬이온의 회수를 향상시키기 위한 ED와 IEM의 서로 다른 변화 전략이 논의된다.

We established pretreatment method of solidified cement ion-exchange resin samples generated before 2003 in nuclear power plants for measurement of non-volatile radionuclide activity. A microwave digestion system (MDS) with mixed acid (HCl-HNO3-HF-H2O2) was used to dissolve cement and to desorb non-volatile elements such as Ce, Co, Cs, Fe, Nb, Ni, Re, Sr and U from mixed ion-exchange resin. The content of Ce, Co, Fe, Nb, Ni, Re, Sr, U and Cs after pretreatment of cement plus mixed ion-exchange resin was measured by ICP-AES and ICP-MS, respectively. As iron and strontium are also present in cement, their content after dissolving a certain amount of cement was measured by ICP-AES. All elements except Nb were quantitatively recovered. Especially since the Nb recovery was low at 72.0±2.5%, the MDS following addition of the mixed acid to the resin was operated once more for desorbing Nb from it. Finally the recovery of Nb was over 95%. This sample pretreatment method will be applied to solidified cement ion-exchange resin samples generated in nuclear power plants for assessment of radionuclide inventory.

In this study, an experiment is performed to recover the Li in Li2CO3 phase from the cathode active material NMC (LiNiCoMnO2) in waste lithium ion batteries. Firstly, carbonation is performed to convert the LiNiO, LiCoO, and Li2MnO3 phases within the powder to Li2CO3 and NiO, CoO, and MnO. The carbonation for phase separation proceeds at a temperature range of 600oC~800oC in a CO2 gas (300 cc/min) atmosphere. At 600~700oC, Li2CO3 and NiO, CoO, and MnO are not completely separated, while Li and other metallic compounds remain. At 800 oC, we can confirm that LiNiO, LiCoO, and Li2MnO3 phases are separated into Li2CO3 and NiO, CoO, and MnO phases. After completing the phase separation, by using the solubility difference of Li2CO3 and NiO, CoO, and MnO, we set the ratio of solution (distilled water) to powder after carbonation as 30:1. Subsequently, water leaching is carried out. Then, the Li2CO3 within the solution melts and concentrates, while NiO, MnO, and CoO phases remain after filtering. Thus, Li2CO3 can be recovered.

The characteristics of aqueous lithium recovery by ion exchange were studied using three commercial cation exchange resins: CMP28 (porous type strong acid exchange resin), SCR-B (gel type strong acid exchange resin) and WK60L (porous type weak acid exchange resin). CMP28 was the most effective material for aqueous lithium recovery; its performance was even enhanced by modifying the cation with K+. A comparison to Na+ and H+ form resins demonstrated that the performance enhancement is reciprocally related to the electronegativity of the cation form. Further kinetic and equilibrium isotherm studies with the K+ form CMP28 showed that aqueous lithium recovery by ion exchange was well fitted with the pseudo-second-order rate equation and the Langmuir isotherm. The maximum ion exchange capacity of aqueous lithium recovery was found to be 14.28 mg/g and the optimal pH was in the region of 4-10.