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.

Korea Atomic Energy Research Institute’s Post Irradiated Examination Facility safely stores spent nuclear fuel using a wet storage method to conduct research. Here, in order to remove the radioactivity released into the water, the stored water is passed through an ion exchange resin tower, and the radionuclides are exchanged with the bead-shaped ion exchange resin filled inside to lower the radioactivity concentration. At this time, because the stored water passes in one direction, clogging of the ion exchange resin occurs. If this phenomenon continues, the flow rate of the water treatment process decreases and operation efficiency decreases, so a backwashing process is necessary to re-mix the ion exchange resin and secure the flow rate again. In this study, the flow rate reduction trend according to the lifespan of the ion exchange resin and the flow rate recovery according to the backwash process operation amount were analyzed. The flow rate reduction trend of the ion exchange process was analyzed immediately after the backwashing process was started. In addition, the amount of flow recovery according to the backwash process operation amount was evaluated by the amount of waste generated during the backwash process and the number of days of operation until the backwash process was needed again. As a result, the flow rate of the ion exchange process decreased rapidly right after the backwash process until the position of the ion exchange resins was stabilized, and then stabilized. After that, it gradually decreased and reached the point where the backwash process was necessary. However, the decline trend was analyzed to be the same regardless of the lifespan of the ion exchange resin. In addition, the amount of waste generated during the operation of the backwash process was increased in the order of 400 L, 600 L, 1,100 L, 1,400 L, 3,500 L, and 4,200 L to increase the amount of operation of the backwash process. As a result, the number of days of ion exchange resin operation was 285 days, 338 days, and 342 days, was analyzed as 422 days, 322 days, and 720 days. Based on this study, it was confirmed that the flow rate reduction trend is the same regardless of the lifespan of the ion exchange resin, and as the backwash process operation increases, the number of days the ion exchange process can be operated increases, but there is a turning point where the waste treatment cost exceeds the number of days of operation.

Mixed-bed ion exchange resin consist of anion exchange resin and cation exchange resin is used to treat liquid radioactive waste in nuclear power plants. C-14 from heavy water reactors (HWR) is adsorbed on the anion exchange resin and is considered intermediate-level radioactive waste. The total amount of radioactivity of C-14 in spent ion exchange resin exceeds the activity limits for the disposal facility. Therefore, it is necessary to reduce the radioactivity through pre-treatment. There are thermal and non-thermal methods for the treatment of spent ion exchange resin. However, destructive methods have the problem of emitting off-gas containing radionuclides. To solve this challenge, various methods have been developed such as acid stripping, PLO process, activity stripping, thermal treatment and others. In this study, spent ion exchange resin (spent resin) was treated using microwave. The reaction characteristics of the resin to microwave were used to selectively remove the C-14 on the functional groups. Simulated spent anion exchange resin and spent resin from Wolseong NPP were treated with the microwave method, and the desorption rate was over 95%. An integrated process system of 1 kg/batch was built to produce operating data. After the operation of the process, characterization and evaluation of post-treatment for condensate water and adsorbent used in the process were performed. When the process system was applied to treat simulated spent resin and real spent resin, both showed a desorption rated of more than 97%. It means that the C-14 was successfully removed from the radioactive spent resin.

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.

Cardiovascular disease remains as one of the most common causes of high morbidity and mortality worldwide, despite remarkable medical advances in recent decades. Non-invasive techniques play a preeminent role in prevention of cardiovascular disease by diagnosing it at an early stage and guiding optimal patient management. Nuclear imaging is one of the most powerful means available for noninvasive diagnosis and management of poorly perfused myocardial region resulting from the cardiovascular disease. Several radionuclides are available for monitoring blood flow to cardiac tissue. The most validated radionuclides for these measurements are 13N, 15O, 99mTc, 201Tl and 82Rb. Each of 13N, 15O and 201Tl require the presence of an on-site cyclotron, whereas, 82Rb and 99mTc require only a generator. Rubidium (Rb) is an alkali metal ion that acts biologically like potassium and accumulates in cardiac muscle tissue. Rb has a rapid blood clearance profile which allows the use of 82Rb. It also has an ultra-short physical half-life of 75 sec for non-invasive evaluation of regional cardiac blood flow. There are several advantages of 82Rb over other radionuclides. Having a short half-life significantly reduces the radiation dose to the patient. In addition, 82Rb is a positron emitter, which gives the full advantages of PET such as image quantification with superior sensitivity. Several reports have shown superior diagnostic performances of 82Rb-PET over conventional 99mTc-SPECT. 82Rb can be produced from a generator system by the decay of its 25.6-day half-life parent 82Sr. However, the 82Sr parent is difficult to prepare. In routine generator production, certain purity is required to meet the specification of the product. Since there has been no the use of 82Rb radionuclide for research or medical purpose in Korea, we have plans to produce 82Sr with certain purity and develop a 82Sr/82Rb generator system. These studies can also be applied to remove radioactive Sr from radioactive waste waters. Because ion exchange resin, used for purification of 82Sr from impurities, is also utilized to trap radioactive Sr2+ ions from radioactive waste water. After Fukushima Daiichi nuclear accident, interest in the treatment of radioactive waste water has surged. As one of main fission products of nuclear reactor, 90Sr has been regarded as a hazardous radionuclide with half-life of about 29 years. Therefore, the investigation on ion exchange resin is important for removal of 90Sr from radioactive waste water. Here, we optimized 82Sr purification method using ion exchange resin to establish the most suitable procedure.

알칼리 금속 이온과 염소 이온이 포함된 용액으로부터 이온교환수지를 이용한 이온 제거에 대한 연구를 진행하였다. 양이온인 금속이온(Na+와 K+)의 제거에는 양이온교환수지를, 음이온인 염소 이온(Cl-)의 제거에는 음이온교환수지를 사용하였다. 용액 A (Na+를 36,633 ppm, Cl-를 57,921 ppm 함유)의 경우, Na+ 이온과 Cl- 이온은 20분 이내에 99% 이상 제거되었다. 용액 B (K+를 1,638 ppm 함유)의 경우, K+ 이온은 3분 이내에 99% 이상 제거되었다.

CKD 추출액은 시멘트공정에서 발생한 폐기물인 CKD를 시멘트 원료로 재사용하기 위해 공정 방해물질로 작용하는 KCl을 추출한 폐수이며, 폐수처리시설 증설 등의 문제로 추출액 무방류 및 이를 재이용하고자 하였다. 이온교환법을 적용하여 KCl을 제거한 결과, 이온교환 후 추출액의 pH는 12.7 에서 pH 2 미만으로 감소하였으며 양이온교환수지의 H+가 이온교환을 거쳐 추출액에 용해되었음을 확인하였다. 이온교환의 선택성에 의해 Ca2+, K+ 순서로 제거되었으며, K+ 이온을 제거하기 위해 접촉시 간의 증가가 필요함을 판단하였다. 이온교환수지와 직접접촉시간이 약 6배 높은 접촉시간을 갖는 회분 식장치에서 연속흐름식장치 대비 4배 높은 K+ 제거 효율을, 7배 높은 Cl- 제거 효율을 확인하였다. 양이온교환수지의 H+가 음이온교환수지의 OH- 대비 1.2배 빠른 교환속도를 가짐을 추출액 pH 변화를 통해 확인하였다.

최근, 전세계적으로 물 부족 현상과 지역개발 및 산업 고도화, 인구증가와 함께 물 수요는 증가하고 있다. 이를 해결하기 위한 방법으로 해수담수화 방법이 있다. 해수 담수화의 많은 방법 중 이온 교환막을 이용한 실험을 진행하였다. 본 연구에서는 Anion exchange resin을 대체할 수 있는 물질로 Polystyrene Latex입자를 제조 하였다. 제조된 입자에 chloromethylation과 amination을 통해 –NH3+, -NR3+, -PR3+, -SR2+등의 관능기를 도입하였으며, 제조된 입자와 고분자를 합하여 하이브리드 막 제조를 하였다. 특성평가로는 SEM, TGA, DSC, FT-IR, IEC Value를 통한 측정을 진행하였다.

본 연구에서는 수계 내 포함된 양이온들 중 특히 중금속 이온을 효율적으로 분리할 수 있는 양이온 교환막을 개 발하였다. 기저 고분자로는 sulfonated polyetheretherketone (SPEEK)를 사용하였으며 이에 중금속 이온에 결합력이 강한 킬 레이팅 수지를 파우더링하여 첨가하였다. 또한 양이온 교환막의 성능을 최적화시키기 위해 킬레이팅 수지의 함량 및 SPEEK 의 이온교환용량을 제어하였다. 결과적으로 제조된 양이온 교환막을 막 축전식 탈염 공정(membrane capacitive deionization, MCDI)에 적용한 결과 중금속 이온 제거 효율이 20% 이상 향상됨을 확인할 수 있었다.

Crosslinked ion exchange resin composite membranes were prepared by casting sulfonated polystyrene(SPS) solution with suspended ion exchange resin(crosslinked SPS) and crosslinker (trimethylolpropane ethoxylate triacrylate (TMPETA)) follow by gamma-ray irradiation. The physicochemical properties of the composite membranes were evaluated by measuring gel-fraction, ion exchange capacity, water-uptake and dimensional stability. We confirmed that the introduction of ion exchange resin and radiation crosslinking in the membranes improved the water uptake, dimensional stability and permselectivity.

양이온교환수지인 IRN-77을 직접 분해 처리하기 위하여 Fonton시약을 적용하였으며, 분해반응의 특징으로써 반응의 효율 및 안전을 위해 수지를 먼저 건조시키고 용액을 수지에 완전히 흡수시킨 후 를 첨가하는 방법을 적용하였다. 수지 분해반응의 특성은 반응이 개시되기까지 반응유도시간이 필요하였으며, 반응유도시간은 의 농도가 낮을수록 또한 의 초기 첨가량이 적을수록 길었다. 단위량의 수지를 분해하는데 적절한 반응조건으로서 의 농도는 0.9 M 및 15% 의 용액의 첨가량은 수지량에 대해 6-7배 비율로 나타났으며, 반응유도시간을 포함하여 1.5시간 이내에 완전 분해가 가능하였다. 의 첨가방법은 반응 초기 및 반응개시 후로 나누어 첨가하므로서 의 분해효율 및 첨가량을 최소화하였다. 가열효과로서 분해반응 개시 전에 비교적 낮은 온도인 정도로 가열하면 반응유도시간이 5분 이내로 단축되었으며, 수지의 양을 5g 및 10g 으로 증가시킨 결과, 의 첨가비율을 9-10배 정도로 증가시키면 완전분해가 가능하였다.