To achieve permanent disposal of radioactive waste drums, the radionuclides analysis process is essential. A variety of waste types are generated through the operation of nuclear facilities, with dry active waste (DAW) being the most abundant. To perform radionuclides analysis, sample pretreatment technology is required to transform solid samples into solutions. In this study, we developed a dry ashing-microwave digestion method and secured the reliability of the analysis results through a validity evaluation. Additionally, we conducted a comparative analysis of the radioactivity of 94Nb nuclides with and without the chemical separation process, which reduced the minimum detectable activity (MDA) level by more than 65-fold for a certain sample.

The radioactive cesium, released from the normal operation or the accidental operation of nuclear facilities, should be regularly monitored for environmental regulatory compliance. The 135Cs/137Cs isotopic ratios, potentially useful for long-term tracking Cs transport in seawater, can be used as a tool of understanding how radionuclides are transported from different nuclear production source terms and distributed in the ocean. The ultra-high sensitive mass spectrometers (TIMS, SF-ICP-MS and TQ-ICP-MS) have been used to measure the 135Cs/137Cs isotopic ratios. However, the radiochemical separation of Cs from the seawater matrix is essential for the analysis of Cs using the mass spectrometers. An automated radiochemical procedure for the separation of Cs in seawater was developed for the analysis of 135Cs/137Cs isotopic ratios using a sequential column chromatography with AMPPAN and AG50Wx8 cation exchange resins. National Instrument’s LabVIEW is a graphical programming language and a powerful tool for the instrument control. A virtual instrument system for the automated separation of cesium isotopes was developed by the state machine of the fundamental design patterns in LabVIEW. In this study, the conceptual designs of an automated separation system of cesium isotopes, its virtual instrument system based on the LabVIEW state machine architectures and an automated radiochemical procedure were described for the purification of cesium isotopes at trace levels found in seawater discharged from the various nuclear facilities.

The physicochemical similarities of hydrogen isotopes have made their separation a challenging task. Conventional methods such as cryogenic distillation, Girdler sulfide process, chromatography, and thermal cycling absorption have low separation factors and are energy-intensive. To overcome these limitations, research has focused on kinetic quantum sieving (KQS) and chemical affinity quantum sieving (CAQS) effects for selective separation of hydrogen isotopes. Porous materials such as metal-organic frameworks (MOF), covalent organic frameworks (COF), zeolites, carbon, and organic cages have been studied for hydrogen separation. In this study, we focus the enhancement for CAQS to provide the cations due to the chemical affinity between hydrogen isotope and unsaturated sites by cations in zeolite beads. Cation exchanged zeolite beads was synthesized with cobalt, copper, nickel, iron and silver in zeolite 4A beads. Synthesized cation exchanged zeolite was analyzed for the surface area and pore size in N2 and adsorption behaviors of hydrogen isotopes (D2/H2) for various cation exchanged zeolite beads using BET at 77 K. The study predicts the D2/H2 adsorption selectivity based on the results obtained with BET. These hydrogen isotope adsorption results will provide a foundation for future processes for tritium separation.

Tritium is radioactive isotope, emitting beta ray, released as tritiated water from nuclear power plants. Due to the danger of radioactive isotope, the appropriate separation of tritium is essentially carried out for environment and safety. Further, it is also promising material for energy production and research. The tritiated water can be treated by diverse techniques such as water distillation, cryogenic distillation, Girdler-sulfide process, and catalytic exchange. After treatment, it is more desirable to convert as gas phase for storage, comparing to liquid phase. However, achieving complete separation of hydrogen gases with very similar physical and chemical properties is significantly challenging. Thus, it is necessary to develop materials with effective separation properties in gas separation. In this presentation, we present hydrogen isotope separation in the gas phase using modified mesoporous silica. Mesoporous silica is a form of silica that is characterized by its mesoporous structure possessing pores that range from 2 to 50 nm in diameter. This material can be functionalized to selectively capture and separate molecules having specific size and affinity. Here, the silver and copper incorporated mesoporous silica was synthesized to tailor a chemical affinity quantum sieving effect, thereby providing separation efficiency in D2/H2. The adsorption quantities of H2 and D2 were determined by sorption study, and the textural properties of each mesoporous silica were analyzed using N2 physisorption. The selectivity (D2/H2) in diverse feed composition (1:1, 1:9, and 1:99 of D2/H2) was estimated by applying ideal adsorbed solution theory to predict the loading of the gas mixture on bare, Ag- and Cu-mesoporous silica based on their sorption study. Further, the performance of each mesoporous silica was evaluated in the breakthrough adsorption under 1:1 mixture of D2 and H2 at 77 K.

Pt/C catalysts were prepared using black carbon (CB), and evaluated for their potential application as a catalyst of liquid-phase catalystic exchange for tritium treatment. CB was treated with 10% H2O2 solution for 0 and 2 hours at 105°C, Ethylene glycol and 40wt% Pt were added to the dried treated sample to prepare a Pt/C catalyst. The physical and chemical properties of the prepared catalysts were evaluated by BET, XRD, elemental analysis (EA), and TEM analyses. As a result of BET analysis, the surface area of CB without 10% H2O2 was 237.2 m2·g-1, and after treatment with 10% H2O2, it decreased to 181.2 m2·g-1 for 2 hours. However, the internal surface area increased, indicating the possibility that more Pt could be distributed inside the CB treated with 10% H2O2. In the XRD analysis results, the presence of Pt was confirmed by observing the Pt peak in the prepared Pt/C catalyst, and it was also observed through TEM analysis that Pt was evenly distributed within the CB. The elemental analysis (EA) results showed that the ratio of S and N decreased and the ratio of O increased with increasing 10% H2O2 treatment time. The H2O2 treated carbon supported Pt catalysts and polytetrafluoroethylene were then loaded together on a foamed nickel carrier to obtain hydrophobic catalysts. Our hydrophobic Pt catalyst using H2O2 treated black carbon are expected to be usefully used in the tritium treatment system.

Hydrogen isotope separation involves the separation of hydrogen, deuterium, tritium, and their isotopologues. It is an essential technology for removing radioactive tritium contamination and for obtaining valuable hydrogen isotope resources. Among various hydrogen isotope separation technologies, water electrolysis technology exhibits a high separation factor. Consequently, the electrolysis of tritiated water is of paramount importance as a tritium enrichment method for treating tritium-contaminated water and for analyzing tritium in environmental samples. More recently, hydroelectrolysis technology, which utilizes proton exchange membranes (PEM) to reduce water inventory, has gained favor over traditional alkaline hydroelectrolysis. Nevertheless, it is crucial to decrease the hydrogen permeability of the PEM in order to mitigate the explosion risk associated with tritium hydrogen electrolysis devices. Additionally, efforts are needed to enhance the hydrogen isotope selectivity of the PEM and optimize the manufacturing process of the membrane-electrode assembly (MEA), thereby improving both hydrogen isotope separation performance and water electrolysis efficiency. In this presentation, we will delve into two key aspects. Firstly, we’ll explore the reduction of hydrogen permeability and the enhancement of the hydrogen isotope separation factor in PEM through the incorporation of 2D nanomaterial additives. Secondly, we’ll examine the influence of various MEAs preparation methods on electrolysis and isotope separation performances. Lastly, we will discuss the effectiveness of the developed system in separating deuterium and tritium.

Domestic nuclear power plants can affect the environment if multiple devices are operated on one site and even a trace amount of pollutants that may affect the environment after power generation are simultaneously discharged. Therefore, not only radioactive substances but also ionic substances such as boron should be discharged as minimally as possible. We adopted pilot CDI and SD-ELIX sytem to separating and concenrating of boron containing nulcear power plant discharge water. The boron concentration of the initial inflow water tended to decrease over time. The water quality of concentrated water also reached its peak until the initial 60 minutes, but tended to decrease in line with the decrease in the inflow water concentration. The boron removal rate was in the range of 85 to 99% with respect to the initial boron concentration of 15 to 25 mg/L. On the other hand, performance degradation due to the use of electrochemical modules is also observed, and regeneration through low ion-containing water cleaning effective. We shortened processing time by considering the optimal flow rate conditions and conductivity conditions and converting electrochemical modules into series or parallel.

Nuclear power plants in Korea stores approximately 3,800 drums of paraffin solidification products. Due to the lack of homogeneity, these solidification products are not allowed to be disposed of. There is therefore a need for the separation of paraffin from the solidification products. This work developed an equipment for a selective separation of paraffin from the solidification product using the vacuum evaporation and condensational recovery method in a closed system. The equipment mainly consists of a vacuum evaporator and a condensational deposition recovery chamber. Nonisothermal vacuum TGAs, kinetic analyses and kinetic predictions were conducted to set appropriate operation conditions. Its basic operability under the established conditions was first confirmed using pure paraffin solid. Simulated paraffin solidification product fixing dried boric acid waste including nonradioactive Co and Cs were then fabricated and tested for the capability of selective separation of paraffin from the simulated waste. Paraffin was selectively separated without entertainment of Co and Cs. It was confirmed that the developed equipment could separate and recover paraffin in the form of nonradioactive waste.

The separation efficiency of nuclides in molten salt systems was investigated, with a focus on the influence of apparatus configuration and experimental conditions. A prior study revealed that achieving effective Sr separation from simulated oxide fuel required up to 96 hours, reaching a separation efficiency of approximately 90% using a static dissolution reaction in a porous alumina basket. In this study, we explored the impact of agitation on improving Sr separation efficiency and dissolution rates. The simulated oxide fuel composition consisted of 2wt% Sr, 3wt% Ba, 2wt% Ce, 3wt% Nd, 3wt% Zr, 2wt% Mo, and 89wt% U. To quantify the Sr concentration in the salt, we utilized ICP analysis after salt sampling via a dip-stick technique. Furthermore, we conducted ICPOES analysis over a 55-hour duration to assess the separated nuclides. Complementing these analyses, SEM and XRD investigations were performed to validate the crystal structure and morphology of the oxide products.

가스 분리 방법 중에서도, 멤브레인을 이용한 CO2 포집 및 분리는 지속적으로 개발되고 있는 꾸준히 성장하는 분 야이다. 이온성 액체(IL) 기반 복합 막은 CO2를 분리하는 데 있어 우수한 성능값을 보여준다. 유사하게, 다양한 공중합체/IL 복합막 또한 향상된 성능을 보여준다. 이러한 공중합체/IL 복합만에 산화그래핀과 같은 필러를 첨가하면 IL과 유기 필러 사 이에서 발생하는 강한 상호작용으로 인해 필러의 효과가 더욱 향상되며, 이는 결과적으로 CO2의 친화도, 선택도 및 흡착과 같은 요소를 향상시킨다. 금속-유기 구조체(MOF)를 사용하는 공중합체/IL 복합 막은 향상된 CO2 투과도를 보여주었다. 이 총설에서는 이온성 액체와 공중합체복합막의 다양한 조합에 따른 이산화탄소분리성능에 대한 상관관계를 논의한다.

본 연구에서는 중공사형 지지체막을 폴리술폰(polysulfone, PSf) 고분자를 이용하여 비용매 상분리법(non solvent induced phase separation, NIPS)에 의해 제조하였다. 제조된 중공사 지지체막을 PDMS와 Pebax를 코팅하여 중공사형 복합막 을 제조하고 CO2, H2, O2 그리고 N2에 대한 순수 투과도(permeance)와 선택도를 측정하였다. 제조된 복합막 모듈 중에서 선 택도(CO2/H2)가 가장 높은 모듈을 선정하여 모사가스를 사용하여 스테이지컷(stage cut, SC)의 변화에 따라 분리성능을 측정 하였다. 이때 사용된 모사가스는 PSA에서 버려지는 off gas의 농도인 CO2 70% : H2 30%인 것을 사용하였다. 1단 실험에서 는 H2 농도 약 60%, H2 회수율 12%의 값을 얻을 수 있었다. 낮은 H2 농도와 회수율을 극복하기 위해 2단 직렬 테스트를 수 행하였으며, 이를 통해 H2 농도 약 70%, H2 회수율 70%를 달성할 수 있었으며, 이를 통해 CO2/H2 분리에 대하여 분리 공정 구성을 도출할 수 있었다.

공유 유기 프레임워크(COF)는 분자 분리, 염료 분리, 가스 분리, 여과 및 담수화를 포함한 다양한 응용 분야에서 가능성을 보여주었습니다. COF를 막에 통합하면 투과성, 선택성 및 안정성이 향상되어 분리 공정이 향상됩니다. 단일 벽 탄 소 나노튜브(SWCNT)와 COF를 결합하면 염료 분리에 이상적인 높은 투과성과 안정성을 가진 나노 복합막을 생성합니다. COF를 폴리아미드(PA) 막에 통합하면 합성 계면 전략을 통해 투과성과 선택성이 향상됩니다. 혼합 매트릭스 막(MMM)의 3 차원 COF 필러는 CO2/CH4 분리를 향상시켜 바이오가스 업그레이드에 적합합니다. COF와 금속 유기 프레임워크(MOF) 막 을 결합한 모든 나노 다공성 복합재(ANC) 막은 투과성-선택성 트레이드오프를 극복하여 가스 투과성을 크게 향상시킵니다. 가상 COF (hypoCOF)를 사용한 계산 시뮬레이션은 CO2 분리 및 H2 정제와 관련하여 우수한 CO2 선택성과 작업 능력을 입 증합니다. 박막 복합재(TFC) 및 폴리술폰아미드(PSA) 막에 통합된 COF는 유기 오염물, 염 오염물 및 중금속 이온에 대한 거 부 성능을 향상시켜 분리 능력을 향상시킵니다. TpPa-SO3H/PAN 공유 유기 프레임워크 막(COFM)은 대전된 그룹을 활용하 여 정전기적 반발을 통해 효율적인 담수화를 가능하게 함으로써 기존의 폴리아미드 막에 비해 우수한 담수화 성능을 보여 이 온 및 분자 분리의 잠재력을 제시했습니다. 이러한 연구 결과는 투과성, 선택성 및 안정성을 향상시켜 향상된 분리 공정을 위 한 막 기술에서 COF의 잠재력을 강조합니다. 이 검토에서는 분리 공정에 적용된 COF에 대해 논의합니다.

진해만은 태풍 내습 시 피항 선박이 폭주하고 강한 바람 등의 영향으로 주묘가 자주 발생하며 이에 따른 선박 간 충돌 및 좌초 등 해양사고 발생 개연성이 매우 높다. 본 연구에서는 우리나라 해역 특성에 맞는 진해만 정박지의 선박 간 안전이격거리 설정을 위한 연 구를 수행하였다. 진해만 태풍 피항지에는 태풍 내습 시 평균 100 ~ 200여척의 선박이 정박을 하고 있으며 풍속이 25m/s 이상되는 강한 외력에서 전체 선박의 약 70%에 주묘가 발생하는 상황인 것으로 분석되었다. 본 연구에서는 국내외 설계기준 상 제시된 황천 시 정박 선 박간 이격거리, 실제 피항지로서 사용된 진해만 피항선박 간 이격거리, 강한 외력에 따른 선박 표류 시 적정 안전거리 등을 분석하여 제 시하였다. 그 결과 설계기준 상의 최소 기준과 비상조치 시간을 고려하여 약 400 ~ 900m의 안전이격거리가 필요하며, 공간상의 여유가 있 는 경우에는 700 ~ 900m 이격거리 설정이 필요한 것으로 도출하였다. 본 연구의 결과는 향후 진해만 피항지를 이용하는 선박에 대해 선박 간 안전 이격거리를 위한 지침 수립 시 활용이 가능할 것으로 기대된다.

1,3-다이옥솔란은 용매, 전해질 및 시약으로서 화학, 페인트 및 제약 산업에서 광범위한 관심을 받고 있는 화합물 이다. 1,3-dioxolane은 주로 독성, 발암성, 폭발성, 자동 인화성이 없으며 다기능성을 가지고 있어, 대부분의 유기 및 수성 용 매 조건에서 우수한 용해성을 가져 고분자 전구체로서 사용된다. 최근 몇 년 동안 이 물질은 배가스 및 천연 가스 혼합물에 서 CO2를 분리하기 위한 CO2 선택적 고분자 전구체로서 점점 더 많은 관심을 받고 있다. Poly(1,3-dioxolane) (PDXL)은 폴 리에틸렌 옥사이드(PEO)보다 높은 에테르 산소 함량을 가지고 있으며, 이는 극성 에테르 산소 그룹이 CO2에 대해 강한 친화 력을 나타내기 때문에 우수한 막 CO2/N2 분리 특성을 보인다. 따라서 PDXL 기반 분리막은 비극성(N2, H2 및 CH4) 가스에 대해 탁월한 CO2 용해도 선택성을 보인다. 그러나 PEO와 마찬가지로 PDXL의 극성기는 고분자 사슬의 밀집도를 증가시키고 고분자 결정화를 유발하여 기체 투과도를 감소시켜 이에 대한 개선이 필요하다. 이 논문에서는 기체 분리 응용 분야에서 PDXL기반 고분자막의 최근 발전과 한계에 대해 알아보고자 한다. 또한 CO2 분리 공정에서 1,3-dioxolane 기반 고분자의 한 계를 극복하기 위한 몇 가지 분자 설계방안에 대해 다루어 보기로 한다.

This study was performed to evaluate the separation of Sr, Cs, Ba, La, Ce, and Nd using gas pressurized extraction chromatography (GPEC) with anion exchange resin for the quantitation of Neodymium. GPEC is a micro-scaled column chromatography system that provides a constant flow rate by utilizing nitrogen gas. It is overcome the disadvantages of conventional column chromatography by reducing the volume of elution solvent and shortening the analysis time. Here, we compared the conventional column chromatography and the GPEC method. The whole analysis time was decreased by nine times and radioactive wastes were reduced by five times using the GPEC system. Anion exchange resin 1-X4 (200~400 mesh size) was used. The sample was prepared at a 0.8 M nitric acid in methanol solution. The elution solvent was used at a 0.01 M nitric acid in methanol solution. Finally the eluate was analyzed by ICP-MS to determine the identification and recovery. In this case, we applied the natural isotopes of LREEs (139La, 140Ce, and 144Nd) and high activity nuclides (88Sr, 133Cs, and 138Ba) instead of radioactive isotopes for the preliminary test; as a result, unnecessary radioactive waste was not produced. The recoveries were 93.9%, 105.9%, 91.9%, 47.6%, 35.9%, and 79.9% of Sr, Cs, Ba, La, Ce, and Nd, respectively. The reproducibility of recoveries by GPEC were in the range 2.8%–10.9%.

The separation of hydrogen isotopes is a critical issue in various fields, such as deuterium or tritium production and the treatment of radioactively contaminated water. In this presentation, we describe the pervaporative separation of hydrogen isotopes using proton conductive membranes and underlying separation mechanism. We investigated the H/D separation factors of perfluorosulfonic acid (Nafion) and polybenzimidazole membranes using pervaporation, and found that both membranes exhibited similar separation factors of approximately 1.026. Water permeation flux through the membranes was highly dependent on their thickness and type, and increased with operation temperature. However, the effect of temperature on H/D separation factor was negligible. We also demonstrated the cascade separation of H/D, indicating the potential application of multi-stage operation. We found that surface transport mechanisms such as hydron hopping contributed the most to H/D separation during the pervaporation process of proton conductive membranes.

The physicochemical similarities of hydrogen isotopes have made their separation a challenging task. Conventional methods such as cryogenic distillation, Girdler sulfide process, chromatography, and thermal cycling absorption have low separation factors and are energy-intensive. To overcome these limitations, research has focused on kinetic quantum sieving (KQS) and chemical affinity quantum sieving (CAQS) effects for selective separation of hydrogen isotopes. Porous materials such as metal-organic frameworks (MOF), covalent organic frameworks (COF), zeolites, carbon, and organic cages have been studied for hydrogen separation. This study have the literature review for previous research on D2/H2 adsorption and analyzes the D2/H2 adsorption behaviors of hydrogen isotopes for various zeolite using BET at 77 K. The study predicts the D2/H2 adsorption selectivity based on the results obtained with BET. These hydrogen isotope adsorption fundamentals provide a foundation for future processes for tritium separation.

Separating nuclides from spent nuclear fuel is crucial to reduce the final disposal area. The use of molten salt offers a potential method for nuclide separation without requiring electricity, similar to the oxide reduction process in pyroprocessing. In this study, a molten salt leaching technique was evaluated for its ability to separate nuclides from simulated oxide fuel in MgCl2 molten salts at 800°C. The simulated oxide fuel contained 2wt% Sr, 3wt% Ba, 2wt% Ce, 3wt% Nd, 3wt% Zr, 2wt% Mo, and 89wt% U. The separation of Sr from the simulated oxide fuel was achieved by loading it into a porous alumina basket and immersing it in the molten salt. The concentration of Sr in the salt was measured using ICP analysis after sampling the salt outside the basket with a dip-stick technique. The separated nuclides were analyzed with ICP-OES up to a duration of 156 hours. The results indicate that Ba and Sr can be successfully separated from the simulated fuel in MgCl2, while Ce, Nd, and U were not effectively separated.