It has been studied on the disposal area reduction for the used nuclear fuel by the management of high decay-heat nuclides, long-lived nuclides, and highly mobile nuclides. It was investigated on the management of the nuclides in KAERI. Strontium-90 is a high heat-generating nuclide in spent nuclear fuel. It is needed to separate the salt from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. Vacuum distillation was used for the separation of strontium from the molten salt. Potassium carbonate was chosen as a reactive distillation reagent for SrCl2 – LiCl – KCl system by the thermodynamic calculation. Reactive distillation experiments were carried out. The residual was mainly SrCO3 in the XRD analysis. It could be concluded that K2CO3 could be one of the suitable reagents for the reactive distillation. The salt in the long–lived nuclide powders should be removed to prepare the block for disposal. Experiments were carried out using W powders (surrogate) and U3O8 powders to develop a process for the removal of the residual salt from UOx powders. The salts were successfully removed from the W and U3O8 powders by distillation.

Separation of high heat generating-radioactive isotopes from spent nuclear fuel is an important issue because it can reduce the final disposal area. As one of the technologies that can selectively separate only high heat generating-radioactive isotopes without dissolving spent fuel, the methods using molten salt have recently attracted attention. Although studies on chemical changes of Sr oxides in molten salts have been reported, they have limitation in that alternative oxide reagents rather than oxide fuel were used. In this study, the separation behaviors of Sr from simulated oxide fuel using various molten salts were investigated. A powder type containing 95.7wt% of U and 0.123wt% of Sr was used as the simulated oxide fuel. LiCl, LiCl-CaCl2, MgCl2, LiCl-KCl-MgCl2 and NaCl-MgCl2 were used as molten chloride salts. The separation of Sr from the simulated oxide fuel was conducted by loading it in porous alumina basket and immersing it in a salt. The concentration of Sr in the salt was measured by ICP analysis after sampling the salt outside the basket using dip-stick technique. The separation efficiencies of Sr from simulated oxide fuel using the salts were compared. Furthermore, the causes of their separation efficiency were systematically investigated.

PURPOSES : Presently, it is impossible to evaluate the actual number of operational plans and punctuality of bus routes. In this study, a methodology for separating bus stops and sorting departure orders was developed using city bus operation information, and its field applicability was examined. METHODS : Based on current bus data, the "moving speed (km/h)," "Lag data," and "Lead data" were generated. The number of stops was classified based on the location where the order of the stops changed for each vehicle, and an arbitrary ID was assigned for each number of stops. In the case of a preceding vehicle, for each rotation information, a time interval was calculated based on the same point, and the rotation with the smallest time interval was determined as the preceding vehicle. RESULTS : An evaluation was conducted on Bus Route 201 in Daejeon. The vehicles were classified based on the operation data of the previous day. From analyzing the operation time and distance for each operation based on the last stop and starting point, it was found that both the number of operation plans and performance were consistent. The punctuality was analyzed and was found to be inversely proportional to the difference between the stop arrival time and preset arrival time, preset dispatch interval, and actual dispatch interval. The on-time(regularity) ratio was determined as 98.0% and the early departure ratio was 99.3%, indicating very good regularity. CONCLUSIONS : The research results can be used by local governments with insufficient levels of public transportation-based data management, and can be used as basic data for establishing public transportation policies. The results of this study can be referred to when evaluating the transparency of financial support for public transportation operators and operations, with the aim of enhancing public transportation services.

This research was conducted for dewatered sludge cake of industrial wastewater treatment, i.e., as the object of inorganic sludge discharged especially in iron & steel manufacturing shop which used Air drying system to reduce water content. That drying system's single-type cyclone separator was confirmed to have significantly lower separation efficiency on the conditions 20μm and below of particular size through computational fluid dynamics(CFD) analysis. However, we found out the primarily advanced value of separation efficiency on dual-type directly connected. Regarding separation efficiency on size of 10μm, the efficiency of a single-type was presented at 51.91%. On the other side, the efficiency of the dual-type was 97.88%. This advanced effect of the dual cyclone separator was checked at a demo facility of air drying equipment designed by 340m3/min of airflow on site.

대용량 화학 및 청정에너지의 운반체인 수소는 석유화학 산업 및 연료전지 등에서 많이 활용되는 중요한 산업용 기체이다. 특히 수소는 주로 증기개질 및 가스화를 통해 화석 연료에서 생성되며 부산물로 이산화탄소가 발생한다. 따라서 고 순도 수소를 얻기 위해서는 이산화탄소를 제거해야 한다. 본 총설에서는 배러 단위[1 Barrer = 10−10 cm3 (STP) × cm / (cm2 × s × cmHg)]로 보고된 이산화탄소로부터 수소를 분리하는 프리스탠딩 고분자 분리막 및 혼합매질 분리막에 초점을 맞추었 다. 최근 보고된 다양한 논문들을 분석하여 분리막의 구조, 형태, 상호 작용 및 제조 방법에 대해 논의하고 구조-물성 관계를 이해하여 향후 더 나은 분리막 소재를 찾는 데 도움이 되고자 한다. 다양한 분리막의 성능 및 특성 검토를 통해 수소/이산화 탄소 분리에 대한 Robeson 성능 한계선을 제시하고, 가교, 혼합 및 열처리 등의 기술을 사용하여 분리 특성을 개선하는 다양 한 혼합매질 분리막에 대해 논의하였다.

이온교환막(IEM)은 다양한 종류의 단가이온과 다가이온을 분리하기 위해 사용되는 막의 한 종류로, 배터리, 연료 전지, 염화물-알칼리 공정 등에 사용된다. 이온교환막을 통한 막분리는 전기 구동력을 기반으로 한 녹색 분리 방식이며, 해수 담수화와 수처리 분야에서 떠오르는 방식이다. 전기투석(ED)은 양이온과 음이온이 이온교환막을 따라 선택적으로 이동하는 기술이다. 음이온 교환막(AEM)은 전기투석의 중요한 구성 요소 중 하나이며, 공정 효율을 향상시키는 데 상당한 역할을 한 다. 이온교환막에 가교결합을 도입하면 자유 부피의 감소로 인해 이온 선택 분리 성능이 향상된다. 역삼투(RO) 공정을 통한 해수 담수화 시 RO 농축수에 용해된 염이 다량 존재한다. 따라서 1가 양이온 선택막으로 구성된 전기투석 공정은 오염을 줄 이고 막 플럭스를 개선한다. 이 검토는 전기투석, 음이온 교환막, 그리고 양이온 교환막의 세 부분으로 나뉜다.

The purpose of this study is to use the hybrid steam-solvent process, because it is created in the form of water, bitumen, and water/bitumen emulsion by hot steam, so effective separation is required. Methods for separating the emulsion include a chemical separation method by adding a chemical, a separation method using an electrostatic property, a separation method using a membrane, a separation method using a microwave, and the like. Among them, the most used method is the separation method using a chemical, and it is reported that the separation efficiency of the emulsion is the best. In this study, a method for efficiently separating bitumen emulsions using a chemical separation method adding an emulsifier was investigated. In particular, technological trends in oil sand oil treatment technology were analyzed based on patent analysis.

According to Article 4 and 5 of the Nuclear Safety and Security Commission (NSSC) Notice No. 2020-6, radioactive waste packages should be classified by radioactive levels, and finally permanently shipped to underground or surface disposal facilities. The level of the radioactive waste package is determined based on the concentrations of the radionuclides suggested in Article 8 of NSSC Notice No. 2021-26. Since most of the radionuclides in radioactive wastes are beta nuclides, chemical separation and quantification of the target nuclides are essential. Conventional methods to classify chemically non-volatile radionuclides such as Tc-99, Sr-90, Nb- 94, Fe-55 take a lot of time (about 5 days) and have low efficiency. An automated non-volatile nuclide analysis system based on the continuous chemical separation method of radionuclides has been developed to compensate for this disadvantages of the conventional method in this study. The features of the automated non-volatile nuclide separation system are as follows. First, the amount of secondary waste generated during the chemical separation process is very small. That is, by adopting an open-bed resin column method instead of a closed-bed resin column method, additional fittings and connector are unnecessary during the chemical separation. In addition, because the peristaltic pump is supplied for the sample and solution respectively, it is great effective to prevent cross-contamination between radioactive samples and the acid stock solution for analysis. Second, the factors that may affect results, such as solution amount, operating time and flow rate, are almost constant. By mechanically controlling the flow rate precisely, the operating time and additional factors required during the separation process can be adjusted and predicted in advance, and the uncertainty of the chemical separation process can be significantly reduced. Finally, it is highly usable not only in the continuous separation process but also in the individual separation process. It can be applied to the individual separation process because the user can set the individual sequence using the program. As a result of the performance evaluation of the automation system, recovery rates of about 80–90% and reproducibility within 5% were secured for all of the radionuclides. Furthermore, it was confirmed that the actual work time was reduced by more than 50% compared to the previous manual method. (It was confirmed that the operation time required during the separation process was reduced from 6 days to 3 days.) Based on these results, the automation system is expected to improve the safety of workers in radiation exposure, reduce human error, and improve data reliability.

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).

Concrete is one of the largest wastes, by volume, generated during the decommissioning of nuclear facilities, which significantly influences the projected costs for the disposal of decommissioning wastes. Concrete consists of aggregates and a cement binder. In radioactive concrete, the radioisotopes are mainly associated with the cement component. If the radioactive isotope can be separated from the concrete to below the clearance criteria, the volume of radioactive concrete waste could be reduced effectively. We were studied to separate the radioactive materials from the concrete by using the thermomechanical and chemical treatment processes, sequentially. From the study, separated aggregate could be treated to achieve the clearance level. However, these processes generate a large volume of secondary acidic radioactive wastewater, which might be a critical problem to reduce the volume of radioactive concrete waste. In this research, separating the 137Cs and 90Sr from dissolved concrete wastewater to below the discharge criteria by precipitation method, it would be released to the environment under industrial waste guidelines. The experiments were conducted to using a simulated radioactive wastewater, formed by the dissolution of concrete within HCl, which was spiking the 137Cs and 90Sr, respectively. In addition, we applied the chemical precipitation methods with wastewater, using ferrocyanide for 137Cs and BaSO4 coprecipitation for 90Sr. As a result, targeted radionuclides could be removed to the discharge level (137Cs: 0.05 Bq·ml−1, 90Sr: 0.02 Bq·ml−1) by precipitation method. Therefore, it could reduce the secondary wastewater effectively by precipitation method and enhance the additional volume reduction for radioactive concrete waste.

Water electrolysis is a representative technology for tritium enrichment in water. Proton exchange membrane (PEM) water electrolysis has received great attention to replace traditional alkaline water electrolysis which generates concentrated tritiated water containing a large amount of salts. Nafion has been widely used as a polymeric electrolyte for the PEM electrolyzer. However, its low gas barrier property causes explosion, corrosion or degradation of electrolyzer. Furthermore, the traditional polymeric electrolytes have negligible differences in conductivity between hydrogen isotopes. To enhance the tritium separation by water electrolysis, we designed a composite membrane (Nafion/ hexagonal boron nitride (hBN)). The monolayer hBN has a high proton conductivity and gas barrier property, and the hBN can enhance conductivity differences between hydrogen isotopes. We prepared Nafion/hBN composite membranes, and water electrolysis performances and proton/deuterium separation behaviors were investigated.

Bentonite is considered as buffer of engineered barrier for retardation of radionuclide migration. Bentonite has low permeability, high swelling and high sorption capacity for radioactive nuclides. Properties have been widely investigated under various geochemical conditions simulating deep geological environments. The chemical stability of bentonite is an important factor in evaluating the long-term stability of the bentonite buffer. However, the presence of impurities in bentonite clays can reduce the retention capacity for retardation of radionuclide migration value of bentonite. Therefore, the bentonite purification is necessary. In the present study, grade improvement of montmorillonite was conducted using ultrasonic and froth flotation methods. As a result of confirming the grade of montmorillonite according to the optimal ultrasonic intensity for ultrasonic irradiation is 1.0 kHz of bentonite in Gyeongju (KJ-II) increased from 60% to 78%. In case of froth flotation method using PSS (0.1 mM) as a reagent, the grade of montmorillonite increased up to 90%.

A deep geological disposal system, which consists of the engineered and natural barrier components, is the most proven and widely adopted concept for a permanent disposal of the high level radioactive waste (HLW) thus far. The clay-based engineered barrier is designed to not only absorb mechanical stress caused by the geological activities, but also prevent inflow of groundwater to canister and outflow of radionuclides by providing abundant sorption sites. The principal mineralogical constituent of the clay material is montmorillonite, which is a 2:1 phyllosilicate having two tetrahedral sheets of SiO2 sandwiching an octahedral sheet of Al2O3. The stacking of SiO2 and Al2O3 sheets form the layered structures, and ion-exchange and water uptake reactions occur in the interlayer space. In order to reliably assess the radionuclide retention capacity of engineered barrier under wide geochemical conditions relevant to the geological disposal environments, sorption mechanisms between montmorillonite and radionuclides should be explicitly investigated in advance. Thus far, sorption behavior of mineral adsorbents with radionuclides has been quantified by the sorption-desorption distribution coefficient (Kd), which is simply defined as the ratio of radionuclide concentration in the solid phase to that in the equilibrium solution; the Kd value is conditional, and there have been scientific efforts to develop geochemically robust bases for parameterizing the sorption phenomena more reliably. In this framework, application of thermodynamic sorption model (TSM), which is theoretically based on the concept of widely accepted equilibrium models for aquatic chemistry, offers the potential to improve confidence in demonstration of radionuclide sorption reactions on the mineral adsorbents. Specifically, it is generally regarded in the TSM that coordination of radionuclides on montmorillonite takes place at the surficial aluminol and silanol groups while their ion-exchange reactions occur in the interlayer space also. The effects of electrical charge on the surface reactions are additionally corrected in accordance with the numerous theories of electrochemical interface. The present work provides an overview of the current status of application of TSM for quantifying sorption behaviors of radionuclides on montmorillonite and experimental results for physical separation and characterization of Ca-montmorillonite from the newly adopted reference bentonite (Bentonil- WRK) by means of XRD, BET, FTIR, CEC measurement, and acid-base titration. The determined mineralogical and chemical properties of the montmorillonite obtained will be used as input parameters for further sorption studies of radionuclides with the Bentonil-WRK montmorillonite.

When disposing of spent nuclear fuel, there is a risk of exposure that could exceed the annual allowable dose due to human intrusion after the institutional control period. Therefore, it can be treated with the pyroprocess, but the decontamination factor is not sufficient, and an additional actinide recovery is required because molten waste salt-containing actinide is generated. In the case of reducing the element in the spent molten salt through an electrochemical method using a liquid Bi electrode, it is difficult to separate only the actinide element because the two-element groups are reduced together due to the large concentration difference between the actinide and the rare earth element. Therefore, a process of forming a Bi intermetallic compound using a liquid Bi electrode, which has higher element separation efficiency than a liquid Cd electrode, and physically separating the Bi intermetallic compound using the difference in density of the produced compound has been proposed. For this, it is necessary to understand the properties and density separation of the intermetallic compound to be produced, and experiments were planned and conducted for this purpose. Various metals were added to the molten Bi to form an intermetallic compound, and an analysis device such as SEM was used to determine the intermetallics distribution, composition, and internal structure. As the added metal, Ce is a representative element for lanthanide, and Hf with the most similar intermetallic density, decomposition temperature, and standard reduction potential to U, and U as a substitute element for actinide was adopted. As a result of SEM and EDS analysis, it was confirmed that the separation was made in Bi due to the density difference between the produced intermetallic compounds. A Ce-Bi intermetallic compound was observed in the upper part, Hf at a concentration smaller than the error range was detected, and a Hf-Bi intermetallic compound which containing high concentration of Ce was observed in the lower part. Separation of high-purity Ce seems to be possible in the upper part, and it seems difficult to separate high-purity Hf in the lower part. Therefore, to separate highpurity Hf, an additional process suitable for it seems to be necessary.

Molten salt immersion technique has been tested with several Sr oxides, SrZrO3, SrMoO4 and U2SrOy, and MgCl2 based molten salts for the Sr nuclide separation. Reaction time, temperature, and salt composition were varied to effectively separate Sr in chloride forms. ICP-OES, XRD, and SEM analysis were conducted for the conversion efficiency and structure and morphology analysis. It is confirmed that all experiments of SrZrO3 with MgCl2 at 800°C for reaction time 5, 10, 20 hours showed higher conversion efficiency than 99% and in LiCl-KCl-MgCl2 and NaCl-MgCl2 molten salts at 500°C or 600°C, conversion efficiency higher than 97% was obtained. SrMoO4 in MgCl2 immersion experiments for 10 hours showed higher conversion efficiency than 99% when the molar ratio of salt/oxide powder is 7. U2SrOy was also tested with MgCl2 molten salt at 800°C and higher efficiency than 99% and mainly MgUO4 were produced as a reaction product.

알코올의 탈수를 위하여 에너지 다소비 공정인 증류 공정을 투과증발 막 공정으로 대체하려는 연구가 많이 진행 되어져 왔다. 대표적인 투과증발 막인 PDMS 분리막에 대한 시간의존적 분리 거동은 분리 메카니즘의 이해에 매우 중요하다. 따라서 본 연구에서는 50 wt% 에탄올-물 혼합용액에 대하여 50°C에서 막면적 1194 cm2인 PDMS/PSF 분리막 모듈의 시간의 존적인 투과증발 분리 거동을 고찰하였다. 총 유속과 에탄올/물 분리계수는 투과증발 시간이 증가함에 따라서 초기에 증가하 다가 다시 감소하였다. 초기 분리성능의 증가는 건조한 PDMS 분리막에 에탄올이 용해되는데 시간이 걸리기 때문이며, 후기 분리성능의 감소는 주입 탱크의 에탄올 농도가 시간에 따라서 감소하기 때문에 나타나는 현상이었다. 따라서 본 연구로부터 PDMS 분리막을 통한 에탄올의 투과는 용해-확산 메카니즘에 의해 발생된다는 것이 재확인되었다.

정유소와 석유 공장에서 발생하는 폐수는 심각한 환경오염으로 이어진다. 기름이 있는 물을 정수 처리하는 데에 는 많은 방법이 존재하지만, 가장 효과적인 방법은 막을 이용한 기술이다. 물에서 기름기를 제거하는 데 사용되는 유기재료로 만들어진 고분자 막은 파울링이라는 고질적인 문제를 가지고 있다. 무기성 막은 수명이 길다는 점에서 유기성 분리막보다 효 율적이다. 제올라이트 막은 우수한 화학적 안정성을 갖고 있으며 오랜 기간 재활용할 수 있다. 막에서 친수성의 존재는 막의 수 투과량을 증가시킨다. 제올라이트로 만들어진 세라믹 분리막은 물과 기름을 분리하는 데 사용되는 효율적인 무기막 중 하 나이다. 본 리뷰논문은 i) 순수 제올라이트 막과 ii) 다른 물질과 혼합된 제올라이트 복합막, 2가지로 분류되는 제올라이트 기 반의 무기막을 사용하는 물과 기름 분리 기술을 중점으로 다루고 있다.