최근 생활방식의 변화로 인하여 실내 생활이 점점 증가함에 따라 다양한 인테리어 자재의 수요가 증가하고 있으며, 이에 따라 인테리어 스톤 제품 생산 과정에서 발생하는 산업 폐기물인 슬러지의 발 생도 더불어 증가하고 있다. 발생하는 슬러지는 전량 소각 및 매립되어 처리되고 있으며 환경파괴 및 매립지 부족 등의 문제로 슬러지 처리에 어려움을 겪고 있는 실정이다. 이와 더불어 최근 건설 현장의 골재 수급은 매우 어려운 상황이며 이는 직접적으로 레미콘의 품질 및 가격에 영향을 미치게 된다. 이 러한 문제점의 해결을 위하여, 본 연구에서는 모르타르 내부의 잔골재를 인테리어 스톤 슬러지로 치환 하여 슬러지의 친환경적 재활용성을 검토하고자 하였다. 선행 연구를 바탕으로 시멘트, 슬러지, 잔골 재, 고유동화제 등을 활용하여 배합비를 설정하였으며, 이에 대한 시험체를 제작 하였다. 잔골재 무게 대비 슬러지는 각각 5, 10, 15, 20%를 치환하였으며, 각 배합에 대한 유동성과 재령별 압축강도를 측 정하였다. 관입저항 실험을 통해 각 시편의 초결과 종결 시간을 확인하였으며 수은압입법을 통해 시편 별 내부의 공극을 측정하였다.

Highly radioactive waste is solidified to restrict leaching, retain its shape, and maintain its structural stability to prevent it from affecting humans and the environment as much as possible. This operation should be performed consistently regardless of whether the waste is homogeneous or heterogeneous. However, currently, there are no specific performance requirements for heterogeneous waste in Korea. This study reviewed domestic research results and the status of overseas applications, and proposed immobilization requirements for heterogeneous waste to be applied in Korea. IAEA safety standards, domestic laws, and waste acceptance criteria were reviewed. The status of heterogeneous waste immobilization in countries such as the United States, France, and Spain was reviewed. Most countries treat heterogeneous waste by encasing it in concrete, and impose immobilization requirements on this concrete. Based on these data, safety standards for the thickness, compressive strength, and diffusion limit of this concrete material were proposed as immobilization requirements for heterogeneous waste disposal in Korea. Quantitative values for the above requirements need to be derived through quantitative assessments based on the characteristics of domestic heterogeneous waste and disposal facilities.

During the dismantling of nuclear facilities, a large quantity of radioactive concrete is generated and chelating agents are required for the decontamination process. However, disposing of environmentally persistent chelated wastes without eliminating the chelating agents might increase the rate of radionuclide migration. This paper reports a rapid and straightforward ion chromatography method for the quantification of citric acid (CA), a commonly used chelating agent. The findings demonstrate acceptable recovery yields, linearities, and reproducibilities of the simulated samples, confirming the validity of the proposed method. The selectivity of the proposed method was confirmed by effectively separating CA from gluconic acid, a common constituent in concretes. The limits of detection and quantification of the method were 0.679 and 2.059 mg·L−1, respectively, while the recovery yield, indicative of the consistency between theoretical and experimental concentrations, was 85%. The method was also employed for the quantification of CA in a real concrete sample. These results highlight the potential of this approach for CA detection in radioactive concrete waste, as well as in other types of nuclear wastes.

교통량이 증가하고 교량과 같은 특수구조물에 아스팔트 포장이 시공되는 사례가 증가함에 따라 일반적으로 사용되는 아스팔트보다 높은 성능을 가진 아스팔트에 대한 수요가 증가하고 있다. 일반 아스팔트 혼합물은 내구연한이 지나면 재생첨가제 등을 사용하여 다 시 도로포장재료로서 재활용할 수 있는 방안이 마련되어 있으나, 개질 아스팔트가 사용된 폐아스팔트 혼합물은 매립재로 사용하는 것 이외에는 별다른 대안이 없는 실정이다. 이에 본 연구에서는 국토부 지침에 규정된 재활용 아스팔트 혼합물 배합설계법을 적용하여 개질 폐아스팔트 혼합물을 재활용할 수 있는지를 검토해보고자 하였다. 이를 위해 개질 아스팔트를 활용하여 혼합물을 제작하였으며, 현장에서 수거되는 폐아스팔트 혼합물의 노화상태를 모사하기 위해 AASHTO R 30을 참고하여 강제 노화를 실시하였다. 노화 및 추출 과정에서 아스팔트의 물성 변화를 확인하기 위해 절대점도, DSR, MSCR 시험을 실시하였다. 시험결과, 추출 후 바인더의 절대점도는 감소하였으나 G*(복합전단계수)와 δ(위상각)은 증가하는 경향을 보였다. 소성변형 저항성을 확인하기 위해 MSCR(다중 응력 크리프 및 회복) 시험을 실시한 결과,  이 2배 가까이 증가하여 소성변형 저항성이 감소한 것을 확인할 수 있었다. 이러한 결과는 추출시 사용 되는 용매가 개질첨가제를 추출하지 못하여 기인한 결과로 판단된다. 따라서 개질 폐아스팔트 혼합물을 재활용하기 위해서는 기존과 는 다른 별도의 배합설계법이 개발되어야 할 것으로 판단되었다.

This study addresses the environmental impact associated with waste management and natural aggregate production. It explores the potential of utilizing Coal Bottom Ash (CBA) and Reclaimed Asphalt Pavement Aggregate (RAPA) as complete replacements, respectively, for fine and coarse aggregates in concrete. Despite their similarities to natural aggregates, CBA and RAPA often end up in landfills. Laboratory tests were conducted, revealing satisfactory performance in drying shrinkage and air void parameters. However, while the flexural strength met design requirements, the compressive and splitting tensile strengths were lower than predicted. The deviation in strength development behavior from natural aggregate concrete (NAC) was attributed to weak agglomerated aggregates in RAPA and the large size of the interfacial transition zone (ITZ) due to the old asphalt coating surrounding RAPA. To enhance the strength behavior, two methods were employed: compaction in the form of roller-compacted concrete and RAPA abrasion carried out by rolling RAPA in a concrete mixer. Compaction improved aggregate interlock, while RAPA abrasion decreased agglomerated aggregates and minimized asphalt coating, reducing ITZ size. These treatments resulted in improvements in compressive, flexural, and splitting tensile strengths, with the combination of both treatments having the most significant effect. Analysis of relationships between flexural, splitting tensile, and compressive strengths indicated that CBA and RAPA concrete behaved more similarly to NAC after the treatments. This research suggests that with appropriate interventions, it is feasible to utilize CBA and RAPA in concrete, contributing to sustainable construction through improved waste management, carbon footprint reduction, and conservation of natural resources.

The waste secondary battery contains a significant amount of valuable metals, making its recycling highly desirable. However, conventional chemical methods for recycling are environmentally unfriendly and cost-ineffective. Rather than the chemical method, this paper deals with a mechanical method for recovering electrode materials from waste secondary batteries by blowing pressurized air onto the interface area between the electrode and the separator. Especially, in this study, the effective blowing angle were searched by simulating the separation of the electrode material from the separator through 1-way fluid structure interaction analysis based on the Cohesive Zone Modeling technique.

This work involves the development of a novel waste-derived carbon dots (CDs) conjugated with silver (Ag) nanohybrid system-based Fluorescence Resonance Energy Transfer (FRET) sensor for the detection of melamine. CDs and Ag nanoparticles served as energy donors and energy acceptors, respectively. CDs were synthesized from orange peel waste through a combined hydrothermal and ultra-sonication route. The synthesized CDs had hydroxyl, amino, and carboxyl groups on their surface, explaining that waste-derived CDs can act as reducing and stabilizing agents and showed strong absorption and fluorescence emission at 305 and 460 nm, respectively. The bandgap, linear refractive index, conduction band, and valance band potential of CDs were observed to be 2.86, 1.849, 1.14, and 4.002 eV, respectively. No significant difference was observed in the fluorescence properties at different pH (acid and alkaline) and ionic concentrations. Given their fluorescent nature, the synthesized CDs were used for the detection of melamine. The fluorescence of CDs was found to be quenched by Ag+ due to the FRET energy transfer between CDs to Ag. Notably, the zeta potential of Ag@CDs was changed from − 28.7 mV to − 30.6 mV after the incorporation of Ag+. Ag@CDs showed excellent selectivity and sensitivity toward the sensing of melamine in the aqueous solutions with the limit of detection ~ 0.85 μM. Increasing the melamine level also raises the FL intensity of Ag@CDs. The substrate was effectively used in the detection of melamine in milk as a real application and the recovery percentage was found to be 98.03%. Moreover, other adulterants such as urea and formaldehyde can be detected selectively by Ag@CDs. Overall, the synthesized Ag@CDs can be used as an efficient material for sensing applications involving such food adulterants.

국내 중·저준위 방사성폐기물은 영구적 격리를 위해 처분장에 매립하고 있으며 그 위치는 경주에 있다. 이러한 방 사성폐기물의 영구적인 격리를 위한 처분시설은 공학적 방벽과 자연 방벽으로 구성되어 있으며 자연 방벽을 특성을 파악 하기 위하여 한국원자력환경공단에서는 2006년부터 부지특성조사를 수행하였고, 이후 부지감시 및 조사계획에 따른 감시 를 수행하여 부지특성의 변화를 지속적으로 확인하고 있다. 중저준위 방폐장의 수리지화학적 환경은 자연 방벽의 평가를 위해 중요한 요소로 손꼽히고 있으나 동해와 가까운 경주의 지역적 특성상 해수의 영향을 반드시 고려해야 한다. 따라서 본 연구에서는 처분 부지의 지하수 관정 7개 및 관정의 심도별 수질 자료를 취합해 지하수 자료 총 30개를 해수 2개소와 비교 분석하여 수리지화학적 환경을 해석하였다. 분석 자료는 수질 10개 항목(온도, EC, HCO3, Na, K, Ca, Mg, Cl, SO4, SiO2)을 2017년 3분기부터 2022년 3분기까지 총 5년간 20회의 자료를 활용하였다. 특히, EC, HCO3, Na, Cl의 농도 변화 를 통해 연구 지역의 배경 농도 및 관정의 구간별 해수의 영향을 파악하였으며, 시계열 군집 분석을 통해 담수, 기수, 해 수의 분류를 시도하였다. 그 결과, 기존의 모니터링 방법으로는 확인하지 못한 부지내 수리지화학적 변화를 제시하였다.

This study presents a rapid and sequential radiochemical separation method for Pu and Am isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin and TRU resin. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the sample was allowed to evaporate to dryness after filtering the leaching solution with 0.45 micron filter. The Pu isotopes were separated in HNO3 medium with anion exchange resin. For leaching solution passed through anion exchange column, the Am isotopes were separated with TRU resin. The purified Pu and Am isotopes were measured by alpha spectrometer, respectively, after micro-precipitation of neodymium. The sequential radiochemical separation of Pu and Am isotopes in radioactive waste samples using anion exchange resin and TRU resin was validated with ICP-MS system.

Chelating agents, such as EDTA, NTA, and citric acid, possess the capacity to establish complexes with radionuclides, potentially enhancing the migration of such radionuclides from the disposal sites. Hence, quantification of these chelating agents in radioactive wastes is required to ensure secure disposal protocols. The determination of chelating agents in radioactive wastes is mainly composed of two steps, e.g. extraction and detection. However, there are little information on the extraction of the chelators in various radioactive wastes. We endeavored to optimize the extraction conditions for citric acid (CA) found within concrete, a prevalent component in the context of dismantled waste materials. Given the inherent high solubility of CA in water, we applied an aliquot of deionized water to the concrete and conducted a one-hour ultrasonic leaching procedure to facilitate chelate extraction. Subsequently, following the preparation of the concrete leachate via vacuum filtration and centrifugation to yield a clarified solution, we quantified the concentration of CA within the solution using Ion Chromatography (IC). To enhance leaching efficiency, we examined the % recovery variation with respect to the pH of the leaching solution. The optimized extraction method will be applied to diverse chelating agents and radioactive waste categories.

For safe disposal of radioactive wastes, accurate analysis of nuclear isotopes is important. It is known that there are 14 nuclides that have to identify nuclide-specific concentration levels. 63Ni, one of non-volatile nuclear isotopes which is included in those 14 nuclides, has to follow chemical separation for exact analysis. As various analysis methods were developed, various methods for analyzing 63Ni also emerged. Past method has used measurement specimens of 59Ni, after 59Ni measurement has been done. It used HClO4, known as strong oxidizing agent, to dissolve DMG, an organic substance used to form 59Ni precipitates. Nowadays, we analyze 59Ni and 63Ni simultaneously, which enables short analysis time, without use of HClO4. But high accuracy is just as important as short measurement time and efficiency. So, this paper compare 63Ni specific activity value used new method with the value, past method used, using real sample’s data. As a result, all sample data from new method’s relative 63Ni specific activity is within the uncertainty range of past ones based on past specific activity value. Consistency of new method’s result and past method’s data increased the reliability of the data and accuracy of those methods.

In the decommissioning process of nuclear power plants, Ni-59, Ni-63 and Fe-55 present in radioactive waste are crucial radionuclides used as fundamental indicators in determining waste treatment methods. However, due to their low-energy emissions, the chemical separation of these two radionuclides is essential compared to others. Therefore, this study aims to evaluate the suitability of various pre-treatment methods for decommissioning waste materials by conducting characteristic assessments at each chemical separation stage. The goal is to find the most optimized pre-treatment method for the analysis of Ni-59, Ni-63 and Fe-55 in decommissioning waste. The comparative evaluation results confirm that the chemical separation procedures for Fe and Ni are very stable in terms of stepwise recovery rates and the removal of interfering radionuclides. However, decommissioning waste materials, which mainly consist of concrete, metals, etc., possess unique properties, and a significant portion may be low-radioactivity waste suitable for on-site disposal. Considering that the chemical behavior and reaction characteristics may vary at each chemical separation stage depending on the matrix properties of the materials, it is considered necessary to apply cascading chemical separation or develop and apply individual chemical separation methods. This should be done by verifying and validating their effectiveness on actual decommissioning waste materials.

Chelating agents in low and intermediate radioactive wastes can form complexes with radionuclides and increase the mobility of the radionuclides. According to the Korea Radioactive Waste Agency (Acceptance criteria for low and intermediate radioactive waste, WAC-SIL-2022-1), if the amount of residual chelating agents in the waste are greater than 0.1%, the chemical names and residual amounts should be specified; if greater than 1%, the waste must be solidified and contain no more than 8%. The existing method for analyzing chelates in radioactive waste was based on UV–Visible spectrophotometry (UV-Vis), but the new method is based on liquid chromatography/mass spectrometry (LC-MS). The analysis was performed in aqueous solution before applying to real samples. Since the real sample may contain several heavy metals, it is expected that the chelates will exist as complexes. Therefore, 1.0×10-4 mol L-1 of EDTA (Ethylenediaminetetraacetic acid), DTPA (Diethylenetriaminepentaacetic acid), NTA (Nitrilotriacetic acid), and excess metals in aqueous solution were analyzed using HPLC using RP (Reverse Phase) column and HILIC (Hydrophilic interaction) column. When the RP column was used, each substance eluted without separation at the beginning of the analysis. However, when analyzed using a HILIC column, the peaks of each substance were separated. LC-MS measurements using HILIC conditions resulted in separations with better sensitivity.

As existing nuclear power plants reach the end of their lifespan, 22 nuclear power plants in korea are scheduled to be permanently shut down and decommissioned by 2050. Chelates are used as decontamination agents during nuclear power plant operation and decommissioning, and as a result, decommissioning waste contains chelates. Chelates contained in radioactive waste are complexed with radionuclides and increases their mobility. So, qualitative and quantitative analysis of chelates contained in radioactive waste is necessary. However, the spectroscopic method (UVVis), previously used for chelate analysis in Korea takes too much time for analysis and cannot analyze two or more chemically similar chelates at the same time. Due to these problems, new methods for analyzing chelate must be developed. Overseas, many cases of chelate analysis using advanced analysis equipment have been reported. CEA in France has developed a chelate analysis method for application to radioactive waste using HPLC-MS (J. Chromatogram. A, 1276, 20-25, 2013). In this method, the existing method of measuring EDTA using a complex of Fe and EDTA was improved to measuring a complex of Ni and EDTA. Based on such overseas cases, we would like to develop an analysis method for chelates in radioactive waste. For this purpose, we will verify similar overseas papers and develop pretreatment methods for mixtures of chelates (EDTA, DTPA, NTA) and metals (Fe, Ni, Cu, etc.) in various media. Finally, we will develop a separation analysis technology for multi-component chelates in nuclear decommissioning waste based on LCMS.

Concrete is the primary building material for nuclear facilities, making it one of the most common forms of radioactive waste generated when decommissioning a nuclear facility. Of the total waste generated at the Connecticut Yankee and Maine Yankee nuclear power plants in the United States, concrete waste accounts for 83.5% of the total for Connecticut Yankee and 52% for Maine Yankee. In order to dispose of the low- to medium-level radioactive concrete waste generated during the decommissioning of nuclear power plants, it is necessary to analyze the radioactivity concentration of gamma nuclides such as Co-58, Co-60, Cs-137, and Ce-144. Gamma-ray spectroscopy is commonly used method to measure the radioactivity concentration of gamma nuclides in the radioactive waste; however, due to the nature of gamma detectors, gamma rays from sequentially decaying nuclides such as Co-60 or Y-88 are subject to True Coincidence Summing (TCS). TCS reduces the Full Energy Peak Efficiency (FEPE) of specific gamma ray and it can cause underestimation of radioactivity concentration. Therefor the TCS effect must be compensated for in order to accurately assess the radioactivity of the sample. In addition, samples with high density and large volume will experience a certain level of self-shielding effect of gamma rays, so this must also be compensated for. The Radioactive Waste Chemical Analysis Center at the Korea Atomic Energy Research Institute performs nuclide analysis for the final disposal of low- and intermediate-level concrete waste. Since a large number of samples must be analyzed within the facility, the analytical method must simultaneously satisfy accuracy and speed. In this study, we report on the results of evaluating the accuracy of the radioactivity concentration correction by applying an efficiency transfer method that appears to satisfy these requirements to concrete standard reference material.

In Korea, extensive industry-academia-research research has already established many facilities and technologies for materials and chemical experiments on non-radioactive substances. However, few facilities have been built to analyze the physical and chemical properties of substances irradiated through neutron irradiation. Korea is planning to decommission Kori-1 and Wolsong-1 in 2027. Extensive analysis of low-level and intermediate-level materials is required to begin decommissioning these nuclear power plants. The material’s composition and level can be identified by analyzing the structure’s characteristics, and a cutting and decontamination plan can be established based on this. In addition, by conducting a nuclide analysis on the waste generated after cutting, suitability for disposal can be secured, and stable treatment can be performed. Accordingly, the Korea Decommissioning Research Institute (KRID) plans to secure infrastructure (hot cells) to analyze the characteristics of intermediate-level decommissioning waste. The goal is to secure high-dose/high-radiation decommissioning waste processing technology through Korea’s first intermediate-level hot cell, support domestic nuclear power plant decommissioning projects, and secure and verify procedures related to nuclide analysis of intermediate-level using hot cells. In addition, by possessing these material properties and nuclide analysis technology, KRID can have a foundation to conduct continuous research on low- and intermediate-level radioactive materials and decommissioning. The purpose of KRID’s establishment is to use this foundation in the future to improve the technological level of the nuclear industry as a whole through linkage between industry, academia, and research institutes.

For the disposition of radioactive wastes generated from nuclear power plant, radioisotope inventory must be analyzed to determine an activity concentration of radionuclides. Radionuclides in low- and intermediate-low-level of radioactive wastes, however, can be easily classified to easyto- measure (ETM) and difficult-to-measure (DTM) nuclides. ETM nuclides are gamma emitting nuclides that is relatively easy to measure because they do not need to be destroyed for the preprocessing. On the other hands, DTM nuclides are alpha and beta emitting nuclides that need to be destroyed for the preprocessing and also need chemical separation. Currently, measurement methods for DTM nuclides are developed and in this paper measurement methods of Fe-55, Ni-59, Ni-63, Sr-90 and Tc-99 will be introduced.