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.

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.

The primary purpose of high temperature process of radioactive waste is to satisfy the waste acceptance criteria and volume reduction. The WAC offers the guideline of waste form fabrication process. The WAC is defined as quantitative or qualitative criteria specified by the regulatory body, or specified by and operator and approved by the regulatory body, for radioactive waste to be accepted by the operator of a repository for disposal, or by the operator of a storage facility for storage. The main objective of WAC is to protect staff and general public and environment by the containment of radioactive material, limit external radiation level, and prevent criticality. The WAC also offers systematic management of radioactive waste by standardization of waste management operations, facilitation waste tracking, ensure safe and effective operation of operating facilities, etc. Since the high temperature process for radioactive waste is considered in many countries, lots of codes and standards are considered. In many WACs, compressive strength, thermal cycle stability, radiation exposure stability, free liquid, and leachability are evaluation to understand the effect of solidified form to the disposal facility. In this paper, systematical review on waste form will be discussed. In addition, brief result of characterization of waste form will be compared.

In the Kori power plant radioactive waste storage, the concentrated waste and spent resin drums generated in the past are repacked and stored in large concrete drums. Four 200 L drums of solidified concentrated waste are packed in the square concrete. One 200 L drum of spent resin is packed inside the round concrete. In order to build a foundation for disposal of large concrete drums that generated in the past, it is necessary to develop a large concrete drum handling device and disposal suitability evaluation technology. In order to build handling equipment and establishment of disposal base, such as weight and volume, of square and round concrete containers must be identified. In addition, waste information, such as the production record of the built in drum and the type of contents, is required. Therefore, this study plans to comprehensively review the characteristics of the waste by investigating the structure of square and round concrete containers and the records of internal drum production.

Currently, the most promising fuel candidate for use in sodium fast reactors (SFRs) is metallic fuel, which is produced by a modified casting method in which the metallic fuel material is sequentially melted in an inert atmosphere to prevent volatilization, followed by melting in a graphite crucible, and then injection casting in a quartz (SiO2) mold to produce metallic fuel slugs. In previous studies, U-Zr metallic fuel slugs have been cast using Y2O3 reaction prevent coatings. However, U-Zr alloy-based metallic fuel slugs containing highly reactive rare earth (RE) elements are highly reactive with Y2O3-coated quartz (SiO2) molds and form a significant thickness of surface reaction layer on the surface of the metallic fuel slug. Cast parts that have reacted with nuclear fuel materials become radioactive waste. To decrease amount of radioactive waste, advanced reaction prevent material was developed. Each RE (Nd, Ce, Ln, Pr) element was placed on the reaction prevent material and thermal cycling experiments were carried out. In casting experiments with U-10wt% Zr, it was reported that Y2O3 layer has a high reaction prevent performance. Therefore, the reaction layer properties for RE elements with higher reactivity than uranium elements were evaluated. To investigate the reaction layer between RE and NdYO3, the reaction composition and phase properties as a function of RE content and location were investigated using SEM, EDS, and XRD. The results showed that NdYO3 ceramics had better antireaction performance than Y2O3.

In this work, subabul wood biomass was used to prepare carbon adsorbents by physical and chemical activation methods at various carbonization temperatures. The properties of the carbon adsorbents were estimated through characterization techniques such as X-ray diffraction, Fourier transform infrared spectroscopy, X–ray photo electron spectroscopy, laser Raman spectroscopy, scanning electron microscopy, CHNS-elemental analysis and N2 adsorption studies. Subabul-derived carbon adsorbents were used for CO2 capture in the temperature range of 25–70 °C. A detailed adsorption kinetic study was also carried out. The characterization results indicated that these carbons contain high surface area with microporosity. Surface properties were depended on treatment method and carbonization temperature. Among the carbons, the carbon prepared after treatment of H3PO4 and carbonization at 800 °C exhibited high adsorption capacity of 4.52 m.mol/g at 25 °C. The reason for high adsorption capacity of the adsorbents was explained based on their physicochemical characteristics. The adsorbents showed easy desorption and recyclability up to ten cycle with consistent activity.

Kori unit 1, the first PWR (Pressurized Water Reactor) in Korea, was permanent shut down in 2017. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. Solid radioactive waste depending on the characteristics of the generation was classified into reactor vessel and reactor vessel internal, large components, small metals, spent nuclear fuel storage racks, insulation, wires, concrete debris, scattering concrete, asbestos, mixed waste, soil, spent resins and filters, and dry active waste. Radiological characterization of solid radioactive waste is performed to determine the characteristics of radioactive contamination, including the type and concentration of radionuclides. It is necessary to ensure the representativeness of the sample for the structures, systems and components to be evaluated and to apply appropriate evaluation methods and procedures according to the structure, material and type of contamination. Therefore, the radiological characterization is divided into concrete and structures, systems and components, and reactor vessel, reactor vessel internal and bioshield concrete. In this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. The results of this study can be used as a basis for the preparation of the FDP for the Kori unit 1.

As the importance of radioactive waste management has emerged, quality assurance management of radioactive waste has been legally mandated and the Korea Radioactive Waste Agency (KORAD) established the “Waste Acceptance Criteria for the 1st Phase Disposal Facility of the Wolsong Lowand Intermediate-Level Waste Disposal Center (WAC)”, the detailed guideline for radioactive waste acceptance. Accordingly, the Korea Atomic Energy Research Institute (KAERI) introduced a radioactive waste quality assurance management system and developed detailed procedures for performing the waste packaging and characterization methods suggested in the WAC. In this study, we reviewed the radioactive waste characterization method established by the KAERI to meet the WAC presented by the KORAD. In the WAC, the characterization items for the disposal of radioactive waste were divided into six major categories (general requirements, solidification and immobilization requirements, radiological, physical, chemical, and biological requirements), and each subcategories are shown in detail under the major classification. In order to satisfy the characterization criteria for each detailed item, KAERI divided the procedure into a characterization item performed during the packaging process of radioactive waste, a separate test item, and a characterization item performed after the packaging was completed. Based on the KAERI’s radioactive waste packaging procedure, the procedure for characterization of the above items is summarized as follows. First, during the radioactive waste packaging process, the characterization corresponding to the general requirements (waste type) is performed, such as checking the classification status of the contents and checking whether there are substances unsuitable for disposal, etc. Also, characterization corresponding to the physical requirements is performed by checking the void fraction in waste package and visual confirmation of particulate matter, substances containg free water, ect. In addition, chemical and biological requirements can be characterized by visually confirming that no hazardous chemicals (explosive, flammable, gaseous substances, perishables, infectious substances, etc.) are included during the packaging process, and by taking pictures at each packaging steps. Items for characterization using separate test samples include radiological, physical, and chemical requirements. The detailed items include identification of radionuclide and radioactivity concentration, particulate matter identification test, free water and chelate content measurement tests, etc. Characterization items performing after the packaging is completed include general requirements such as measuring the weight and height of packages and radiological requirements such as measurements of surface dose rate and contamination, etc. All of the above procedures are proceduralized and managed in the radioactive waste quality assurance procedure, and a report including the characterization results is prepared and submitted when requesting acceptance of radioactive waste. The characterization of KAERI’s radioactive waste has been systematically established and progressed under the quality assurance system. In the future, we plan to supplement various items that require further improvement, and through this, we can expect to improve the reliability of radioactive waste management and activate the final disposal of KAERI’s radioactive waste.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR-1&2), the first research reactors in South Korea, began in 1997. Approximately 5,000 tons of waste will be generated when the contaminated buildings are demolished. Various types of radioactive waste are generated in large quantities during the operation and decommissioning of nuclear facilities, and in order to dispose of them in a disposal facility, it is necessary to physico-chemically characterize the radioactive waste. The need to transparently and clearly conduct and manage radioactive waste characterization methods and results in accordance with relevant laws, regulations, acceptance standards is emerging. For radioactive waste characterization information, all information must be provided to the disposal facility by measuring and testing the physical, chemical, and radiological characteristics and inputting related documents. At this time, field workers have the inconvenience of performing computerized work after manually inputting radioactive waste characterization information, and there is always a possibility that human errors may occur during manual input. Furthermore, when disposing of radioactive waste, the production of the documents necessary for disposal is also done manually, resulting in the aforementioned human error and very low production efficiency of numerous documents. In addition, as quality control is applied to the entire process from generation to treatment and disposal of radioactive waste, it is necessary to physically protect data and investigate data quality in order to manage the history information of radioactive waste produced in computerized work. In this study, we develop a system that can directly compute the radioactive waste characterization information at the field site where the test and measurement are performed, protect the stored radioactive waste characterization data, and provide a system that can secure reliability.

The recycling of solid waste materials to fabricate carbon-based electrode materials is of great interest for low-cost green supercapacitors. In this study, porous carbon foam (PCF) was prepared from waste floral foam (WFF) as an electrode material for supercapacitors. WFF was directly carbonized at various temperatures of 600, 800, and 1,000 oC under an inert atmosphere. The WFF-derived PCF (C-WFF) was found to have a specific surface area of 458.99 m2/g with multi-modal pore structures. The supercapacitive behavior of the prepared C-WFF was evaluated using a three-electrode system in a 6 M KOH aqueous electrolyte. As a result, the prepared C-WFF as an active material showed a high specific capacitance of 206 F/g at 1 A/g, a rate capability of 36.4 % at 20 A/g, a specific power density of 2,500 W/kg at an energy density of 2.68 Wh/kg, and a cycle stability of 99.96 % at 20 A/g after 10,000 cycles. These results indicate that the C-WFF prepared from WFF could be a promising candidate as an electrode material for high-performance green supercapacitors.

Radioactive materials emitted from nuclear accident or decommissioning cause soil contamination over wide areas. In the event of such a wide area of contaminated soil, decontamination is inevitable for residents to reside and reuse as industrial land. There are many ways to decontaminate these contaminated soils, but in urgent situations, the soil washing, which has a short process period and relatively high decontamination efficiency, is considered the most suitable. However, the soil washing process of removing fine soil and cesium by using washing liquid as water and adding a flocculating agent (J-AF) generates slurry/sludge-type secondary waste (Cs-contaminated soil + flocculating agent). Since this form of sludge contaminants cannot be disposed, solidification is needed using an appropriate solidification agent to treat wastes for disposal. Therefore, this study devised a treatment method of contaminated fine soils occurring after the soil washing process. This investigation prepared the simulated wastes of contaminated fine soils generated after the soil washing, and pelletized the samples using a roll compactor under the optimum operating conditions. The optimum conditions of the device were determined in the pre-test. Roll speed, feeding rate, and hydraulic pressure were 1.5 rpm, 25 rpm, and 28.44 MPa, respectively. The waste forms were manufactured by incorporating created pellets (H 6.5 × W 9.4 mm) using polymers as solidification agents. Used polymers were main ingredient (YD-128), hardener (G-1034), and diluent (LGE). The optimum mixing ratio was YD-128 : G-1034 = 65 : 35 phr, and LGE was added in an amount of 10wt% of the total mixture. To confirm the disposal suitability of the manufactured waste forms, characterization evaluation was carried out (compressive strength, thermal cycling, immersion, and leaching test). Characterization evaluation revealed a minimum compressive strength of 23.1 MPa, far exceeding 3.44 MPa of the disposal facility waste acceptance criteria. Compressive strength increased to the highest value of 31.90 MPa after immersion test. To examine leaching characteristics, the pH, Electrical Conductivity (EC) and leachability index (􀜮􀯜) of leachates were identified. As results, pH and EC consistently increased or remained constant with leaching time. The average of Co, Cs and Sr nuclides was 17.76, 17.38 and 14.04, respectively, exceeding the value of 6 in the waste acceptance criteria. Effective waste treatment/ disposal can be achieved without increasing volumes of sludge/slurry by enhancing the technique of this research by performing additional studies in the future.

This work reveals a modified method for the preparation of activated carbon (P-ACA) using low-cost materials (mix natural asphalt: polypropylene waste). The P-ACA was prepared at 600 °C by assisting KOH and HF. The morphological variations and chemical species of the P-ACA were characterized using SEM–EDX and FTIR. The active surface area, density and ash content of the P-ACA were also investigated. Adsorption properties of P-ACA were used for the thermodynamic and kinetic study of 4-((2-hydroxy naphthalenyl) diazenyl) antipyrine (HNDA), which was prepared as a novel azo dye in this work. The optimal conditions (initial concentration, adsorbent dose, contact time and temperature) of the adsorption process were determined. Adsorption isotherms (Freundlich and Langmuir) were applied to the experimental data. These isothermal constants were used to describe the nature of the adsorption system, and the type of interaction between the dye and the P-ACA surface. The results have indicated that the mixture (Natural asphalt-polypropylene waste) is efficient for the synthesis of P-ACA. The synthesized P-ACA demonstrates the presence of pores on the surface with various diameter ranges (from 1.4 to 4.5 μm). Furthermore, P-ACA exhibits an active surface area of 1230 m2 g−1, and shows a high adsorption capacity for HNDA.

Hierarchically porous carbon materials with high nitrogen functionalities are extensively studied as highperformance supercapacitor electrode materials. In this study, nitrogen-doped porous carbon textile (N-PCT) with hierarchical pore structures is prepared as an electrode material for supercapacitors from a waste cotton T-shirt (WCT). Porous carbon textile (PCT) is first prepared from WCT by two-step heat treatment of stabilization and carbonization. The PCT is then nitrogendoped with urea at various concentrations. The obtained N-PCT is found to have multi-modal pore structures with a high specific surface area of 1,299 m2 g−1 and large total pore volume of 1.01 cm3 g−1. The N-PCT-based electrode shows excellent electrochemical performance in a 3-electrode system, such as a specific capacitance of 235 F g−1 at 1 A g−1, excellent cycling stability of 100 % at 5 A g−1 after 1,000 cycles, and a power density of 2,500 W kg−1 at an energy density of 3.593 Wh kg−1. Thus, the prepared N-PCT can be used as an electrode material for supercapacitors.

The radionuclide inventory in radioactive waste from nuclear power plants should be determined to secure the safety of final repositories. As an alternative to time-consuming, labor-intensive, and destructive radiochemical analysis, the indirect scaling factor (SF) method has been used to determine the concentrations of difficult-to-measure radionuclides. Despite its long history, the original SF methodology remains almost unchanged and now needs to be improved for advanced SF implementation. Intense public attention and interest have been strongly directed to the reliability of the procedures and data regarding repository safety since the first operation of the low- and intermediate-level radioactive waste disposal facility in Gyeongju, Korea. In this review, statistical methodologies for SF implementation are described and evaluated to achieve reasonable and advanced decision-making. The first part of this review begins with an overview of the current status of the scaling factor method and global experiences, including some specific statistical issues associated with SF implementation. In addition, this review aims to extend the applicability of SF to the characterization of large quantities of waste from the decommissioning of nuclear facilities.

우리는 실험과 MCNP 시뮬레이션을 통해 전알파 분석법의 한계를 설명하였다. 국내에서 중·저준위 방사성폐기물 인도 규 정 관련, 전알파 분석법은 방사성폐기물을 처분하기 위해 반드시 규명해야 할 방사성 특성평가 인자이다. 전알파 분석법은 시료 준비 절차가 간단하고 신속한 분석 결과를 제공하지만, 정량분석 인자로 사용하는 것은 적절하지 않다. KCl과 241Am 을 이용하여 시편 건조고형물 무게에 따른 전알파 계측효율을 평가하였다. 동일한 무게의 시편일지라도 계측효율의 차이가 20% 나는 것을 확인하였고, 이는 시편의 물리적 형태가 서로 다르기 때문인 것으로 보인다. 토양 중 우라늄을 화학분리 한 후, ICP-MS로 우라늄을 직접 측정한 결과와 전알파 농도를 비교하였다. 전알파는 실제 우라늄 농도에 비해 50% 과소평가되 었다. 알파핵종별 전알파 계측효율이 최대 3배 차이 나기 때문에, 전알파 분석결과는 개별 알파핵종의 합과 비교하기 보다는 스크리닝 개념으로 사용하는 것이 적절하다.

In this study, hair waste was converted into active carbon for the first time and its characteristics were analyzed. As chemical activation tool, zinc chloride ( ZnCl2) was impregnated and then carbonized under different temperatures (250–300 °C). Scanning Electron Microscope (SEM) images showed an increase in the pore density, radius and volume of pores. X-ray diffraction analysis (XRD) showed that the samples had an amorphous structure. In Fourier-transform infrared (FT-IR) spectroscope analysis, C=C and N–H vibrations observed in 1515–1520 cm−1 wave number of protein molecules were found to disappear with the increase in temperature. With Raman spectroscopy, the behaviors of D peak at 1344 cm−1 wave number and G peak at 1566 cm−1 wave number expressing structure layout in carbonized structures were analyzed depending on the temperatures. Between these intensities, (ID/IG) the rate was found to differ in direct proportion to temperature. XRD spectrums showed that the samples are converted into a more irregular crystal structure. All these results implied that the waste hair mass could be used as an adsorbant material.