An electrical double-layer capacitor is fabricated with biomass-derived activated carbon (AC) and multi-walled carbon nanotubes (MWCNTs), which are synthesized from Pongamia pinnata fruit shell and its seed oil, respectively. The activated carbon is produced by the chemical activation process at varying carbonization temperatures from 600 to 900 °C for 5 h at a rate of 10 min in an N2 atmosphere. The surface area of activated carbon and MWCNTs is 1170 m2 g− 1 and 216 m2 g− 1, respectively. The total pore volumes of activated carbon and MWCNTs are 1.51 cm3 g− 1 and 0.5907 cm3 g− 1, respectively. The as-prepared AC and MWCNTs are characterized by surface area analysis Brunner–Emmett–Teller method (BET), X-ray diffraction, X-ray photoelectron spectroscopy and Raman spectroscopic analysis, field emission scanning electron microscopy, high-resolution transmission electron microscopy and energy-dispersive X-ray spectroscopy. The electrochemical performances of AC-AC, MWCNTs-MWCNTs and AC-MWCNTs (25:75) symmetric electrodes are studied by cyclic voltammetry, galvanostatic charge–discharge and electrochemical impedance spectroscopy. The AC-MWCNTs (25:75) single electrode performance is also studied in two different electrolytes, such as 0.5 M Na2SO4 and 0.5 M H2SO4. The fabricated AC-MWCNTs (25:75) symmetric supercapacitor cell exhibits excellent electrochemical performance in 0.5 M Na2SO4. It shows a specific capacitance of 55.51 Fg− 1, energy density 4.852 Wh Kg− 1 and power density of 199.18 W Kg− 1 at a current density of 1 Ag− 1 in the voltage window of 0–1.8 V. The AC-AC and AC-MWCNTs (25:75) symmetric supercapacitor electrodes show outstanding performance.

국제해사기구(IMO, International Maritime Organization)를 중심으로 자율운항선박 도입을 대비한 해사안전 및 보안관련 국제 협약 을 제정하고 있다. 국내에서도 선급 및 산업체를 중심으로 자율운항시스템 기술개발에 착수하고 있으며 연안선박에서 발생하는 사고를 줄이기 위해 연안선박을 대상으로 하는 자율운항선박 기술적용 방안 연구가 진행되고 있다. 국내외적으로 자율운항선박에 대한 관심이 크게 증가하고 있으며 실제 개발된 기술의 검증을 위한 해상실증이 본격적으로 추진되고 있다. 본 연구에서는 연안선박에 적용하기 위한 디지털트윈 기술 관련 실증선박과 육상 플랫폼(원격지원센터)의 설계를 위한 기초연구를 진행하였다. 디지털트윈 기술을 선박에 적용하 기 위해 8m 소형 배터리 전기추진선박을 대상으로 선정하였으며, 선박과 육상 플랫폼 간 통신을 통해 선박 항해 및 운전 데이터가 서버 시스템에 저장되고 전기추진선박의 원격제어 명령이 가능한 디지털트윈 통합 플랫폼의 기본 설계를 진행하였다. 이러한 디지털트윈 기술 을 적용한 선박 성능관리, 운항 및 운영 최적화, 예지제어 등이 가능할 것으로 판단되며, 위기상황에 대응이 가능한 안전하고 경제성 있는 디지털트윈 기술의 선박적용이 가능할 것이라 사료된다.

Niobium (Nb) is present in Ni-based alloys and stainless steels used in nuclear reactors as structural materials. Nb-93 is a naturally occurring and stable isotope of niobium and Nb-94 (half-life = 20,000 years) is produced by neutron activation of Nb-93. Nb-94 can be present in waste streams from dismantling of nuclear power plants and treatment of the primary coolant circuit. Hence, the radioactive wastes containing active Nb-94 are disposed of in the repositories for low- and intermediate-level waste (LILW). Nb predominantly exhibits a pentavalent oxidation state (i.e., +V) within the stability field of water. Cementitious materials (concrete, mortar, and grout) are extensively utilized in LILW disposal systems as structural components and chemical agents for the stabilization of waste. Solubility defines the source term (i.e., upper concentration limit) in the repository system. However, the solubility behavior of Nb in cementitious systems at high pH remains ill-defined, and information available on the Nb solid phases controlling the solubility is scarce and often ambiguous. Sorption on cementbased materials is one of the main mechanisms controlling the retention of niobium(V) in a LILW repository, and distribution coefficients (Rd) are necessary to evaluate the retention capacity by sorption in the safety assessment of disposal systems. Available sorption data of Nb(V) on cement showed a large discrepancy in Rd, moreover, no sorption data is available for Nb(V) under conditions characterizing the first degradation stage of cement (young cement condition) at pH 13 – 13.5. In this context, the solubility of Nb was extensively investigated in porewater conditions representative of the cement degradation stage I, as well as in CaCl2-Ca(OH)2 systems. Special focus was given to the accurate characterization of the solubility-controlling solid niobium phases. We also studied the sorption of Nb(V) by hardened cement pastes (HCP) and calcium silicate hydrates (CSH, major hydrate of HCP). This work provides the results on Rd, sorption isotherm and sorption mechanisms of Nb(V). Besides, the impact of ISA (polyhydroxycarboxylic acid generated by the degradation of cellulose) on Nb(V) sorption and the dissolution of cement materials was investigated.

Nuclear power plants decommissioning is planned to be started in middle of the 2020. It is necessary to develop safety evaluation and verification technology during decommissioning to ensure the safety of security monitoring measures and maintenance measures, appropriate emergency plans and preparations for decommissioning, and the use of proven engineering when establishing decommissioning plan. For this purpose, a nuclear power plant decommissioning plan is prepared in several stages before decommissioning. When a lifetime of a nuclear power plant has reached, it needs to be decommissioned and therefore operator company should submit decommissioning plans to the National Safety and Security Commission. And safety analysis should be included in this document and it is explained in chapter 6. According to the NSSC Notice No. 2021-10, it is largely divided into principles and standards, exposure scenarios, dose assessment, residual radioactivity, abnormal events, and risk analysis. When unexpected radiological accident is happened, both public and occupational dose analysis should be conducted. However, research on the former can be found easily on the other hands, research on the latter is not active. In this paper, method of choosing scenarios of accidents during the decommissioning the nuclear power plants is briefly introduced. Accidents during nuclear power plants decommissioning cases in USA is chosen and its risk is evaluated by using risk matrix and ranked by AHP method. During the decommissioning phases, varieties of radioactive waste is expected to be generated such as contaminated concrete and metal. On the other hand, Dry Active Waste (DAW) is generated and its amount is and its amount is 7,353 drums. Characteristic of DAW is highly flammable compared to concrete or metal. Moreover, depending on method of radioactive waste conditioning and type of radioactive nuclides, release rate of the nuclides varies. Thus this type of radioactive waste is critical to fire accidents and such accident can occur extra dose exposure which exceeds the guideline of the regulatory body to workers. Therefore, in this paper, occupational dose exposure during the fire accident is conducted.

This study evaluated the synthesis of optimal materials for high efficiency adsorption and removal characteristics of Cs-137 for radioactive contaminated water, and considered thermal treatment methods to stabilize the spent adsorbent generated after treatment. We synthesized a composite adsorbent with a combination of impregnating metal ferrocyanide that improves the selectivity of Cs adsorption with zeolite capable of removing Cs as a support. The Cs removal efficiency of the composite adsorbent was evaluated, and the stability change of Cs according to the high-temperature sintering was evaluated as a stabilization method of the spent adsorbent. The metal ferrocyanide content of the adsorbent was in the range of 11.8~36.0%. The adsorption experiments were performed using a simulated liquid waste to have a total Cs concentration of 1 mg/L while containing a trace amount of Cs-137, and then gamma radioactivity was analyzed. In order to evaluate the stabilization of the spent adsorbent, heat treatment was performed in the range of 500~1,100°C, and the volatilization rate of Cs during heat treatment and the leaching rate of Cs after heat treatment were compared. In the adsorption experiment, the Cs removal efficiency was higher than 99%, regardless of the amount of metal ferrocyanide in the composite adsorbent. In the sintering experiment on the spent adsorbent, it was confirmed that there was no volatilization of Cs up to 850°C, and then the volatilization rate increased as the heating temperature increased. On the other hand, the leaching rate of Cs in the sintered adsorbent tends to significantly decrease as the heating temperature increases, so that Cs can be stabilized in the sintered body. In addition, as the content of metal ferrocyanide increases, the volatilization rate of Cs rapidly increases, indicating that the unstable metal ferrocyanide in the adsorbent may adversely affect the removal of Cs as well as the thermal treatment stability.

Radioactive waste generated in large quantities from NPP decommissioning has various physicochemical and radiological characteristics, and therefore treatment technologies suitable for those characteristics should be developed. Radioactively contaminated concrete waste is one of major decommissioning wastes. The disposal cost of radioactive concrete waste is considerable portion for the total budget of NPP decommissioning. In this study, we developed an integrated technology with thermomechanical and chemical methods for volume reduction of concrete waste and stabilization of secondary waste. The unit devices for the treatment process were also studied at bench-scale tests. The volume of radioactive concrete waste was effectively reduced by separating clean aggregate from the concrete. The separated aggregate satisfied the clearance criteria in the test using radionuclides. The treatment of secondary waste from the chemical separation step was optimally designed, and the stabilization method was found for the waste form to meet the final disposal criteria in the repository site. The final volume reduction rates of 56.4~75.4% were possible according to the application scenario of our processes under simulated conditions. The commercial-scale system designs for the thermomechanical and chemical processes were completed. Also, it was found that the disposal cost for the contaminated concrete waste at domestic NPP could be reduced by more than 20 billion won per each unit. Therefore, it is expected that the application of this technology will improve the utilization of the radioactive waste disposal space and significantly reduce the waste disposal cost.

Decommissioning waste is generated at all stages during the decommissioning of nuclear facilities, and various types of radioactive waste are generated in large quantities within a short period. Concrete is a major building material for nuclear facilities. It is mixed with aggregate, sand, and cement with water by the relevant mixing ratio and dried for a certain period. Currently, the proposed treatment method for volume reduction of radioactive concrete waste was involved thermomechanical and chemical treatment sequentially. The aggregate as non-radioactive materials is separated from cement components as contaminated sources of radionuclides. However, to commercialize the process established in the laboratory, it is necessary to evaluate the scale-up potential by using the unit equipment. In this study, bench-scale testing was performed to evaluate the scale-up properties of the thermomechanical and chemical treatment process, which consisted of three stages (1: Thermomechanical treatment, 2: Chemical treatment, 3: Wastewater treatment). In the first stage, lab, bench, and pilot scale thermomechanical tests were performed to evaluate the treated coarse aggregate and fines. In the second stage, the fine particles generated by the thermomechanical treatment process, were chemically treated using dissolution equipment, after then the removal efficiency and residual of cement in the small aggregate was compared with laboratory results. The final stage, the secondary wastewater containing contaminant nuclides was treated, and the contaminant nuclides could be removed by chemical precipitation method in the scale-up reactors. Furthermore, an additional study was required on the solid-liquid separation, which connected each part of the equipment. It was conducted to optimize the separation method for the characteristics of the particles to be separated and the purpose of separation. Therefore, it is expected that the basic engineering data for commercialization was collected by this study.

An induction melting facility includes several work health and safety risks. To manage the work health and safety risks, care must be taken to identify reasonably foreseeable hazards that could give rise to risks to health and safety, to eliminate risks to health and safety so far as is reasonably practicable. If it is not reasonably practicable to eliminate risks to health and safety, attention have to be given to minimize those risks so far as is reasonably practicable by implementing risk control measures according to the hierarchy of control in regulation, to ensure the control measure is, and is maintained so that it remains, effective, and to review and as necessary revise control measures implemented to maintain, so far as is reasonably practicable, a work environment that is without risks to health or safety. The way to manage the risks associated with induction melting works is to identify hazards and find out what could cause harm from melting works, to assess risks if necessary – understand the nature of the harm that could be caused by the hazard, how serious the harm could be and the likelihood of it happening, to control risks – implement the most effective control measures that are reasonably practicable in the circumstances, and to review control measures to ensure they are working as planned.

The Korea government decided to shut down Kori-1 and Wolsung-1 nuclear power plants (NPPs) in 2017 and 2019, respectively, and their decommissioning plans are underway. Decommissioning of a NPP generates various types of radioactive wastes such as concrete, metal, liquid, plastic, paper, and clothe. Among the various radioactive wastes, we focused on radioactive-combustible waste due to its large amount (10,000–40,000 drums/NPP) and environmental issues. Incineration has been the traditional way to minimize volume of combustible waste, however, it is no longer available for this amount of waste. Accordingly, an alternative technique is required which can accomplish both high volume reduction and low emission of carbon dioxide. Recently, KAERI proposed a new decontamination process for volume reduction of radioactivecombustible waste generated during operation and decommissioning of NPPs. This thermochemical process operates via serial steps of carbonization-chlorination-solidification. The key function of the thermochemical decontamination process is to selectively recover and solidify radioactive metals so that radioactivity of the decontaminated carbon meets the release criteria. In this work, a preliminary version of mass flow diagram of the thermochemical decontamination process was established for representative wastes. Mass balance of each step was calculated based on physical and chemical properties of each constituent atoms. The mass flow diagram provides a platform to organize experimental results leading to key information of the process such as the final decontamination factor and radioactivity of each product.

The decommissioning of nuclear-related facilities at the end of their design life generates various types of radioactive waste. Therefore, the research on appropriate disposal methods according to the form of radioactive waste is needed. This study is about the solidification of uranium contaminated soils that may occur on the site of nuclear facilities. A large amount of radioactively contaminated soil waste was generated during the decommissioning of the uranium conversion plant in KAERI, and research on the proper disposal of this waste has been actively conducted. Numerous minerals in the soil can become glass-ceramic through the phase change of minerals during the sintering process. This method is effective in reducing the volume of waste and the glassceramic waste form has excellent mechanical strength and leaching resistance. In this study, the optimum temperature and time conditions were established for the production of glass-ceramic sintered body of soil. The compressive strength and leachability of the sintered body made by applying the optimal conditions to simulated waste was confirmed. The basic physicochemical properties of simulated soil waste were identified by measuring the pH, moisture content, density, and organic matter content. The elemental compositions in the soil was confirmed by XRF. Soils were classified by particle size, and each sample was compressed with a pressure of 150 MPa or more to prepare a green body. Based on the TG-DSC analysis, an appropriate heating temperature was set (>1,000°C), and the green body was maintained in a muffle furnace for 2~6 hours. The optimal sintering conditions were selected by measuring the compressive strength and volume reduction efficiency of the sintered body for each condition. The difference between the green body and sintered body was observed by XRD and SEM. In the experiments for evaluation of additives, the selected chemical substances were mixed with the soil sample in a rotator. Based on the results of TG-DSC, sintered body was made at 850°C, and the compressive strength and volume reduction were compared. Based on the results, the most effective additive was determined, and the appropriate ratio of the additive was found by adjusting the range of 1~5 wt%. This study was confirmed that the sintered soil waste showed sufficient stability to meet the disposal criteria and effective volume reduction for final disposal.