우리나라에서는 최근 아동 돌봄서비스에 대한 사회적 요구에 따라 다양한 서비스의 확대를 통해 사각지대를 해소하고자 노력해 왔지만, 공간적 관점은 부족하였다. 이 연구의 주된 목적은 아동의 생애주기를 고려하여 돌봄서비스들의 공간적 접근성을 종합적으로 분석함으로써 돌봄 사막과 같은 사각지대의 공간적 패턴을 탐색하는 것이다. 이를 위해 먼저 아동 돌봄서비스의 공간적 접근성에 대한 이론적 논의와 선행연구를 검토하였다. 이어 2SFCA를 이용해 생애주기에 따른 돌봄서비스의 공간적 접근성을 측정하고, 이를 종합해 돌봄서비스의 사각지대를 파악하기 위한 방법을 제시하였다. 경상북도의 시군을 사례로 분석한 주요 결과는 다음과 같다. 첫째, 돌봄 사각지대는 크게 거리와 공급의 부족으로 구분될 수 있으며, 두 경우 모두 중학생 돌봄시설 에 대하여 가장 접근성이 취약한 것으로 나타났다. 둘째, 모든 시기에 있어서 도시지역이 농촌지역보다 접근성이 우수하였고, 읍･면 지역의 돌봄의 사각지대가 두드러졌다. 셋째, 접근성이 취약한 지역, 이른바 돌봄 사막은 영유아 시설에서 중학생시설에 이르기까지 공간적으로 유사한 패턴을 보였다. 생애주기를 고려하여 아동 돌봄서비스의 사각지대가 공간적 패턴을 이루고 있음을 파악한 본 연구의 결과는 향후 돌봄서비스 제공에 있어 중요한 정책적 시사점을 제공한다.

Democratic People’s Republic of Korea (DPRK) has produced weapon-grade plutonium in a graphite-moderated experimental reactor at the Yongbyon nuclear facilities. The amount of plutonium produced can be estimated using the Graphite Isotope Ratio Method (GIRM), even without considering specific operational histories. However, the result depends to some degree on the operational cycle length. Moreover, an optimal cycle length can maximize the number of nuclear weapons made from the plutonium produced. For conservatism, it should be assumed that the target reactor was operated with an optimal cycle length. This study investigated the optimal cycle length using which the Calder Hall MAGNOX reactor can achieve the maximum annual production of nuclear weapons. The results show that lower enrichment fuel produced a greater number of critical plutonium spheres with a shorter optimal cycle length. Specifically, depleted uranium (0.69wt%) produced 5.561 critical plutonium spheres annually with optimal cycle lengths of 251 effective full power days. This research is crucial for understanding DPRK’s potential for nuclear weapon production and highlights the importance of reactor operational strategy in maximizing the production of weapons-grade plutonium in MAGNOX reactors.

PURPOSES : The study aims to establish a comprehensive life cycle assessment model for bridges in South Korea considering domestic carbon emission factors. The main aims are to evaluate the carbon emission of bridge construction, focusing on the Seong-ri Bridge as a case study, and to improve national environmental policies and management strategies. METHODS : We utilized the life cycle assessment (LCA) methodology, adhering to standards set by ISO, to categorize each phase of the bridge's life cycle. The process involved selecting the bridge type based on the compilation of a detailed analysis range. The analysis covered various stages from raw material supply (A1-A3) to construction (A4-A5) and maintenance (B2-B5), excluding certain stages due to data unavailability. Carbon emission factors were then applied to quantify emissions at each stage. RESULTS : The findings indicate that the raw material production phase (A1-A3) contributes to approximately 96% of the total carbon emissions, highlighting its significant impact. We report detailed calculations of emissions using domestically developed emission factors for materials such as steel and concrete and establish a carbon emission per unit length measure for comparative analysis with other infrastructure. CONCLUSIONS : We leveraged LCA ISO standards to analyze each stage of the Seong-ri bridge, calculating its carbon emissions based on domestic factors for CO2, CH4, and N2O. By tailoring the study to Korea-specific emission factors, we develop a greenhouse gas model closely aligned with the nation’s environmental conditions. The results contribute to improving environmental impact assessments and strategically aiding national policy and management decisions.

본 연구는 소량 다품종의 화학물질을 다루는 연구실의 안전관리를 강화하기 위해 전과정관리 체계 알고리즘을 제안하는데 그 목적이 있다. 이를 위해 최근 10년간의 연구실 사고 사례, 관련 법규, 선행 연구 및 FGI (Focus Group Interview) 연구실 현황 분석을 통해 요구사항을 도출하고 전과정 관리체계를 개발하였으며, 이를 기반으로 전과정관리체계 알고리즘을 설계, 개발하여 연구실에 적용 및 검증하였다. 연구실 화학물질 사고 예방을 위한 안전관리는 우선 연구실 내 화학물질의 종류와 양을 정확히 파악하는 것부터 시작해야 한다. 이를 위해 화학물질관리의 전 과정을 단계별로 나누어 관리체계를 수립하고, 각 단계에서 활용되는 물질 정보, 법규 정보, 화학물질 성상별 분리 정보를 다루는 데이터베이스 항목 추출 및 알고리즘 개발 방안을 제시하였다. 본 알고리즘을 세 개의 기관에서 적용한 결과, 사고 예방 및 법규 준수 측면에서 높은 효과를 보인 것으로 평가되었으나 사고 발생률에 미치는 유의한 영향에 대해 알기 위해서는 더 많은 적용 연구가 필요하다. 본 연구에서 개발한 전과정관리체계 알고리즘의 활용은 실험실 안전과 사회적 안전을 도모하는 동시에 기업의 ESG 경영의 중요한 요소 중 하나인 기업활동의 리스크 관리와 책임경영에 도움이 될 것으로 기대한다.

내구연한이 도래한 아스팔트 혼합물은 사용자의 주행성 및 안전성 확보를 위해 주기적인 유지·보수를 실시한다. 과거에는 유지·보수 과정에서 발생된 폐아스팔트 혼합물을 각종 건설현장에서 단순 매립재로서 활용하였으나, 재활용 아스팔트 혼합물의 배합설계 기술이 확립된 이후에는 도로포장재료로서의 재활용되어왔다. 하지만 현재 시공된 재활용 아스팔트 혼합물 또한 내구연한이 다가옴에 따라 노화된 재활용 아스팔트 혼합물의 처리방안 수립이 필요한 시점이다. 본 연구에서는 한번 재활용된 아스팔트 혼합물이 기존의 재활용 배합설계법으로 반복적인 재활용이 가능한지 검증해보고자 하였다. 이를 위해 아스팔트 혼합물의 물성을 결정하는 가장 큰 요인 중 하나인 아스팔트의 물성이 노화 및 재생을 반복할때 어떻게 변하는지를 시험을 통해 분석하였다. 아스팔트 바인더의 노화를 모사하기 위해 공용성등급시험에 사용되는 단기노화 장비(Rolling Thin Film Oven, RTFO)와 장기노화 장비(Pressure Aging Vessel, PAV)을 활용하 였다. 노화된 아스팔트의 회생을 위한 재생첨가제 사용량은 국토부 시공지침의 배합설계법을 참고하였다. 실험결과, 노화된 바인더는 회생시 원바인더에 비해  sin는 감소하였으나, 회생된 바인더 간에는 유사한 결과값을 보였다. 반면 단기노화 시료는 회생이 반복 됨에 따라  sin이 감소한 경향을 보였으며, 장기노화시료는 회생이 반복되어도  sin가 유사한 것으로 확인되었다.

Wastewater management is increasingly emphasizing economic and environmental sustainability. Traditional methods in sewage treatment plants have significant implications for the environment and the economy due to power and chemical consumption, and sludge generation. To address these challenges, a study was conducted to develop the Intermittent Cycle Extended Aeration System (ICEAS). This approach was implemented as the primary technique in a full-scale wastewater treatment facility, utilizing key operational factors within the standard Sequencing Batch Reactor (SBR) process. The optimal operational approach, identified in this study, was put into practice at the research facility from January 2020 to December 2022. By implementing management strategies within the biological reactor, it was shown that maintaining and reducing chemical quantities, sludge generation, power consumption, and related costs could yield economic benefits. Moreover, adapting operations to influent characteristics and seasonal conditions allowed for efficient blower operation, reducing unnecessary electricity consumption and ensuring proper dissolved oxygen levels. Despite annual increases in influent flow rate and concentration, this study demonstrated the ability to maintain and reduce sludge production, electricity consumption, and chemical usage. Additionally, systematic responses to emergencies and abnormal situations significantly contributed to economic, technical, and environmental benefits.

산오이풀(Sanguisorba hakusanensis)은 한국의 자생식물 이며 정원소재로써 가치가 있지만, 생육 및 생리적 특성 및 정 원 적응 여부에 대하여 알려진 정보가 많지 않아 이용에 어려 움을 겪고 있다. 본 연구에서는 자생식물인 산오이풀의 관수 주기 및 NaCl 농도에 따른 생장, Fv/Fm, NPQ, 성분 변화, 무기성분 변화를 조사하여 내건 및 내염성 보유 여부, 생육 한 계 범위, 스트레스 환경에서 생육을 유지하기 위한 반응을 파 악하고자 했다. 실험 결과 NaCl 무처리구의 관수주기에 따른 성분 분석에서 엽록소 함량의 감소를 제외하고 유의한 차이가 나타나지 않았으나 이는 토양수분함량이 건조 스트레스를 유 발할 정도로 감소하지 않았기 때문으로 보인다. 염 처리에서 는 2주 이후 급격한 스트레스 반응이 나타났고 3주차부터 고 사하기 시작하여 6주차에 모든 개체가 최종 고사했다. 이러한 결과는 2주까지 염 스트레스에 의해 유발되는 2가지 스트레 스 중 초기에 나타나는 삼투 스트레스에는 저항하였으나 이후 나타나는 NPQ의 감소 등 이온 스트레스에 의해 유발된 광합 성 기구 붕괴로 인해 정상적인 생육을 유지할 수 없었기 때문 에 나타난 것으로 보인다. 그러나 무기이온 분석은 이온 스트 레스에 저항하기 위한 메커니즘의 존재 가능성을 시사하였다. 상대적으로 염 농도가 낮을 때에는 세포내 Ca2+ 및 K+ 수준이 높았는데, 이는 Ca2+ 수준이 높아짐에 따라 Na+를 세포 밖으 로 방출시키는 단백질, Na+를 K+와 함께 수송하는 단백질이 기능하여 Na+축적을 지연시키는 반응이 있었음을 시사한다. 그러나 NaCl을 고농도로 처리했을 때는 이러한 반응이 관찰 되지 않았다. 따라서 산오이풀은 염 스트레스에 의해 야기되 는 삼투 스트레스에 강한 저항성을 가지고 있고 이온 독성을 줄이기 위한 메커니즘으로 Na+ 세포내 축적을 지연시키는 것으로 보이지만, 심한 염 스트레스를 받았을 때 나타나는 급격 한 반응에서 이러한 메커니즘이 기능하지 못하고 이온독성에 매우 취약한 것으로 여겨진다. 본 연구를 통해 자생식물인 산 오이풀의 활용을 늘리는 데 기초적인 자료를 제공할 수 있을 것으로 생각된다.

PURPOSES : This study is aimed to economic analysis of the ferronickel slag pavement method carried out to suggest the necessity of developing ferronickel slag pavement technology. METHODS : A life cycle cost analysis of the application of the Ferronickel Slag pavement method and the cutting + overlay pavement method was performed to compare the economic indicators and greenhouse gas emissions for each pavement method. RESULTS : As a result of the analysis, regardless of the Ferronickel Slag mixing rate, if the common performance of the Ferronickel Slag pavement method is the same or superior to the existing pavement method, it is more economical than the existing pavement method. Furthermore, the lower the maintenance cost of the Ferronickel Slag pavement method, the higher the economic feasibility due to the high Ferronickel Slag mixing rate. Greenhouse gas emissions can be reduced from at least 9% to up to 53% through the application of the Ferronickel Slag pavement method, except for some scenario analysis results. CONCLUSIONS : This study provided that the Ferronickel Slag pavement method was superior to the existing pavement method in terms of economic and environmental aspects. Therefore, it was found that the objective justification of developing road pavement technology using Ferronickel Slag was secured.

The need for the development of sustainable, efficient, and green radioactive waste disposal methods is emerging with the saturation of spent nuclear waste storage facilities in the Republic of Korea. Conventional radioactive waste management methods like using cement or glass have drawbacks such as high porosity, less chemical stability, high energy consumption, carbon dioxide production, and the generation of secondary wastes, etc. To address this gigantic issue of the planet, we have designed a study to explore the potential of alternative materials having easy processability, low carbon emissions and more chemical stability such as ceramic (hydroxyapatite, HAP) and alkali-activated materials (geopolymers, GP) to capture the simulated radioactive cobalt ions from the contaminated water and directly solidify them at low temperatures. Physical and mechanical properties of HAP alone and 15wt% GP incorporated HAP (HAP-GP- 15) composite were studied and compared. The surface of both materials was fully sorbed with an excess amount of Co(II) ions in the aqueous system. Co(II) sorbed powders were separated from aqueous media using a centrifuge machine operating at 5,000 RPM for 10 minutes and dried at 100°C for 8 hours. The dried powders were then placed in stainless steel molds, and shaped into cylindrical pellets using a uniaxial press at a pressure of 1 metric ton for 1 minute. The pellets were sintered at 1,100°C for 2 hours at a heating rate of 10°C/min. Following this, the water absorption, density, porosity, and compressive strength of the polished pellets were measured using standard methods. Results showed that HAP has a greater potential for decontamination and solidification of Co(II) due to its higher density (2.65 g/cm3 > 1.90 g/cm3), less open porosity (16.2±2.9% < 42.4 ±0.9%) and high compressive strength (82.1±10.2 MPa > 6.9±0.8 MPa) values at 1,100°C compared to that of HAP-GP-15. Nevertheless, further study with different constituent ratio of HAP and GP at various temperatures is required to fully optimize the HAP-GP matrix for waste solidifications.

A comprehensive understanding of actinide coordination chemistry and its structure is essential in many aspects of the nuclear fuel cycle, such as fuel reprocessing, waste management, reactor safety, and non-proliferation efforts. Managing radioactive waste generated during the nuclear fuel cycle has recently become more important, accordingly increasing the importance of designing appropriate waste forms and storage solutions for long-term waste disposal. Compared to the increase in the need for understanding the chemistry of major radioactive elements, the information on the local structure of the radioactive elements, especially actinides, remains unknown. To probe this issue, X-ray absorption fine structure (XAFS) can be applied. By analyzing the EXAFS (extended X-ray absorption fine structure) and XANES (X-ray absorption near edge structure), the local structure around atoms can be determined. However, the radioactive properties of the nuclides hindered the measurement of EXAFS and XANES, due to the difficulties of preparation, containment, and transfer of the sample. To measure the EXAFS of various compounds regarding the back-end nuclear fuel cycle, laboratory-based EXAFS (hiXAS, HP spectroscopy) has been introduced which can measure the EXAFS and XANES at the energy range of 5-18 keV. Compounds of Copper (Cu foil, CuO samples), Zirconium (Zr foil), and Europium (Eu2O3) were used for the verification of the laboratory -based EXAFS at a given energy range. The measured EXAFS spectrum of various compounds exhibit good agreement with the theoretical data, showing an R-factor of less than 0.02. It was found that each graph has a first peak corresponding to 2.55Å for Cu foil (Cu-Cu), 1.93Å for CuO samples (Cu-O), 3.23Å for Zr foil (Zr-Zr), and from 2.32Å to 2.34Å for Eu2O3 (Eu-O), which agree well with other values from the literature. From the result, it can be implied that this equipment can be used especially in the back-end nuclear fuel cycle field to enhance the understanding of local structure in radiochemistry.

Molten chloride salts have received considerable research attention as potential nuclear fuel and coolant candidates for molten salt reactors. However, there are several challenges, especially for structural materials due to the selective dissolution of chromium (Cr) in the molten chloride salts environment. Understanding the compatibility of uranium (U), which is used as nuclear fuel in molten salt reactors, with Cr in molten chloride salts is critical for designing the molten salt reactor structure. Therefore, in this study, the cyclic voltammetry (CV) was used to investigate the electrochemical behaviors of U and Cr. The diffusion coefficients and formal potentials were obtained. The electrochemical properties of uranium and chromium were investigated by CV in molten NaCl-MgCl2 salt at 600°C. Tungsten rods for working and counter electrode, and Ag/AgCl for reference electrode were utilized in this experiment. UCl3 made from the chemical dissolution of U rods and CrCl2 (Sigma-Aldrich, 99.99%) were used. Diffusion coefficients (D) of U and Cr were calculated by measuring reduction peak current of U3+/U and Cr2+/Cr from CV curves and using the Berzins-Delahay equation; D (U3+/U) = 3.0×10-5 cm2s-1 and D (Cr2+/Cr) = 3.3×10-5 cm2s-1. The formal potentials were also calculated by using the reduction peak potential obtained from CV results; E0’ (U3+/U) = -1.173 V and E0’ (Cr2+/Cr) = -0.321 V. The ionization tendency was investigated by comparing each reduction peak potential. The reduction peak potential Ep,c was increasing order of Ep,c (U3+/U) < Ep,c (Cr2+/Cr) < Ep,c (U4+/U3+). It can be seen that in the presence of U4+ and Cr metals, the Cr in the alloy can dissolve into Cr2+, but in the presence of U3+ and Cr metals, the Cr in the alloy does not dissolve into Cr2+. By analyzing the CV curve, diffusion coefficients and formal standard potentials were obtained. The result of comparing reduction peak potentials suggests that the nuclear fuel using U4+ should be inhibited to prevent the selective dissolution of Cr.

The ultimate objective of deep geological repositories is to achieve complete segregation of hazardous radioactive waste from the biosphere. Thus, given the possibility of leaks in the distant future, it is crucial to evaluate the capability of clay minerals to fulfill their promising role as both engineered and natural barriers. Selenium-79, a long-lived fission product originating from uranium- 235, holds significant importance due to its high mobility resulting from the predominant anionic form of selenium. To investigate the retardation behaviors of Se(IV) in clay media by sorption, a series of batch sorption experiments were conducted. The batch samples consisted of Se(IV) ions dissolved in 0.1 M NaCl solutions, along with clay minerals including kaolinite, montmorillonite, and illite-smectite mixed layers. The pH of the samples was also varied, reflecting the shift in the predominant selenium species from selenious acid to selenite ion as the environment can shift from slightly acidic to alkaline conditions. This alteration in pH concurrently promotes the competition of hydroxide ions for Se(IV) sorption on the mineral surface as the pH increases and impedes the selective attachment of selenium. The acquired experimental data were fitted through Langmuir and Freundlich sorption isotherms. From the Freundlich fit data, the distribution coefficient values of Se(IV) for kaolinite, montmorillonite, and illite-smectite mixed layer were derived, which exhibited a clear decrease from 91, 110, 62 L/kg at a pH of 3.2 to 16, 6.3, 12 L/kg at a pH of 7.5, respectively. These values derived over the pH range provide quantitative guidance essential for the safety assessment of clay mineral barriers, contributing to a more informed site selection process for deep geological repositories.

This program aims to build a specialized and converged educational platform for the training of students in the back-end nuclear fuel cycle and cultivate integrated human resources encompassing majors, generations, and fields. To achieve this, we have established an infrastructure for integrated education and training in the radiochemistry and back-end nuclear fuel cycle and operated specialized educational courses linked with special lectures, experimental practices, and field trips. Firstly, to construct an integrated educational and training infrastructure for the back-end nuclear fuel cycle, we formed a committee of experts from both inside and outside the institution and built an advanced radiochemistry laboratory equipped with physical and chemical analysis instruments. Through a comprehensive educational program involving theory, experiments, and discussions, we have established an integrated curriculum across adjacent majors and interdisciplinary studies. We also operate short-term education and experimental training programs (e.g., summer and winter schools for the back-end nuclear fuel cycle). Secondly, the program has connected leading researchers domestically and internationally, as well as the next generation of scholars. The program offers long-term educational opportunities and internships targeting both undergraduate and graduate students. To support this, we continuously offer expert colloquiums and individual research internships. Through regular committee meetings and workshops, we focus on nurturing the integrated talents necessary for the back-end nuclear fuel cycle. Through this program, students from various fields are being trained as competent integrated human resources capable of addressing various issues in the back-end nuclear fuel cycle. It is expected that this will enable us to supply specialized technical personnel in the back-end nuclear field in line with mid-to-long-term demands.

One of the important components of a nuclear fuel cycle facility is a hot cell. Hot cells are engineered robust structures and barriers, which are used to handle radioactive materials and to keep workers, public, and the environment safe from radioactive materials. To provide a confinement function for these hot cells, it is necessary to maintain the soundness of the physical structure, but also to maintain the negative pressure inside the hot cell using the operation of the heating, ventilation, and air conditioning (HVAC) systems. The negative pressure inside the hot cells allows air to enter from outside hot cells and limits the leakage of any contaminant or radioactive material within the hot cell to the outside. Thus, the HVAC system is one of the major components for maintaining this negative pressure in the hot cell. However, as the facility ages, all the components of the hot cell HVAC system are also subject to age-related deterioration, which can cause an unexpected failure of some parts. The abnormal operating condition from the failure results in the increase of facility downtime and the decrease in operating efficiency. Although some major parts are considered and constructed in redundancy and diversity aspects, an unexpected failure and abnormal operating condition could result in reduction of public acceptance and reliability to the facility. With the advent of the 4th Industrial Revolution, prognostics and health management (PHM) technology is advancing at a rapid pace. Korea Hydro & Nuclear Power, Siemens, and other companies have already developed technologies to constantly monitor the integrity of power plants and are applying the technology in the form of digital twins for efficiency and safety of their facility operation. The main point of PHM, based on this study, is to monitor changes and variations of soundness and safety of the operation and equipment to analyze current conditions and to ultimately predict the precursors of unexpected failures in advance. Through PHM, it would be possible to establish a maintenance plan before the failure occurs and to perform predictive maintenance rather than corrective maintenance after failures of any component. Therefore, it is of importance to select appropriate diagnostic techniques to monitor and to diagnose the condition of major components using the constant examination and investigation of the PHM technology. In this study, diagnostic techniques are investigated for monitoring of HVAC and discussed for application of PHM into nuclear fuel cycle facilities with hot cells.

In the nuclear fuel cycle (NFC) facilities, the failure of Heating Ventilation and Air Conditioning (HVAC) system starts with minor component failures and can escalate to affecting the entire system, ultimately resulting in radiological consequences to workers. In the field of air-conditioning and refrigerating engineering, the fault detection and diagnosis (FDD) of HVAC systems have been studied since faults occurring in improper routine operations and poor preventive maintenance of HVAC systems result in excessive energy consumption. This paper aims to provide a systematic review of existing FDD methods for HVAC systems therefore explore its potential application in nuclear field. For this goal, typical faults and FDD methods are investigated. The commonly occurring faults of HVAC are identified through various literature including publications from International Energy Agency (IEA) and American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE). However, most literature does not explicitly addresses anomalies related to pressure, even though in nuclear facilities, abnormal pressure condition need to be carefully managed, particularly for maintaining radiological contamination differently within each zone. To build simulation model for FDD, the whole-building energy system modeling is needed because HVAC systems are major contributors to the whole building’s energy and thermal comfort, keeping the desired environment for occupants and other purposes. The whole-building energy modeling can be grouped into three categories: physics-based modeling (i.e., white-box models), hybrid modeling (i.e., grey-box models), and data-driven modeling (i.e., black-box models). To create a white-box FDD model, specialized tools such as EnergyPlus for modeling can be used. The EnergyPlus is open source program developed by US-DOE, and features heat balance calculation, enabling the dynamic simulation in transient state by heat balance calculation. The physics based modeling has the advantage of explaining clear cause-and-effect relationships between inputs and outputs based on heat and mass transfer equations, while creating accurate models requires time and effort. Creating a black-box FDD model requires a sufficient quantity and diverse types of operational data for machine learning. Since operation data for HVAC systems in existing nuclear cycle facilities are not fully available, so efforts to establish a monitoring system enabling the collection, storage, and management of sensor data indicating the status of HVAC systems and buildings should be prioritized. Once operational data are available, well-known machine learning methods such as linear regression, support vector machines, random forests, artificial neural networks, and recurrent neural networks (RNNs) can be used to classify and diagnose failures. The challenge with black-box models is the lack of access to failure data from operating facilities. To address this, one can consider developing black-box models using reference failure data provided by IEA or ASHRAE. Given the unavailability of operation data from the operating NFC facilities, there is a need for a short to medium-term plan for the development of a physics-based FDD model. Additionally, the development of a monitoring system to gather useful operation data is essential, which could serve both as a means to validate the physics-based model and as a potential foundation for building data-driven model in the long term.

Recently, as carbon-neutral energy sources become increasingly important worldwide, SMRs (Small Modular Reactors), which offer significantly enhanced safety, versatility, and mobility compared to conventional nuclear reactors, are gaining attention as a viable alternative. SMR generally refers to small modular reactors with a power output of 300 MWe or less. Unlike conventional reactors, SMRs are characterized by an all-in-one design where peripheral systems and equipment are all integrated into the reactor itself, leading to enhanced reliability and durability. Additionally, the nuclear fuel reloading cycle is significantly extended compared to traditional reactors, resulting in a substantial reduction in maintenance difficulty and costs. Researchers have taken note of these characteristics of SMRs, particularly the extended fuel reloading cycle. Therefore, we have initiated the initial design of an ultra-small Micro Modular Reactor with an electricity generation capacity of 10 MWe and a fuel cycle of up to 55 years, with the goal of using it as a propulsion power source for various transportation modes, especially ships. Our design of MMR, called ‘ARA,’ is primarily distinguished by its use of U233 and Th232 fuels instead of conventional UO2 fuel. Due to various features of ‘ARA,’ including different fuel compositions, ARA is predicted to exhibit several characteristic features compared to conventional PWRs. In this study, among these characteristics, we focused on predicting changes in material composition within the fuel rod during the extended cycle operation of high-enriched fuel, rather than short-cycle operation using low-enriched fuel, unlike conventional reactors. The primary goal of this research is to observe the behavior of the composition of the materials used in the fuel cycle of the MMR, which utilizes U233 and Th232 fuels instead of UO2. Considering the difficulties in the spent nuclear fuel disposal process, many different trials were made to minimize the fission products of ARA, which differs from conventional reactors in terms of fuel type, size, and fuel cycle, in relation to waste generation.

The Fukushima-Daiichi accident in 2011 revealed the limitations of Zr-alloys in accident scenarios where severe steam oxidation led to the liberation of heat and hydrogen and the destruction of the reactor core. In response to this accident, there has been a concerted effort by industry, national laboratories, and universities to develop cladding and fuel materials for lightwater reactors (LWRs) that are more accident tolerant. The near-term approach has been to develop coatings for Zr-alloys that would provide additional safety and operational margin by virtue of its excellent corrosion/oxidation resistance at both normal and accident conditions. The designs being considered for implementation by major nuclear fuel suppliers include a thin Cr or a ceramic coating on the conventional LWR fuel cladding. For improved economics, the industries are also considering ATF coated cladding with high enrichment fuel (up to 8%) to achieve high burnup (> 75 GWd/MTU). While the development of ATF concepts (i.e., the front end of the fuel cycle), including coated claddings and doped fuels have progressed at an accelerated pace, relatively less attention has been devoted to the used fuel disposition of ATF fuels (i.e., the backend of the fuel cycle). For accelerated deployment of the ATF designs in the current LWR fleet, it is necessary to investigate technical aspects of the ATF used nuclear fuel (UNF) management in transportation, storage, and disposal. This presentation will provide a brief overview of state-of-the-art ATF developments and list out potential considerations to apply the fuels into back-end fuel cycle. New test plan should be planned to compare the characteristics of current LWR used nuclear fuels with those of the new fuel designs. For example, research focus can be understanding of ATF used fuel particulate size and quantity (at high burnup condition) and mechanical integrity of coated cladding under normal and off-normal conditions during transportation and long-term storage. Finally, the impacts of CRUD on the new fuel cladding, increased container weight, temperature, and radiation level to the back-end fuel cycle activities need to be investigated.