PURPOSES : Under the Traffic Safety Act, the installation and management of transportation facilities (facilities and attachments necessary for the operation of transportation, such as roads, railways, and terminals) must take necessary measures to ensure traffic safety, such as enhancing safety facilities. Recently, railway operators have graded the congestion level inside railway stations and vehicles, addressing safety and convenience issues arising from congestion and providing this information to users. However, for bus-related transportation facilities (such as bus stops, terminals, and transfer facilities), criteria and related research for assessing traffic congestion are lacking. Therefore, this study developed a model for the congestion risk factors of four bus-related transportation facilities and proposed criteria for classifying congestion risk levels. METHODS : This study involved selecting congestion risk influence variables for each traffic facility through field surveys, calculating congestion risk index values through evacuation and pedestrian simulations, and constructing a congestion risk influence model based on the ridge model. RESULTS : The factors influencing congestion were selected to include the number of people waiting, effective sidewalk width, and number of bus stops. As a result of developing congestion risk grades, the central bus stops were determined to be in a severe stage if the Average Waiting Time (AWT) was 2.7 or above. Roadside bus stops were considered severe at 4.2, underground metropolitan transit centers at 3.7, and bus terminals at 5.9 or above. CONCLUSIONS : This study can help establish a foundation for a safety management system for congested areas in transportation facilities. When the congestion risk prediction results correspond to cautionary or severe levels, measures that can reduce congestion risk must be applied to ensure the safety of road users.

자율주행차 상용화 시대를 가속화하기 위해 실제 도로에서 다양한 실증 프로젝트를 수행중이다. 그러나, 자율주행차와 비자율주행차 가 혼재된 혼합교통류 환경에서 발생할 수 있는 다양한 문제의 원인을 파악하고 선제적인 안전대책을 강구하는 노력은 미비한 실정이 다. 특히, 기존 비자율주행차 측면의 주행안전성을 고려하여 설계된 도로 시설 특성으로 인해 자율주행차의 주행안전성이 저하될 수 있다. 또한 기존 비자율주행차의 주행안전성을 저해함과 동시에 자율주행차의 주행안전성도 저해하는 도로 시설 특성이 존재할 가능 성이 있다. 본 연구에서는 상암 자율주행차 시범운행지구에서 수집된 automated vehicle data (AVD)를 활용하여 자율주행차와 비자율주 행차의 주행안전성을 평가하고 도로 시설 특성 측면의 영향요인을 도출하였다. 주행모드별 주행안전성 평가를 위해 autonomous emergency braking system (AEBS) 위험 이벤트 기반의 driving risk index (DRI)를 개발하였다. 구간별 DRI가 발생하지 않은 구간을 very good으로 정의하고 발생한 구간을 25 percentile로 구분하여 good, moderate, poor, very poor 등급으로 정의하여 총 5개의 등급으로 구분 하였다. 또한, 현장조사을 수행함으로써 구간별 포함되어 있는 도로 시설 특성을 수집하였다. 주행모드별 주행안전성에 영향을 미치는 도로 시설 특성을 도출하기 위해 이항로지스틱 회귀분석을 수행하였다. 종속 변수의 경우 DRI 기반 안전등급 중 poor 이상 등급을 1, 그 외의 등급을 0으로 정의하였으며, 독립변수의 경우 현장조사를 통해 수집된 교차로 유형, 차로 수, 차로 폭, 추가차로 유무, 차량 진행방향, 불법주정차 유무, 버스정류장 유무, 자전거 차로 유무에 대해 명목형 변수로 설정하였다. 도출된 주행모드별 주행안전성 영 향 요인을 검토하고 향후 자율주행차 시대에 대비하여 선제적으로 개선이 요구되는 도로 시설 특성을 도출하고 도로 운영성 및 효율 성, 안전성 측면의 개선 방향을 제시하였다.

본 연구는 지역 문화예술공연장 규모에 따른 서비스품질과 관람만족의 차이 분 석, 그리고 서비스품질과 관람만족의 관계를 규명하고자 진행되었다. 이를 위해 공연장 서비스품질(유형성, 신뢰성, 응답성, 확신성, 공감성)에 대한 공연장 규모 별 차이, 그리고 관람만족에 영향을 미치는 공연장 서비스품질 요인을 파악하고자 설문조사를 통하여 연구가설을 설정한 후 진행하였다. 연구지역은 광주광역시와 전라남도 소재에 있는 문화예술공연장(소극장, 중극장, 대극장)을 찾아온 관람객 을 대상으로 불성실한 답변을 제외한 총 313부를 유효 표본으로 활용하였다. 분석 결과, 첫째, 공연장 규모에 따른 서비스품질 차이분석 결과 응답성, 확신성, 공감 성에서 소극장이 차이를 보이고 있었다. 둘째, 공연장 규모에 따른 관람만족 차이 분석 결과 소극장에서 관람만족이 가장 높은 것으로 나타났다. 셋째, 공연장 서비 스품질의 공감성, 유형성, 응답성, 확신성은 관람만족에 긍정적 영향을 미치는 것 으로 나타났다. 본 연구 결과는 지역의 문화예술공연장 규모에 따른 서비스품질 중요요인을 추출했다는 점에서 학술적, 현장적 의미가 있다고 할 수 있다.

This article presents the crucial role played by the French underground research laboratory (URL) in initiating the deep geological repository project Cigéo. In January 2023, Andra finalized the license application for the initial construction of Cigéo. Depending on Government’s decision, the construction of Cigéo may be authorized around 2027. Cigéo is the result of a National program, launched in 1991, aiming to safely manage high-level and intermediate level long-lived radioactive wastes. This National program is based on four principles: 1) excellent science and technical knowledge, 2) safety and security as primary goals for waste management, 3) high requirements for environment protection, 4) transparent and openpublic exchanges preceding the democratic decisions and orientations by the Parliament. The research and development (R&D) activities carried out in the URL supported the design and the safety demonstration of the Cigéo project. Moreover, running the URL has provided an opportunity to gain practical experience with regard to the security of underground operations, assessment of environmental impacts, and involvement of the public in the preparation of decisions. The practices implemented have helped gradually build confidence in the Cigéo project.

본 연구의 목적은 노인장기요양보험법 관련 시설에 종사하는 요양보호사의 직무만족도와 자기효능감, 그리고 근무안정성과 의 관계를 고찰하는 것이며 특히 관련 시설 중 노인요양시설과 재가노인복지시설을 대표하는 주야간보호센터를 비교 분석하고 자 한다. 노인장기요양보험법 관련 시설 요양보호사의 근무안 정성을 제고하기 위하여 요양보호사의 전문성 향상 및 고용에 대한 안정성을 확보하기 위한 방안을 마련할 수 있는 구체적인 운영전략을 제시하고자 한다. 이를 위해 노인장기요양보험 관 련 시설별 요양보호사의 직무만족도와 근무안정성과의 관계에 서 자기효능감의 조절효과를 분석하기 위해 서울시 소재 노인 요양시설과 재가노인복지시설에서 근무하는 요양보호사 810명 을 대상으로 자기 기입식 설문조사를 실시하였다. 수집된 자료 를 기술통계, 상관관계분석 및 위계적 회귀분석을 실시한 결과 다음과 같은 결론이 도출되었다. 첫째, 노인요양시설의 경우 학 력, 직무만족도가 근무안정성에 영향을 미친 것으로 나타났으 므로, 학력별 근무안정성을 높이기 위한 노력과 직무만족도가 낮은 요양보호사들을 대상으로 근무안정성을 향상시키기 위한 개입 노력이 필요하다. 둘째, 재가노인복지시설과 노인요양시설 모두 직무만족도는 근무안정성에 정적인 영향을 미친 것으로 나타났으므로, 요양보호사의 근무안정성을 높이기 위해서는 직 무만족도를 향상시켜야 한다. 셋째, 시설 유형별 요양보호사의 직무만족도와 근무안정성과의 관계에서 자기효능감의 조절효과 가 나타났으므로 자기효능감을 향상시키기 위한 프로그램이 필 요하다. 본 연구의 결론을 토대로 노인장기요양 관련 시설별 요양보 호사들의 근무안정성을 높이기 위해 직무만족도 수준의 향상과 자기효능감을 증진하기 위한 실천적방안을 제언하였다.

Korea Atomic Energy Research Institute (“KAERI”) has been developing pyroprocess technology for the sustainable use of nuclear energy and radioactive waste reduction, and is conducting design studies for a Pyroprocess Commercializing Research Facility (PCRF). High-level radioactive materials such as spent nuclear fuel, which are handled in the hot cell of the PCRF, physically change materials directly or cause chemical changes through ionization or excitation depending on the energy and types of radiation. Therefore, all facilities, including process equipment and remote handling equipment, installed into the hot cell must be evaluated for radiation hardness to be maintained in the radiological environmfent so that processes can proceed throughout the design life of the facility. In addition, as the maintenance paradigm has recently shifted from corrective maintenance to predictive maintenance, it is necessary to know in advance the condition of the equipment or facility in the radiological environment. In this study, an analysis of the radiation environment of the hot cell in the PCRF was conducted through source term, and the radiological dose impact was evaluated through the results of irradiation experiments of major components by reference data. Then, the actual dose contribution was identified through dose assessment using the MCNP code based on the pyroprocess equipment, and the radiation hardness requirements for the facility and equipment in the hot cell were derived by the above results.

Korea Atomic Energy Research Institute (“KAERI”) has been developing various studies related to the nuclear fuel cycle. Among them, KAERI was focusing on the pyroprocess, which recycles some useful elements white reducing the volume and toxicity of spent nuclear fuel (SNF). Pyroprocess involves the handling of SNF, which cannot be handled directly by the facility worker. Therefore, SNF is handled and processed through remote handling device within a shielded facility such as a hot cell. Nuclear Facilities with such hot cells are called nuclear fuel cycle facilities, and unlike other facilities, heating, ventilating, and air conditioning (HVAC) system are particularly important in nuclear fuel cycle facilities to maintain the atmosphere in the hot cell and remove radioactive materials. In addition, due to the nature of the pyroprocess, which uses molten salt, corrosion is a problem in air atmosphere, so the process can only be carried out in an inert gas atmosphere. KAERI has a nuclear fuel cycle facility called the Irradiation Material Examination Facility (IMEF), and has built and operated the ACPF inside the IMEF, which operates an inert atmosphere hot cell for the demonstration of the pyroprocess. For efficient process development of the pyroprocess, it is necessary to put the developed equipment into the hot cell, which is a radiationcontrolled area, after sufficient verification in a mock-up facility. For this purpose, the ACPF mock-up facility, which simulates the system, space, and remote handling equipment of the ACPF, is operated separately in the general laboratory area. The inert gas conditioning system of the ACPF consists of very complex piping, blowers, and valves, requires special attention to maintenance. In addition, if there is a small leak in the piping within these valves or piping, radioactive materials can be directly exposed to facility workers, so continuous monitoring and maintenance are required to prevent accident. In this study, the applicability of acoustic emission technology and ultrasonic technology for leak detection in the inert gas conditioning system of ACPF mock-up facility was investigated. For this purpose, new bypass pipes and valves were installed in the existing system to simulate the leakage of pipes and valves. Acoustic emission sensors are attached directly to pipes or valves to detect signals, while ultrasonic sensors are installed at a distance to detect signals. The optimal parameters of each technology to effectively suppress background noise were derived and, and the feasibility of identifying normal and abnormal scenarios in the system was analyzed.

To secure approval for a decommissioning plan in Korea, it is essential to evaluate contamination dispersion through groundwater during the decommissioning process. To achieve this, licensees must assess the groundwater characteristics of the facility’s site and subsequently develop a groundwater flow model. It is worth noting that Combustible Radioactive Waste Treatment Facility (CRWTF) is characterized by their simplicity and absence of liquid radioactive waste generation. Given these facility characteristics, the groundwater flow model for CRWTF utilizes data from neighboring facilities, with the feasibility of using reference data substantiated through comparative analysis involving groundwater characteristic testing and on-site modeling. To enable a comparison between the actual site’s groundwater characteristics and the referenced modeling, two types of hydraulic constant characterization tests were conducted. First, hydraulic conductivity was determined through long-term pumping and recovery tests. The ‘Theis’ and ‘Cooper-Jacob’ equations, along with the ‘Theis recovery’ equation, were applied to calculate hydraulic conductivity, and the final result adopted the average of the calculated values. Secondly, a groundwater flow test was conducted to confirm the alignment between the main flow direction of the referenced model and the groundwater flow in the CRWTF, utilizing the particle tracking technique. The evaluation of hydraulic conductivity from the hydraulic constant test revealed that the measured value at the actual site was approximately 1.84 times higher than the modeled value. This variance is considered valid, taking into consideration the modeling’s calibration range and the fact that measurements were taken during a period characterized by wet conditions. Furthermore, a close correspondence was observed between the groundwater flow direction in the reference model (ranging from 90° to 170°) and the facility’s actual flow direction (ranging from 78° to 95°). The results of reference data for the CRWTF, based on the nearby facility’s model, were validated through the hydraulic properties test. Consequently, the modeling data can be employed for the demolition plan of CRWTF. It is also anticipated that these comparative analysis methods will be instrumental in shaping the groundwater investigation plans for facilities with characteristics similar to CRWTF.

The critical hazards generated from operation of a melting facility for metal radioactive waste are mainly assumed to be such as vapor explosion, ladle breakthrough and failure in the hot-cell or furnace chamber using remote equipment. In case of vapor explosion, material containing moisture and/or enclosed spaces may, due to rapid expansion of gases when heated, cause an explosion and/or violent boiling. The rapid expansion of gases may lead to ejection of molten radioactive metal from the furnace into the furnace hall. If there is a large amount of liquid the explosion may damage or destroy technical barriers such as facility walls. The consequences for the facility ranges from relatively mild to very severe depending on the force of the explosion as well as the type of waste being melted. Nonradiological consequences may be physical damage or destruction of equipment and facility barriers, such as walls. Due to the radiological consequences a longer operational shutdown would likely be required. Cleanup efforts would include cutting of solidified metal in a problematic radiological environment requiring use of remote technology before damage and repair requirements can be assessed. Even though there is a risk for direct physical harm to operators for example in the control room and hot-cell, this analysis focuses mainly on the radiological impact. The extent to which remote equipment could be used in the decontamination effort will largely determine the health consequences to the workers. It is reasonable to assume that there will be a need for workers to participate manually in the effort. Due to the potentially large dose rates and the physical environment, it is possible that the maximum allowable dose burden to a worker will be reached. No major consequence for the environment is expected as most of the radioactivity is bound to the material. In case of ladle breakthrough, a ladle breakthrough involves loss of containment of the melt due to damage of the ladle. This may be caused e.g. by increased wear due to overheating in the melt, or from physical factors such as mechanical stress and impact from the waste. A ladle breakthrough may lead to spread of molten metal in the furnace hall. Molten metal coming into contact with the surrounding cooling equipment may cause a steam explosion. The consequences of a ladle breakthrough will depend on the event sequence. The most severe is when the molten metal comes into contact with the cooling system causing a vapor explosion. The basic consequences are assumed to be similar to those of the vapor explosion above. While the ejection of molten metal is likely more local in the ladle breakthrough scenario, the consequences are judged to be similar. In case of failure in the hot-cell or furnace chamber using remote equipment, the loss of electric supply or technical failure in the furnace causes loss of power supply. If not remedied quickly, this could lead to that the melt solidifies. A melt that is solidified due to cooling after loss of power cannot be removed nor re-melted. This may occur especially fast if there is not melted material in the furnace. An unscheduled replacement of the refractory in the furnace would be required. It could be unknown to what degree remote equipment can be used to cut a solidified melt. It is therefore assumed that personnel may need to be employed. This event could not have any impact on environment

In light of recent significant seismic events in Korea and worldwide, there is an urgent need to reevaluate the adequacy of seismic assessments conducted during facility construction. This study reexamines the ongoing viability of the Safety Shutdown Earthquake (SSE) criteria assessment for the Combustible Radioactive Waste Treatment Facility (CRWTF) site at the Korea Atomic Energy Research Institute (KAERI), originally established in 1994. To validate the SSE assessment, we delineated 13 seismic structure zones within the Korean Peninsula and employed two distinct methodologies. Initially, we updated earthquake occurrence data from 1994 to the present year (2023) to assess changes in the site’s horizontal maximum earthquake acceleration (g). Subsequently, we conducted a comparative analysis using the same dataset, contrasting the outcomes derived from the existing distance attenuation equation with those from the most recent attenuation equations to evaluate the reliability of the applied attenuation model. The Safety Shutdown Earthquake (SSE) criterion of 0.2 g remains unexceeded, even when considering recent earthquake events since the original evaluation in 1994. Furthermore, when applying various assessment equations developed subsequently, the maximum value obtained from the previously utilized ‘Donvan and Bornstein’ attenuation equation is 0.1496 g, closely resembling the outcome derived from the recently employed ‘Lee’ reduction equation of 0.1451 g. The SSE criteria for CRWTF remain valid in the current context, even in light of recent seismic occurrences such as the 2016 Gyeongju earthquake. Additionally, the attenuation equation employed in the evaluation consistently yields conservative results when compared to methodologies used in recent assessments. Consequently, the existing SSE criteria remain valid at present. This study is expected to serve as a valuable reference for confirming the SSE criterion assessment of similarly constructed facilities within KAERI.

Plasma torch melting technology has been considered as a promising technology for treating or reducing the radioactive waste generated by nuclear power plants. In 2006, IAEA announced that the technology is able to treated regardless of the type of target wastes. Because of the advantage, many countries have been funding, researching and developing the treatment technology. In this study, oversea plasma torch melting facilities for radioactive wastes treatment are reviewed. Also, plasma torch melting facility developed by KHNP CRI is briefly introduced.

Recently, the nuclear decommissioning and environmental restoration industries has significantly attracted as a new industry field due to the decision to decommission the KORI#1 and WOLSONG #1 nuclear power plant. In order to dispose of the decommissioning radioactive wastes generated during nuclear decommissioning, proper analysis is required, and disposal decisions are determined based on the analysis results. When dismantling a nuclear power plant, a few thousand of tons decommissioning waste are produced, so these require analysis for proper disposal. Therefore, a radionuclide facility for decommissioning waste analysis is essential for the disposal of the large quantities of decommissioning waste generated during nuclear power plant decommissioning. Korea Research Institute of Decommissioning (KRID) was established radionuclide analysis facilities to address above issues and support nuclear power plant decommissioning projects. The plan is to perform classification by type and radionuclide for all waste produced during nuclear power plant decommissioning and to support the disposal of radioactive wastes. In addition, we plan to establish validation methods for samples where verification methods are not established, in order to conduct efficient analysis and management. In this presentation, we will introduce the radionuclide facility currently under construction at KRID and present the space design, equipment layout, and utilization plans.

The development of existing radioactive waste (RI waste) management technologies has been limited to processing techniques for volume reduction. However, this approach has limitations as it does not address issues that compromise the safety of RI waste management, such as the leakage of radioactive liquid, radiation exposure, fire hazards, and off-gas generation. RI waste comes in various forms of radioactive contamination levels, and the sources of waste generation are not fixed, making it challenging to apply conventional decommissioning and disposal techniques from nuclear power plants. This necessitates the development of new disposal facilities suitable for domestic use. Various methods have been considered for the solidification of RI waste, including cement solidification, paraffin solidification, and polymer solidification. Among these, the polymer solidification method is currently regarded as the most suitable material for RI waste immobilization, aiming to overcome the limitations of cement and paraffin solidification methods. Therefore, in this study, a conceptual design for a solidification system using polymer solidification was developed. Taking into account industrial applicability and process costs, a solidification system using epoxy resin was designed. The developed solidification system consists of a pre-treatment system (fine crush), solidification system, cladding system, and packing system. Each process is automated to enhance safety by minimizing user exposure to radioactive waste. The cladding system was designed to minimize defects in the solidified material. Based on the proposed conceptual design in this paper, we plan to proceed with the specific design phase and manufacture performance testing equipment based on the basic design.

Structural stability of a waste form can be provided by the waste form itself (steel components, etc.), by processing the waste to a stable form (solidification, etc.), or by emplacing the waste in a container or structure that provides stability (HICs or engineered structure, etc.). The waste or container should be resistant to degradation caused by radiation effects. In accordance with the requirements for the domestic waste acceptance criteria, irradiation testing of solidified waste forms containing spent resin should be conducted on specimens exposed to a dose of 1.0E+6 Gy and other material 1.0E+7 Gy. Expected cumulative dose over 300 years is about 1.770E+6 Gy for spent resin and 0.770E+6 Gy for dried concentrated waste generated from NPPs generally. According to NRC Waste Form Technical Position, to ensure that spent resins will not undergo adverse degradation effects from radiation, resins should not be generated having loadings that will produce greater than 1E+6 Gy total accumulated dose. If it necessary to load resins higher than 1E+6 Gy, it should be demonstrated that the resin will not undergo radiation degradation at the proposed higher loading. This is the recommended maximum activity level for organic resins based on evidence that while a measurable amount of damage to the resin will occur at 1E+6 Gy, the amount of damage will have negligible effect on disposal site safety. Cementitious materials are not affected by gamma radiation to in excess of 1E+6 Gy. Therefore, for cement-stabilized waste forms, irradiation qualification testing need not be conducted unless the waste forms contain spent resins or other organic media or the expected cumulative dose on waste forms containing other materials is greater than 1E+7 Gy. Testing should be performed on specimens exposed to IE+6 Gy or the expected maximum dose greater than 1E+6 Gy for waste forms that contain ion exchange resins or other organic media or the expected maximum dose greater than 1E+7 Gy for other waste forms. This is suggestion as a review result that requirement for irradiation testing of solidified waste forms has something to be revise in detail and definitively.

In Korea, most temporary storage facilities for spent nuclear fuel are nearing saturation. As an alternative to this, the 2nd basic plan for high-level radioactive waste management specified the operation plan of dry interim storage facility. Meanwhile, the NSSC No. 2021-19 stipulates that it is necessary to evaluate the possibility and potential effect of accident before operating interim storage facility. Therefore, this study analyzed the categories of accident scenarios that may occur in dry storage facility as part of prior research on this. We investigated the case of categorization of dry storage facility accident scenarios of IAEA, NRC, KAREI, and KINS. The IAEA presented accident scenarios that could occur in on-site dry storage facility operated with silo and cask method. NRC has classified accident scenarios in dry storage facility and estimated the probability of accidents for each. KAERI and KINS selected major accident scenarios and analyzed the processes for each, in preparation for the introduction of dry storage facility in Korea in the future. Overall, a total of 10 accident scenarios were considered, and the scenarios considered by each institution were different. Among 10 scenarios, cask drop and aircraft collision were included in the categorization of most institutions. The results of this study can be used as basic data for cataloging accidents subject to safety evaluation when introducing dry interim storage facility in Korea in the future.

South It is necessary to develop the future technologies to improve the sustainability and acceptability of nuclear power plants generation. Currently, our company is preparing to build the dry storage facility on-site in accordance with the basic plan for managing high-level radioactive waste announced by the government in 2021. However, studies on technologies for the volume reduction of spent nuclear fuel to increase the efficiency of on-site spent fuel dry storage facilities are very not enough. Accordingly, in this study, the storage efficiency and appropriateness for the SF volume reduction processing technologies such as SF oxide processing technology and consolidation technology are evaluated. Finally, the goal is to develop the optimized technologies to improve the storage efficiency of spent nuclear fuel. As a result in this study is followings. [Safety] After removing volatile fission products (Xe, Kr, I, etc.), Xe, Kr, etc. are removed during storage of the sintered structures. UO2 has a high melting point of approximately 1,000°C after cesium (Cs) has been removed, and heat can be removed by natural convection. [Economy]1999 DUPIC unit facility unit price reference, 2020 standard 328 $/kg estimated. A Comprehensive Approach Considering the Whole System is needed. Benefit from replacement and continuous operation of metal storage containers. Changes in economic efficiency obtained in conjunction with fluctuations in electricity prices and disposal. [Waste filter] A separated solidification facility high-level waste filter is required, and overseas outsourcing must be considered. [Waste cladding]. Cannot be accommodated in low-level disposal site. This reason is why the Ni nuclides occur to be in bulk. [Metal structural material] It is possible to reduce the initial volume by 7.6% or more when compressed or melted, but the technology needs to be advanced. [Oxide blocks] Larger size and density are expected to improve storage and disposal efficiency. [Facilities operation waste] Expected to be able to be disposed of at mid-to-low level decommissioning sites in Gyeongju city. [Solidified volatile nuclides and activated metals] Expected to improve storage efficiency when used volume is reduced and stored, such as outsourced reprocessing. [Oxide block] Radioactivity and decay heat are estimated to be reduced by half during oxide treatment. 75% reduction in volume and 40% reduction in storage area compared to used nuclear fuel before treatment. [Merits/Shortages] Improvement of storage and disposal efficiency empirical research such as large-capacity [real-scale] oxide block production is required. Oxide processing facilities are likely to be classified as post-use nuclear fuel processing facilities. It is determined that additional documents such as a Radiation Environmental Report (RER) must be submitted. Existence of possible external leaks of glass, highly mobile radionuclides from the point of view of nuclear criticality and heat removal. Acceptancy requirements of citizens in the process of creating additional sites for oxide treatment facilities. Considering social public opinion, it is necessary to secure the acceptability such as residents’ opinions convergence. Characteristics of high nuclear non-propagation compared to other processing technologies involving chemical processing. Also, Expectation of volume reduction effect for spent nuclear fuel itself. Volume reduction methods for solid waste and gaseous waste are required.

Safeguards systems and measures are determined through diversion scenario analysis based on the facility design information submitted to the IAEA when a new nuclear facility is introduced. While the concept of safeguards-by-design (SBD), which considers the safeguards from the design phase for a facility operator to minimize unplanned changes or disruption to facility operations as well as for the IAEA to increase the efficiency and effectiveness in safeguards implementation, has been emphasized for more than a decade, there is no practical tool or guidance on how to apply it. In this study, we develop a diversion path analysis tool and introduce how to apply SBD using it. A diversion path analysis tool was developed based on the elements that constitute diversion and the algorithm generated based on the initial information of facility and nuclear material flow. The results of utilizing the analysis tool depending on a different level of facility information and the safeguards set-ups were compared through examples. Taking a typical light water reactor as an example, the test analyzed the automatic generation of dedicated routes, configuration of safeguards measures, and diversion path analysis. Through this, the application and limitations of the analysis tool are discussed, and ideas for utilization according to the SBD concept and necessary regulatory guidance are proposed. The results of this study are expected to be directly utilized to domestic nuclear control during the regulation process for a construction of new nuclear power systems, and furthermore, to enhance national credibility in the engagement with the IAEA for implementation of safeguards.

The Republic of Korea (ROK), as a member state of the IAEA, is operating the State’s System of Accounting for and Control (SSAC) and conducting independent national inspections. Furthermore, an evaluation methodology for the material unaccounted for (MUF) is being developed in ROK to enhance capabilities of national inspection. Generally, physical and chemical changes of nuclear material are unavoidable due to the operating system and structure of facilities, an accumulation of material unaccounted for (MUF) has been issued. IAEA developed statistical MUF evaluation method that can be applied to all facilities around the world and it mainly focuses on the diversion detection of nuclear materials in facilities. However, in terms of the national safeguard inspection, an evaluation of accountancy in facilities is additionally needed. Therefore, in this research, a new approach to MUF evaluation is suggested, based on the Guide to the Expression of Uncertainty in Measurement (GUM) that an evaluation of measurement uncertainty factors is straightforward. A hypothetical list of inventory items (LII) which has 6,118 items at the beginning and end of the material balance period, along with 360 inflow and outflow nuclear material items at a virtual fuel fabrication plant was employed for both the conventional IAEA MUF evaluation method and the proposed GUM-based method. To calculate the measurement uncertainty, it was assumed that an electronic balance, gravimetry, and a thermal ionization mass spectrometer were used for a measurement of the mass, concentration, and enrichment of 235U, respectively. Additionally, it was considered that independent and correlated uncertainty factors were defined as random factors and systematic factors for the ease of uncertainty propagation by the GUM. The total MUF uncertainties of IAEA (σMUF) and GUM (uMUF) method were 37.951 and 36.692 kg, respectively, under the aforementioned assumptions. The difference is low, it was demonstrated that the GUM method is applicable to the MUF evaluation. The IAEA method demonstrated its applicability to all nuclear facilities, but its calculated errors exhibited low traceability due to its simplification. In contrast, the calculated uncertainty based on the GUM method exhibited high reliability and traceability, as it allows for individual management of measurement uncertainty based on the facility’s accounting information. Consequently, the application of the GUM approach could offer more benefits than the conventional IAEA method in cases of national safeguard inspections where factor analysis is required for MUF assessment.

A seal is one of the primary means of safeguards along with surveillance. The International Atomic Energy Agency (IAEA) uses various types of seals to verify the diversion of nuclear materials and is developing new seals according to the development of technology. Independent of the IAEA, ROK uses national safeguards seals for state-level regulation. A national safeguards inspector binds the nuclear material storage by combining a seal with a metal wire and checks the serial number of the RFID chip inserted in the seal with a reader. The Wolsong spent fuel dry storage facility has 14 modules, each with 24 seals, and thus a maximum of 336 national seals will be installed. Although dependent on the sealing method, it takes about 5 minutes to verify one seal. As such, a considerable workforce is required for verification, and both the IAEA and the ROK are currently conducting random inspections. In addition, there are cases where verification is impossible because old seals are damaged due to harsh environments and long exchange periods. Therefore, in this study, we analyzed cases in areas where sealing technology has been developed to improve the problems of the existing national safeguards seals. And we proposed a method for improving national seals by finding requirements of seals considering spent fuel dry storage facility characteristics. In international logistics, sealing is essential in product transport verification, terrorism prevention, and tariff imposition. Accordingly, the field of container sealing has been extensively developed, and the International Organization for Standardization (ISO) has regulated the mechanical requirements of the seal as ISO 17712 and the electronic requirements as ISO 18185. Mechanical seals include metal and plastic seals and metal seals include bolt seals, ball seals, and cable seals. In addition, there are various electronic seals, such as radio frequency identification (RFID), near field communication (NFC), infrared (IR). Recently, there has been a trend to use active seals that have a built-in battery and can implement various additional functions. Among the various seals, the main requirements for selecting seals suitable for dry storage facilities are as follows. First, use of a sealing tube longer than 10m should be possible. Second, it should have corrosion resistance so that it can be used for more than five years in the coastal area. Third, it must be a passive seal without a power supply. Fourth, it should not be overly costly. Finally, the seal verification time should be short. As a seal that satisfies these requirements, an electronic seal with application of the passive RFID method to the mechanical form of a metal cable seal is suitable. Since it is not an active seal, it is difficult to determine the time of breakage. Therefore, designing the seal such that the RFID is also damaged when the metal seal is broken will be helpful for verification. In this study, the requirements for national safeguards seals in dry storage facilities were defined, and measures to improve the existing national seals were studied. Field applicability will be evaluated through future sealing device design and demonstration tests.