In 2012, POSIVA selected a bentonite-based (montmorillonite) block/pellet as the backfilling solution for the deposition tunnel in the application for a construction license for the deep geological repository of high-level radioactive waste in Finland. However, in the license application (i.e. SC-OLA) for the operation submitted to the Finnish Government in 2021, the design for backfilling was changed to a granular mixture consisting of bentonite (smectite) pellets crushed to various sizes, based on NAGRA’s buffer solution. In this study, as part of the preliminary design of the deep geological repository system in Korea, we reviewed history and its rationale for the design change of Finland’s deposition tunnel backfilling solution. After the construction license was granted by the Finnish Government in 2015, POSIVA conducted various lab- and full-scale in-situ tests to evaluate the producibility and performance of two design alternatives (i.e. block/pellet type and granular type) for backfilling. Principal demonstration tests and their results are summarized as follows: (a) Manufacturing of blocks using three types of materials (Friedland, IBeco RWC, and MX-80): Cracking and jointing under higher pressing loads were found. Despite adjusting the pressing process, similar phenomena were observed. (b) 1:6 scale experiment: Confirmation of density difference inhomogeneity due to the swelling of block/pellet backfill and void filling due to swelling behavior into the mass loss area of block/pellet. (c) FISST (Full-Scale In situ system Test): Identification of technical unfeasibility due to the inefficient (too manual) installation process of blocks/pellets and development of an efficient granular in-situ backfilling solution to resolve the disadvantage. (d) LUCOEX-FE (Large Underground Concept Experiments – Full-scale Emplacement) experiment: Confirmation of dense/homogeneous constructability and performance of granular backfilling solution. In conclusion, the simplified granular backfill system is more feasible compared to the block/ pellet system from the perspective of handling, production, installation, performance, and quality control. It is presumed that various experimental and engineering researches should be preceded reflecting specific disposal conditions even though these results are expected to be applied as key data and/or insights for selecting the backfilling solution in the domestic deep geological repository.

The high-level radioactive waste repository must ensure its performance for a long period of time enough to sufficiently reduce the potential risk of the waste, and for this purpose, multibarrier systems consisting of engineered and natural barrier systems are applied. If waste nuclides leak, the dominating mechanisms facilitating their movement toward human habitats include advection, dispersion and diffusion along groundwater flows. Therefore, it is of great importance to accurately assess the hydrogeological and geochemical characteristics of the host rock because it acts as a flow medium. Normally, borehole investigations were used to evaluate the characteristics and the use of multi-packer system is more efficient and economical compared to standpipes, as it divides a single borehole into multiple sections by installing multiple packers. For effective analyses and groundwater sampling, the entire system is designed by preselecting sections where groundwater flow is clearly remarkable. The selection is based on the analyses of various borehole and rock core logging data. Generally, sections with a high frequency of joints and evident water flow are chosen. Analyzing the logging data, which can be considered continuous, gives several local points where the results exhibit significant local changes. These clear deviations can be considered outliers within the data set, and machine learning algorithms have been frequently applied to classify them. The algorithms applied in this study include DBSCAN (density based spatial clustering of application with noise), OCSVM (one class support vector method), KNN (K nearest neighbor), and isolation forest, of which are widely used in many applications. This paper aims to evaluate the applicability of the aforementioned four algorithms to the design of multi-packer system. The data used for this evaluation were obtained from DB-2 borehole logging data, which is a deep borehole locates near KURT.

Even though a huge amount of spent nuclear fuels are accumulated at each nuclear power plant site in Korea, our government has not yet started to select a final disposal site, which might require more than several km2 surface area. According to the second national plan for the management of high-level radioactive waste, the reference geological disposal concept followed the Finnish concept based on KBS-3 type. However, the second national plan also mentioned that it was necessary to develop the technical alternatives. Considering the limited area of the Korean peninsula, the authors had developed an alternative disposal concepts for spent nuclear fuels in order to enhance the disposal density since 2021. Among ten disposal concepts shown in the literature published in 2000’s, we narrowed them to four concepts by international experiences and expert judgements. Assuming 10,000 t of CANDU spent nuclear fuels (SNF), we designed the engineered barriers for each alternative disposal concept. That is, using a KURT geological conditions, the engineered barrier systems (EBS) for the following four alternative concepts were proposed: ① mined deep borehole matrix, ② sub-seabed disposal, ③ deep borehole disposal, and ④ multi-level dispoal. The quantitative data of each design such as foot prints, safety factors, economical factors are produced from the conceptual designs of the engineered barriers. Five evaluation criteria (public acceptance, safety, cost, technology readiness level, environmental friendliness) were chosen for the comparison of alternatives, and supporting indicators that can be evaluated quantitatively were derived. The AHP with domestic experts was applied to the comparison of alternatives. The twolevel disposal was proposed as the most appropriate alternative for the enhancement of disposal efficiency by the experts. If perspectives changes, the other alternatives would be preferred. Three kinds of the two-level disposal of CANDU SNF were compared. It was decided to dispose of all the CANDU spent nuclear fuels into the disposal holes in the lower-level disposal tunnels because total footprint of the disposal system for CANDU SNF was much smaller than that for PWR SNF. Currently, we reviewed the performance criteria related to the disposal canister and the buffer and designed the EBS for CANDU SNF. With the design, safety assessment and cost estimates for the alternative disposal system will be carried out next year.

The thermal evaluations for the conceptual design of the deep geological repository considering the improved modeling of the spent fuel decay heat were conducted using COMSOL Multiphysics computational program. The maximum temperature at the surface of a disposal canister for the technical design requirement should not exceed 100°C. However, the peak temperature at the canister surface should not exceed 95°C considering the safety margin of 5°C due to several uncertainties. All thermal evaluations were based on the time-dependent simulation from the emplacement time of the canister to 100,000 years later. In particular, the heat source condition was set to the decay heat rate and axial decay heat profile of the PLUS7 fuel with 4.0wt% U-235 and 45 GWD/MTU. The thermal properties of the granitic rock in South Korea were applied to the host rock region. For the reference design case, the cooling time of the SNF was set to 40 years, the distance between the deposition holes 8 meters and that between the deposition tunnels 30 meters. However, the peak temperature at the canister surface at 10 years was 95.979°C greater than 95°C. This design did not meet the thermal safety requirement and needed to be modified. For the first modified case, when the distance between the deposition tunnels was set to 30 meters, three cooling time cases of 40, 50 and 60 years and five distances of 6, 7, 8, 9 and 10 meters between the deposition holes were considered. The design with the distances of 9 and 10 meters between the deposition holes for the cooling time of 40 years and all five distances for 50 and 60 years were less than 95°C. For the second modified case, when the distance between the deposition holes was set to 8 meters, three cooling time cases of 40, 50 and 60 years and five distances of 20, 25, 30, 35 and 40 meters between the deposition tunnels were considered. The design with the distances of 35 and 40 meters between the deposition tunnels for the cooling time of 40 years, the distances of 25, 30, 35 and 40 meters for 50 years and all five distances for 60 years were less than 95°C. As a result, the peak temperature at the canister surface decreased as the cooling time and the distance between the deposition holes and the tunnels increased.

Spent nuclear fuel management is a high-priority issue in South Korea, and addressing it is crucial for the country’s long-term energy sustainability. The KORAD (Korea Radioactive Waste Agency) is leading a comprehensive, long-term project to develop a safe and effective deep geological repository for spent nuclear fuel disposal. Within this framework, we have three primary objectives in this work. First, we conducted statistical analysis to assess the inventory of spent nuclear fuel in South Korea as of 2021. We also projected future generation rates of spent nuclear fuels to identify what we refer to as reference spent nuclear fuels. These reference spent nuclear fuels will be used as the design basis spent fuels for evaluating the safety of the repository. Specifically, we identified four types of design basis reference spent nuclear fuels: high and low burnup from PLUS7 (with a 16×16 array) and ACE7 (with a 17×17 array) assemblies. Second, we analyzed radioactive nuclides’ inventory, activities, and decay heats, extending up to a million years after reactor discharge for these reference spent nuclear fuels. This analysis was performed using SCALE/TRITON to generate the burnup libraries and SCALE/ORIGEN for source term evaluation. Third, to assess the safety resulted from potential radioactive nuclides’ release from the disposal canister in future work, we selected safety-related radionuclides based on the ALI (Annual Limit of Intake) specified in Annex 3 of the 2019-10 notification by the NSSC (Nuclear Safety and Security Commission). Conservative assumptions were made regarding annual water intake by humans, canister design lifetime, and aquifer flow rates. A safety margin of 10-3 of the ALI was applied. We selected 56 radionuclides that exceed the intake limits and have half-lives longer than one year as the safety-related radionuclides. However, it is crucial to note that our selection criteria focused on ALI and half-lives. It did not include other essential factors such as solubility limits, distribution coefficients, and leakage processes. So, some of these nuclides can be removed in a specific analysis area depending on their properties.

The design and fabrication of suitable waste forms with high thermal and structural stability are essential for the safe management and disposal of radioactive wastes. In particular, the thermal properties and temperature distribution of waste form containing high heat-generating nuclides such as Cs and Sr can be used to evaluate its thermal stability, but also provide useful information for the design of canisters, storage systems, and repositories. In this study, a new program code-based thermal analysis framework has been developed to facilitate the characterization, design, and optimization of the waste form. Matlab was used as a software development platform because it provides powerful mathematical computation and visualization components such as the partial differential equation (PDE) toolbox for solving heat transfer problems using finite element method, the App Designer for graphical user interface (GUI), and the MATLAB Compiler for sharing MATLAB programs as standalone applications and web applications. The thermal analysis results such as temperature distribution, heat flux, maximum/ minimum temperature, and centerline/surface temperature profile are visualized with graphs and tables. To evaluate the effectiveness of the developed program, several design and optimization studies were carried out for the SrTiO3 waste form, selected as a stable form of strontium nuclide.

The PRIDE scale mechanical decladder is decladding apparatus for separating and recovering fuel material and cladding hull by horizontally slitting rod-cut. In order to enhance mechanical decladdng efficiency, the main requirements were considered as follows. Decladding of the fuel rods may be performed by rotation of three circular cutting blades inserted among the rollers arranged at 120° portion. In a mechanical decladder, a slitting assembly as a unit for slitting the cladding tube may include cutting blades for slitting and rollers for guiding extrusion of the cladding tube. Rotation of the cutting blades may be caused by the fuel rods being extruded from a plurality of rollers. Slitting intervals of rod-cuts having different diameters may be controlled by adding or removing a spacing plate between the cutting blade and a ranch bolt for fixing the slitting blade to the slitting assembly. An extrusion velocity with respect to the fuel rods may be controlled by a hydraulic pressure applied to the fuel rods. A force for cutting the fuel rods may be adjusted by controlling steel plates. Forces applied to a plurality of rollers may be generated by the hydraulic cylinder. The hydraulic pressure may be controlled by hydraulic pressure controller. The PRIDE scale mechanical decladder mainly consists of auto feeding module, hydraulic cylinder module and blade module. A load cell was installed between the hydraulic cylinder and the extrusion pin to measure the decladding force and slitting velocity, and a data acquisition system capable of obtaining data by using the RSC 232 was constructed. Also, the control panel can control the forward and backward movement of the extrusion pin, the hydraulic flow rate, and the hydraulic velocity. In the mechanical decladding test, 40 pieces of simulated rod-cuts were loaded in two auto feeding basket and slit by utilizing the 3-CUT blade modules in the housing, and hulls and simulated pellets were collected in the collection container. As a result, 80 pieces of simulated rodcut (brass pellets + Zry4 tube) were slit continuously without any problem. About 35 min was required to slit 80 rod-cuts and average decladding force was 260 kg. The decladding force of the ceramic simulated rod-cuts (castable) requires 25 kg less force than the brass pellets. Therefore, it is estimated that the spent fuel rod-cut can be fully split into three pieces using the mechanical decladder.

As part of strengthening pyro safety measures, the Korea Atomic Energy Research Institute is developing LIBS (Laser-Induced Breakdown Spectroscopy) application technology to analyze molten salt components in electrolytic recovery device in real time. LIBS performs qualitative and quantitative analysis by analyzing the spectrum of energy emitted by atomizing and ionizing elements on the surface of a salt sample with a high-focused laser. Since salt easily corrodes metal, it must be managed in an environment with a dew point of -40°C or lower. In this study, we designed and manufactured a device that places a rod-type sampling stick on a mounting base, automatically moves it to the optimal measurement position for LIBS, and retrieves the sample. Its characteristics are as follows. First, LIBS is stationary and does not move. Second, the sample stick is placed on a mounting base and can rotate 360 degrees. Third, according to the command, the sample stick automatically moves to the optimal measurement position of LIBS with three degrees of freedom (X, Y, Z). Fourth, the salt attached to the sampling stick is recovered for chemical analysis by driving the gripper mounted at the bottom of the Z axis, Z axis, and rotation axis (R). The X, Y, and Z movement distances of this device are each 100 mm, rotation is 360 degrees, grip stroke is 50 mm, and position accuracy is ±20 m. Once the performance test of the automated salt sample analysis device is completed, it will be installed in a dry room with a dew point of - 40°C or lower. Samples will be collected remotely in connection with the electrolytic recovery device and gantry robot built in the dry room. We plan to conduct experiments to seat the sample stick. Ultimately, we plan to conduct comprehensive experiments in conjunction with LIBS.

For efficient design and manufacture of PWR spent fuel burnup detector, data simulated with various condition of spent fuel in the NPP storage pool is required. In this paper, to derive performance requirements of spent fuel burnup detector for neutron flux and dose rates were evaluated at various distances from CE16 and WH17 types of fuel, representatively. The evaluation was performed by the following steps. First, the specifications of the spent fuel, such as enrichment, burnup, cooling time, and fuel type, were analyzed to find the conditions that emit maximum radioactivity. Second, gamma and neutron source terms of spent fuel were analyzed. The gamma source terms by actinides and fission products and neutron source terms by spontaneous and (α, n) reactions were calculated by SCALE6 ORIGAMI module. Third, simulation input data and model were applied to the evaluation. The material composition and dose conversion factor were referred as PNNL-15870 and ICRP-74 data, respectively and dose rates were displayed with the MCNP output data. It was assumed that there was only one fuel modeled by MCNP 6.2 code in pool. The evaluation positions for each distance were selected as 5 cm, 10 cm, 25 cm, 50 cm, and 1 m apart from the side of fuel, respectively. Fourth, neutron flux and dose rates were evaluated at distance from each fuel type by MCNP 6.2 code. For WH 17 types with a 50 GWd/MTU burnup from 5 cm distance close to fuel, the maximum neutron flux, gamma dose rates and neutron dose rates are evaluated as 1.01×105 neutrons/sec, 1.41×105 mSv/hr and 1.61×101 mSv/hr, respectively. The flux and dose rate of WH type were evaluated to be larger than those of CE type by difference in number of fuel rods. The relative error for result was less than 3~7% on average secured the reliability. It is expected that the simulated data in this paper could contribute to accumulate the basic data required to derive performance requirements of spent fuel burnup detector.

Most Small Modular Reactors currently under development are pursuing designs that can demonstrate flexibility in terms of construction and operation, and seeking to adopt innovative technologies to implement them, which is a very big challenge not only from the developer’s perspective but also from the regulator’s perspective. For the successful development of SMRs, it is necessary to move away from the existing prescriptive regulatory approach and exercise regulatory flexibility to sufficiently reflect design characteristics. The reason why SMR development is actively progressing around the world is to overcome the limitations of existing Light Water Reactors. Licensing advanced reactors such as MSR, VHTR, and microreactors requires a different approach from the existing conservative regulatory framework, taking into account changing regulatory environment. With the development of information technology and artificial intelligence, new types of threats are emerging, most of which are related to nuclear security. The IAEA, as well as leading countries such as the United States and Canada, require that safety, security, and safety measures be reflected in the early stages of design (Safety, Security, Safeguards by Design) and should be applied in the regulatory process. In addition, it is recommended to design a system that can achieve synergy effects by identifying in potential issues that may cause regulatory interference between safety and security (SSI, Safety-Security Interface). The competitiveness of SMR in the international market will be highly dependent on the degree to reflect the importing countries’ requirements. Since most SMRs currently under development do not have significant differences in safety goals, multi-purpose usability, etc., it is necessary to faithfully reflect the environmental factors necessary for SMR operation in the adopting country in the design. The major issues expected are mostly nuclear security-related rather than safety, and the physical characteristics of the site including the geopolitical conditions of the importing country must be taken into consideration. In this paper, the necessity of SMR nuclear security regulation and the way how to reflect the Security by Design on SMR will be presented.

The equivalent static load for non-structural elements has a limitation in that the sloshing effect and the interaction between the fluid and the water tank cannot be considered. In this study, the equations to evaluate the impulse and convective components in the design codes and previous research were compared with the shaking table test results of a rectangular water tank with flexible wall panels. The conclusions of this study can be summarized as follows: (1) It was observed that the natural periods of the impulsive component according to ACI 350.3 were longer than system identification results. Thus, ACI 350.3 may underestimate the earthquake load in the case of water tanks with flexible walls. (2) In the case of water tanks with flexible walls, the side walls deform due to bending of the front and back walls. When such three-dimensional fluid-structure interaction was included, the natural period of the impulsive component became similar to the experimental results. (3) When a detailed finite element (FE) model of the water tank was unavailable, the assumption   could be used, resulting in a reasonably conservative design earthquake load.

MicroRNAs (miRNAs) are emerging materials as ideal biomarkers for noninvasive cancer detection in the early phase. In this article, a simple and label-free electrochemical miRNA biosensor was developed. A single-stranded DNA (ss-DNA) probes were successfully mapped to f-MWCNT and hybridized with the target miR-141 sequence. The optimum peak points of the obtained hybridization were determined using Cyclic Voltammetry (CV) and Differential Pulse Voltammetry (DPV) methods. Significant peaks were observed in the results, depending on miR-141 at different concentrations. The linear relationship (ν) between redox peak currents (Ip) and scanning rate indicated that electron transfer (ET) between miR-141 and the electrode surface was accomplished successfully. In DPV measurements, miR-141 was measured with a low detection limit (LOD) in the 1.3–12 nM concentration range, and the LOD and limit of quantification (LOQ) results were found to be 3 and 9.1 pM, respectively. Besides, selectivity test was investigated for the biosensor using different target analytes and a significant difference in value was observed between the peak currents of miR-141, and other target molecules. This developed strategy has been found to detect miR-141 sensitively, selectively and without tags, and its integration into mobile devices has been successfully carried out.

이 연구는 일본 규슈 북쪽 서단(西端)에 위치한 작은 도시지만, 제1,2 차 세계대전에서 구(舊)일본 해군의 군사적 요충지 역할을 맡은 군항도시 사세보(佐世保)가 관광도시로 전환되는 그 과정의 설계와 실천 과정에 기반을 두고 있다. 1983년 개장한 나가사키(長崎) ‘오란다무라(Holland Village)’는 지역경제에 많은 파급효과를 가져왔고, 1992년 개장한 ‘하우 스텐보스(Huis Ten Bosch)’는 관광도시로의 전환에 있어 ‘결정체’와 같 은 역할을 했다. 오랜 기간 군항도시의 역할이 지속된 사세보는 전전기 (戦前期)와 전후기(戦後期)를 거치며 어두운 전쟁의 체험과 역사적 흔적 을 남겼다. 하지만 사세보 시민들이 염원하던 ‘평화산업항만도시’로의 전 환에 있어 결정체 역할을 한 ‘하우스텐보스’의 개장은 마치 어두운 전쟁 의 역사의 흔적을 지우듯 나가사키의 역사와 문화라는 옷을 입혀 그 부 지위에 자리 잡았다. 하우스텐보스가 조성된 부지는 전전기에는 해군병 교 하리오(針尾)분교가, 전후기에는 인양자들의 숙소이자 인양 원호국이 자리잡고 있었던 곳이었다. 이처럼 사세보의 어두운 역사의 땅 위에 하 우스텐보스가 지어지며 군항도시에서 관광도시로 전환된 과정을 흥미롭 게 살펴보고자 한다.

With the advancement of optical design and manufacturing technology, optical components have found diverse applications, spanning from semiconductors to the aerospace industry. A reflective mirror is a basic component in optics and plays a crucial role as the medium to reflect light. In this paper, a large mirror with a 700mm diameter was designed as a primary mirror using fused silica. The rear side of the mirror was subdivided into several equal angles, and neighboring vertices on the circumference were connected to establish a polygon. Accordingly, the geometric shapes of triangle, square, pentagon, and hexagon were formed. Furthermore, the mirror structure was strengthened by employing straight lines passing the vertices and the center of the circle. Based on the finite element analysis, deformations of the mirrors caused by the gravitational force were evaluated. Weight and deformation of the mirror structures were compared and analyzed to find a proper structure to reduce weight and deformation. This paper, therefore, presents a structural solution aimed at reducing the weight and deformation of a large aperture mirror induced by gravitational forces, thereby suggesting a geometric shape based structure to reduce surface deformation of a mirror.

Recently, many countries are performing researchs about weapon systems or communication systems using laser. Because laser weapons are relatively easy to use and can exert powerful energy with high efficiency versus cost. Also laser communication systems has many advantages compared to RF communication systems, such as big data transmission through giga-bit communication, high security and so on. In these systems, one of highly important key components is Fast Steering Mirror(FSM) to control the laser beam precisely and accurately. Therefore, in this paper, we perform static and dynamic analysis to predict performances of Fast Steering Mirror using voice coil actuators. Also we manufacture prototype of FSM on the basis of static and dynamic analysis results, and perform the performance test about four items. As a result, we lay a foundation for research about FSM and laser systems, and expact to contribute improvement of performances of systems using laser.

In this study, In this study, structural analysis of a fuel tank for an SUV (sports utility vehicle) was performed for crack prevention design. Reservoir tank analysis was conducted for crack prevention design, and improvement measures for weak areas were discovered and reflected in the design. Pressure analysis was performed on the existing model to analyze weak areas. As a result of analysis through various design changes, it was found that the strength problem of the reservoir tank was due to the discontinuity of the rib inside the tank, and to improve this, it was necessary to minimize the discontinuity section.

In general, a large mirror without weight reduction in large optical or space telescope systems can increase the system’s weight or lead to significant deformation of the mirror surface. Thus, it is imperative to pursue lightweight design strategies. In this paper, the structure design of a spherical mirror, a diameter of 600mm and a mirror radius of 2,000mm, was investigated to reduce weight and minimize deformation. To establish load paths for internal and external loads, stiffeners were added across the lateral supports. This approach effectively reduced both weight and deformation caused by gravity. Weight reduction and reduction percentages were quantified, and the mirror deformation was evaluated by using finite element analysis (FEA). The proposed structures were compared with honeycomb structures for weight reduction. This evaluation allowed to assess the deformation characteristics and the potential advantages of the proposed structures for lightweight mirrors.

목적 : 본 연구는 시뮬레이터 훈련이 척수손상 환자의 운전시뮬레이터 도로 주행시간과 주행조작 능력에 미치는 영향을 알아보고자 하였다. 연구방법 : 본 연구는 단일사례실험연구 AB설계로 진행되었으며, 기초선 3회기, 중재기 10회기를 적용하였다. 기초선 3회, 중재기 5회 도로주행코스(중급) 수행 시 운전시간, 주행조작 능력 자료가 수집되었으며, 연구결과 분석을 위하여 시각적 분석 방법과 평균 ±2*표준 편차를 사용한 양적 분석을 동시에 사용하였다. 결과 : 연구대상자 3명의 총 주행시간은 기초선 A보다 중재기 B에서 3분 내외의 감소하였고, 세 명 모두에서 통계학적으 로 의미있는 주행 능력의 향상이 확인되었다. 주행조작능력 또한 오류 점수가 감소하였고, 첫 번째 참여자와 세 번째 참 여자의 경우 그 변화가 통계학적으로 유의하였다. 결론 : 본 연구에서 연구대상자의 총 주행시간 및 수행 오류의 감소가 확인되어 운전시뮬레이터 훈련의 효과가 있었다. 이 와 같은 결과는 운전시뮬레이터 훈련의 적용 가능성을 뒷받침 한다.