안전은 해군과 같은 위험성이 높은 환경에서 활동하는 조직에게 필수적이다. 효과적인 안전관리는 지속적인 개선과 보완을 통 해 유지되어야 하며, PDCA cycle을 활용하는 것이 일반적이다. 하지만 해군에서는 안전 규정 강화와 교육 확대에도 불구하고 안전사고가 지속적으로 발생하고 있다. 이는 안전사고 분석 및 분류 시스템 개선의 필요성을 보여준다. 본 연구에서는 해군 안전사고 분류체계를 분 석하고 문제점을 파악하여 효과적인 분류체계를 구축하는데 중점을 두었다. 이를 통해 안전사고 결과를 데이터화하고, 사고의 근본 원인 을 파악하며, 중장기적인 안전관리 정책 수립에 기여할 수 있도록 12자리의 해군 안전사고 분류 코드를 제안하였다.

In the military, ammunition and explosives stored and managed can cause serious damage if mishandled, thus securing safety through the utilization of ammunition reliability data is necessary. In this study, exploratory data analysis of ammunition inspection records data is conducted to extract reliability information of stored ammunition and to predict the ammunition condition code, which represents the lifespan information of the ammunition. This study consists of three stages: ammunition inspection record data collection and preprocessing, exploratory data analysis, and classification of ammunition condition codes. For the classification of ammunition condition codes, five models based on boosting algorithms are employed (AdaBoost, GBM, XGBoost, LightGBM, CatBoost). The most superior model is selected based on the performance metrics of the model, including Accuracy, Precision, Recall, and F1-score. The ammunition in this study was primarily produced from the 1980s to the 1990s, with a trend of increased inspection volume in the early stages of production and around 30 years after production. Pre-issue inspections (PII) were predominantly conducted, and there was a tendency for the grade of ammunition condition codes to decrease as the storage period increased. The classification of ammunition condition codes showed that the CatBoost model exhibited the most superior performance, with an Accuracy of 93% and an F1-score of 93%. This study emphasizes the safety and reliability of ammunition and proposes a model for classifying ammunition condition codes by analyzing ammunition inspection record data. This model can serve as a tool to assist ammunition inspectors and is expected to enhance not only the safety of ammunition but also the efficiency of ammunition storage management.

The HADES (High-level rAdiowaste Disposal Evaluation Simulator) was developed by the Nuclear Fuel Cycle & Nonproliferation (NFC) laboratory at Seoul National University (SNU), based on the MOOSE Framework developed by the Idaho National Laboratory (INL). As an application of the MOOSE Framework, the HADES incorporates not only basic MOOSE functions, such as multi-physics analysis using Finite Element Method (FEM) and various solvers, but also additional functions for estimating the performance assessment of Deep Geological Repositories (DGR). However, since the MOOSE Framework does not have complex mesh generation and data analyzing capabilities, the HADES has been developed to incorporate these missing functions. In this study, although the Gmsh, finite element mesh generation software, and Paraview, finite element analysis software, were used, other applications can be utilized as well. The objectives of HADES are as follows: (i) assessment of the performance of a Spent Nuclear Fuel (SNF) disposal system concerning Thermal-Hydraulic-Mechanical-Chemical (THMC) aspects; (ii) Evaluation of the integrity of the Engineered Barrier System (EBS) of both general and high-efficiency design perspective; (iii) Collaboration with other researchers to evaluate the disposal system using an open-source approach. To achieve these objectives, performance assessments of the various disposal systems and BMTs (BenchMark Test), conducted as part of the DECOVALEX projects, were studied regarding TH behavior. Additionally, integrity assessments of various DGR systems based on thermal criteria were carried out. According to the results, HADES showed very reasonable results, such as evolutions and distributions of temperature and degree of saturation, when compared to validated code such as TOUGH-FLAC, ROCMAS, and OGS (OpenGeoSys). The calculated data are within the range of estimated results from existed code. Furthermore, the first version of the code, which can estimate the TH behavior, has been prepared to share the contents using Git software, a free and open-source distribution system.

Dry storage of nuclear fuel is compromised by threats to the cladding integrity, such as creep and hydride reorientation. To predict these phenomena, spent fuel simulation codes have been developed. In spent fuel simulation, temperature information is the most influential factor for creep and hydride formation. Traditional fuel simulation codes required a user-defined temperature history input which is given by separate thermal analysis. Moreover, geometric changes in nuclear fuel, such as creep, can alter the cask’s internal subchannels, thereby changing the thermal analysis. This necessitates the development of a coupled thermal and nuclear fuel analysis code. In this study, we integrated the 2D FDM nuclear fuel code GIFT developed at SNU with COBRA -SFS. Using this, we analyzed spent nuclear stored in TN-24P dry storage cask over several decades and identified conditions posing threats due to phenomena like creep and hydrogen reorientation, represented by the burnup and peak cladding temperature at the start of dry storage. We also investigated the safety zone of spent nuclear fuel based on burnup and wet storage duration using decay heat.

This paper presents a study on the design and implementation of a secure contactless system leveraging Quick Response (QR) codes as a core component. The main goal of this system is to bridge the gap between strong security and improved user experience within the realm of digital interaction. The system's versatility can be expanded with broad compatibility with a variety of applications. Utility can be expanded to areas such as contactless payments, electronic ticketing, secure identity verification, and convenient access to medical records. The international standardization of QR codes ensures seamless cross-platform compatibility, strengthening their role in the digital ecosystem. We actually create and develop a non-contact security QR code system and check the expandability of the system. This study highlights the pivotal role of QR codes within the realm of secure contactless systems. Through its effective balance of digital security and user convenience, QR codes are emerging as an important element in the continued development of a secure and user-friendly digital environment. The potential for future research lies in exploring more complex use cases and further advancements that improve both security and user-centered design.

In case a spent nuclear fuel transport cask is lost in the sea due to an accident during maritime transport, it is necessary to evaluate the critical depth by which the pressure resistance of the cask is maintained. A licensed type B package should maintain the integrity of containment boundary under water up to 200 m of depth. However, if the cask is damaged during accidents of severity excessing those of design basis accidents, or it is submerged in a sea deeper than 200 m, detailed analyses should be performed to evaluated the condition of the cask and possible scenarios for the release of radioactive contents contained in the cask. In this work, models to evaluate pressure resistance of an undamaged cask in the deep sea are developed and coded into a computer module. To ensure the reliability of the models and to maintain enough flexibility to account for a variety of input conditions, models in three different fidelities are utilized. A very sophisticated finite element analysis model is constructed to provide accurate response of containment boundary against external pressure. A simplified finite element model which can be easily generated with parameters derived from the dimensions and material properties of the cask. Lastly, mathematical formulas based on the shell theory are utilized to evaluate the stress and strain of cask body, lid and the bolts. The models in mathematical formula will be coded into computer model once they show good agreement with the other two model with much higher fidelity. The evaluation of the cask was largely divided into the lid, body, and bottom, bolts of the cask. It was confirmed that the internal stress of the cask was increased in accordance with the hydrostatic pressure. In particular, the lid and bottom have a circular plate shape and showed a similar deformation pattern with deflection at the center. The maximum stress occurred where the lid was in the center and the bottom was in contact with the body. Because the body was simplified and evaluated as a cylinder, only simple compression without torsion and bending was observed. The maximum stress occurred in the tangential direction from the inner side of the cylinder. The bolt connecting the lid and the body was subjected to both bending and tension at the same time, and the maximum stress was evaluated considering both tension and bending loads. In general, the results calculated by the formulas were evaluated to have higher maximum stresses than the analysis results of the simplified model. The results of the maximum stress evaluation in this study confirms that the mathematical models provide conservative results than the finite element models and can be used in the computer module.

The measurement activities to evaluate material balance of nuclear material are usually performed by operator. It is because that the IAEA does not have enough manpower to carry out nuclear measurement accountancy of all nuclear materials in the world. Therefore, the IAEA should consider scenarios which facility operator tries to divert nuclear material for misuse by distorting measurement record. It is required to verify the operator’s measurement data whether it is normal or not. IAEA measures inventory items using their own equipment which is independent of facility operator equipment for verification. Since all inventory lists cannot be verified due to limited resources, the number of items to be verified is determined through statistical method which is called as sample size calculation. They measure for the selected items using their own equipment and compares with operator’s record. The IAEA determines sample size by comprehensively considering targeted diverted nuclear material amount and targeted non-detection probability and performance of measurement equipment. In general, the targeted diverted nuclear material amount is considered significant quantity (plutonium: 8 kg, uranium-235: 75 kg). If the targeted non-detection probability or the performance of the verification equipment is low, the sample size increases, and on the contrary, in the case of high non-detection probability or good performance of verification equipment, even a small sample size is satisfied. It cannot be determined from a single sample size calculation because there are so many sample size combinations for each verification equipment and there are many diversion scenarios to be considered. So, IAEA estimates initial sample size based on statistical method to reduce calculation load. And then they calculate non-detection probability for a combination of initial sample size. Through the iteration calculation, the sample size that satisfies the closest to the target value is derived. The sample size calculation code has been developed to review IAEA’s calculation method. The main difference is that IAEA calculates sample size based on approximate equation, while in this study, sample size is calculated by exact equation. The benchmarking study was performed on reference materials. The data obtained by the code show similar results to the reference materials within an acceptable range. The calculation method developed in this study will be applied to support IAEA and domestic inspection activities in uranium fuel fabrication facility.

Prior to the investigations on fuel degradation it is necessary to describe the reference characteristics of the spent fuel. It establishes the initial condition of the reference fuel bundle at the start of dry storage. In a few technology areas, CANDU fuels have not yet developed comprehensive analysis tools anywhere near the levels in the LWR industry. This requires significantly improved computer codes for CANDU fuel design. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, clad stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking, delayed hydride cracking, and sheath splitting due to UO2 oxidation for a defective fuel. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. The second step was an examination of the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the in-house code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

본 논문에서는 국내 최초로 건축구조기준(KBC 2016)에 기반하여 확률적 영역에서 초과손상확률 형태와 평균손상확률 형태 의 강풍 취약도 평가 방법론을 개발하였다. 본 연구에서는 풍하중에 대한 3초 순간풍속의 영향을 고려할 수 있는 풍하중 산정식을 건 축구조기준을 기초로 유도하였다. 또한 풍하중과 관련된 문헌을 기초로 유도된 3초 순간풍속 기반의 풍하중 산정식에 적용할 수 있는 풍하중 산정계수의 통계치를 제시하였다. 본 연구에서는 초과손상확률 형태와 평균손상확률 형태의 강풍 취약도를 평가하기 위하여 몬테카를로 모사(Monte Carlo Simulation) 기법을 이용하여 해석적 확률 모델을 개발하였다. 제안한 강풍 취약도 평가 방법론의 신뢰 성은 저층 건축물 모형의 지붕 쉬딩 패널 시스템(roof sheathing panel system)을 대상으로 ASCE(American Society of Civil Engineers) 풍하중 기준을 적용한 취약도 평가 방법론의 결과와 비교·검증되었다. 본 연구는 국내 건축구조기준의 풍하중 산정식을 이용하여 강 풍 취약도의 평가 방법론을 보이며, 제안된 방법론에 의한 강풍 취약도는 기존 ASCE 기반 방법론의 결과와 비교하여 작은 오차 범위 내에서 잘 일치함이 확인되었다. 본 연구에서 제시한 강풍 취약도 평가 방법론은 자연재해저감계획 등에 따른 강풍 피해 예측 시 취약 도 구축 방법으로 적용될 수 있을 것으로 판단된다.

Recognize the QR code and develops the position and orientation of the robot can recognize the robot. It is expected to become the innovative technology of robotic navigation systems and logistics systems. The existing vision of the position recognition method(Vision) or artificial surface(Artificial Landmark)based positioning of pushing the location recognition promoted to use a commercially available wireless signal. When commercially available through these technology are expected to be able to make the logistics robot capable of precise position recognition excellent in cost and performance. In the case of the Amazon by Kiva Systems of automation and robotics technology and logistics system in the same way that suggests supplied to the consumer in the short term it innovates in the current logistics. This same technology is location-aware robot control system of the Amazon and is expected to be an innovative logistics system to transfer after development is complete.

When a radiation detector is applied to the measurement of the radioactivity of high-level of radioactive materials or the rapid response to the nuclear accident, several collimators with the different inner radii should be prepared according to the level of dose rate. This makes the in-situ measurement impractical, because of the heavy weight of the collimator. In this study, an IRIS collimator was developed so as to have a function of controlling the inner radius, with the same method used in optical camera, to vary the attenuation ratio of radiation. The shutter was made to have the double tungsten layers with different phase angles to prevent the radiation from penetrating owing to the mechanical tolerance. The performance evaluation through the MCNP code was conducted by calculating the attenuation ratio according to the inner radius of the collimator. The attenuation ratio was marked on the outer scale ring of the collimator. It is expected that when a radiation detector with the IRIS collimator is used for the in-situ measurement, it can change the attenuation ratio of the incident photon to the detector without replacing the collimator.

PURPOSES: A viscoelastic axisymmetric finite element analysis code has been developed for stress analysis of asphalt pavement structures. METHODS: Generalized Maxwell Model (GMM) and 4-node isoparametric element were employed for finite element formulation. The code was developed using C++ computer program language and named as KICTPAVE. For the verification of the developed code, a structural model of a pavement system was constructed. The structural model was composed of three layers: asphalt layer, crushed stone layer, and soil subgrade. Two types of analysis were considered for the verification: (1)elastic static analysis, (2)viscoelastic time-dependent analysis. For the elastic static analysis, linear elastic material model was assigned to all the layers, and a static load was applied to the structural model. For the viscoelastic time-dependent analysis, GMM and linear elastic material model were assigned to the asphalt layer and all the other layers respectively, and a cyclic loading condition was applied to the structural model. RESULTS: The stresses and deformations from KICTPAVE were compared with those from ABAQUS. The analysis results obtained from the two codes showed good agreement in time-dependent response of the element under the loading area as well as the surface deformation of asphalt layer, and horizontal and vertical stresses along the axisymmetric axis. CONCLUSIONS: The validity of KICTPAVE was confirmed by showing the agreement of the analysis results from the two codes.