This study investigates the risk reduction effect and identifies the optimal capacity of Multi-barrier Accident Coping Strategy (MACST) facilities for nuclear power plants (NPPs) under seismic hazard. The efficacy of MACST facilities in OPR1000 and APR1400 NPP systems is evaluated by utilizing the Improved Direct Quantification of Fault Tree with Monte Carlo Simulation (I-DQFM) method. The analysis encompasses a parametric study of the seismic capacity of two MACST facilities: the 1.0 MW large-capacity mobile generator and the mobile low-pressure pump. The results demonstrate that the optimal seismic capacity of MACST facilities for both NPP systems is 1.5g, which markedly reduces the probability of core damage. In particular, the core damage risk is reduced by approximately 23% for the OPR1000 system, with the core damage fragility reduced by approximately 72% at 1.0g seismic intensity. For the APR1400 system, the implementation of MACST is observed to reduce the core damage risk by approximately 17% and the core damage fragility by approximately 44% under the same conditions. These results emphasize the significance of integrating MACST facilities to enhance the resilience and safety of NPPs against seismic hazard scenarios, highlighting the necessity for continuous adaptation of safety strategies to address evolving natural threats.

본 연구는 북한이 2024년을 전쟁 준비 완성의 해로 선언하고 연이어 미사일을 발사하여 안보를 위협하는 상황에서, 빅데이터 분석을 활용하 여 한국 언론보도와 포털 사이트에 나타난 북핵 및 미사일 위협에 대한 담론과 인식의 특성을 실증적으로 분석하고, 그에 따른 시사점을 도출하 는 것을 목적으로 한다. 이를 위해 국내 주요 언론보도와 포털 사이트에 서 총 33,318건의 데이터를 수집하여, TF-IDF 분석을 통해 상위 50개 의 주요 키워드를 도출하고, 사회연결망 분석을 통해 각 키워드 간의 연 결 정도와 구조를 파악하였다. 분석 결과, 러시아-우크라이나 전쟁, 이스 라엘-하마스 전쟁 등 국제적 안보 불안과 동북아에서의 북-러 군사협력 및 한-미-일 군사협력의 대립 구도 등이 사회적 담론 형성에 큰 영향을 미친 것으로 나타났다. 이에 따라 한-미-일 군사협력 강화와 확장 억제 전략의 신뢰성을 높이고, 사회적 차원에서 위기의식과 안보의식의 제고 가 필요하다는 시사점이 도출되었다.

PURPOSES : Recently, interest in radioactive accidents has increased due to domestic and international nuclear power plant accidents. In particular, local residents' concerns are increasing due to safety issues such as radioactive leaks at the Hanbit Nuclear Power Plant in South Korea. As Gwangju Metropolitan City is not included in the emergency planning area set by the Nuclear Safety and Security Commission, there are significant limitations to establishing disaster prevention measures for nuclear power plant accidents. Considering the Fukushima and Hanbit nuclear power plant accidents, the improvement of Gwangju Metropolitan City's radioactive leak accident response manual is urgently required. This study aimed to establish disaster prevention measures to respond to nuclear power plant accidents in Gwangju Metropolitan City in the event of a Hanbit Nuclear Power Plant accident and to improve resident protection measures by estimating the arrival time of radioactive materials and radiation dosage through a nuclear power plant accident simulation. Additionally, we aimed to supplement the on-site action manual for radioactive leaks at the Hanbit Nuclear Power Plant. METHODS : This study focused on establishing disaster prevention measures centered on Gwangju Metropolitan City in the event of a major accident such as a radioactive leak at the Hanbit Nuclear Power Plant. Simulations were conducted assuming a major accident such as a radioactive leak, measures to improve resident protection were established by calculating the arrival time of radioactive materials and radiation dosage in the Gwangju area in the event of a nuclear power plant accident, and on-site response action manuals were supplemented in response to a radioactive leak. RESULTS : This study considered the concerns of local residents due to the Fukushima nuclear power plant accident and the Hanbit nuclear power plant failure, conducted a simulation to derive the impact on Gwangju Metropolitan City, and examined the effectiveness of an on-site response manual for radioactive leaks to derive improvement measures. CONCLUSIONS : In the event of an accident at the Hanbit Nuclear Power Plant in Gwangju Metropolitan City, insufficient portions of the on-site response action manual should be supplemented, and close cooperation with local governments within the emergency planning area should be ensured to respond to radioactive disasters. Therefore, based on the revised on-site response action manual for radioactive leaks, close cooperation and a clear division of roles among local governments will enable effective resident protection measures to be implemented in the event of a radioactive disaster.

원자력 발전소에 설치되는 안전관련 기기의 손상은 심각한 사고로 이어질 수 있으므로 반듯이 지진안전성을 확보하여야 한 다. MCC, Switchgear, Inverter, Battery charger 등의 전기캐비닛은 대표적인 안전관련 기기이다. 대부분의 실험적 연구는 실험대 상기기의 크기와 실험장비의 성능한계 등으로 인하여 주요부품을 대상으로 하며, 실제 원자력발전소에 납품하는 전기캐비닛을 이용하 여 3축 동시가진에 의한 진동대 실험을 수행한 연구는 많지 않다. 따라서 실제기기를 대상으로 3축 진동대 실험을 통하여 내진성능과 한계상태를 직접적으로 평가하기 위한 연구가 필요하다. 이러한 한계상태평가의 주요 목적은 다양한 부품으로 구성된 캐비닛 단위 실 제기기의 임계 가속도 및 고장 모드를 조사하는 것이다. 본 논문에서는 3축 진동대 실험으로 한계상태 내진성능실험을 수행하여 원자 력발전소에 납품되는 것과 동일한 4종의 전기캐비닛들의 한계상태를 분석하였다.

중국은 경제성장에 치우친 나머지 중국의 환경오염 문제가 국제화되자, 이에 대한 원인을 화석연료의 과다한 사용으로 보고, 이를 해결하고자 화석 연료를 줄이는 대신 안정적인 전력공급을 위해 원자력에너지로의 전환을 계 획하며 많은 원자력 발전소의 추가 건립을 추진하고 있다. 그러나 2011. 3. 후쿠시마 원자력 발전소 사고에서 보았듯이, 효율적인 측면에서 원자력 발 전소는 여느 에너지 공급원보다 대기오염 발생률이 낮고 안정적인 전력을 공급해 주는 것은 사실이지만, 사고 발생 시 국민의 건강과 생명을 위협할 수도 있다. 그럼에도 불구하고 경제성장에 필요한 안정적인 에너지 공급원 이 절실하게 필요한 중국으로서는 이러한 위험을 감수하고라도 경제성장을 위해 선택할 수밖에 없는 에너지원이 원자력에너지일 것이다. 중국은 원자력 발전 기술 수준이 세계적이고, 일부 원자력 발전 기술 분 야의 경우는 세계 최초이며 다른 나라에도 수출하고 있을 정도라며 자부심 을 드러내고 있다. 그러나 동북아시아에서 중국의 지리적 위치와 중국 내 추가 건립되는 원자력 발전소의 입지 현황 등을 고려하면, 제2의 후쿠시마 원자력 발전소 사고와 같은 사고의 발생 가능성을 배제할 수는 없다. 따라서 이 논문에서는 중국의 원자력 발전소 건립계획이 경제성장에 맞춘 부득이한 현실적 선택이라면, 국민의 건강과 안전보장을 위해 최소한의 예 측이 담보할 수 있도록 관련 원자력 안전 관련법의 개정이 필요하다. 이를 위해 첫째, 중국의 원자력 발전소 추가 건립의 위험성을 검토하고, 둘째, 원 자력 발전소 관련 법제를 안전 관련 법제를 중심으로 살펴보고, 셋째, 원자 력의 비중이 가장 높은 프랑스의 법제와 최근에 원자력 사고로 관련 법제를 정비한 일본의 법제 및 한국의 법제를 살펴본 후, 마지막으로 중국의 원자 력 관련 법제가 나아가야 할 방향 및 시사점을 제시하고자 하였다.

Kori Unit 1, the first commercial nuclear power plant (NPP) in Korea, was permanently shut down in 2017 and was scheduled for decommissioning. Various programs must be planned early in the decommissioning process to safely decommission NPPs. Radiological characterization is a key program in decommissioning and should be a high priority. Radiological characterization involves determining the decommissioning technology to be applied to a nuclear facility by identifying the radiation sources and radioactive contaminants present within the facility and assessing the extent and nature of the radioactive contaminants to be removed from the facility. This study introduces the regulatory requirements, procedures, and implementation methods for radiological characterization and proposes a methodology to link the results of radiological characterizations for each stage. To link radiological characteristics, this study proposes to conduct radiological characterization in the decommissioning phase to verify the results of radiological characterization in the transitional phase of decommissioning NPPs. This enables significantly reducing the scope and content of radiological characterization that must be performed in the decommissioning phase and maintaining the connection with the previous phase.

In Korea, two types of spent nuclear fuels (SNFs) are generated, pressurized light water reactor type (PWR) and pressurized heavy water reactor type (PHWR; CANDU), that differ greatly in size, decay heat, and radioactive characteristics. Technology development for the disposal of SNFs has mainly focused on PWR SNFs that are large in size and have extremely high decay heat and radioactivity. However, CANDU SNFs should be considered differently from PWR SNFs in deep geological disposal systems because their characteristics significantly differ from those of PWR SNFs in terms of their dimensions, number of SNF bundles, and handling systems in nuclear power plant sites. In this paper, after reviewing the status of the CANDU SNF disposal concept by Canada and Korea, concepts related to the direct geological disposal of CANDU SNFs were described, and two concepts were proposed based on the results of the development. The engineered barrier systems developed using these two concepts were comparatively analyzed in terms of disposal safety, disposal efficiency, and technical maturity. Based on the results of the comparative analyses, a vertical-type emplacement disposal concept was determined as a reference concept for the deep geological disposal of CANDU SNFs.

Interim dry cask storage systems comprising AISI 304 or 316 stainless steel canisters have become critical for the storage of spent nuclear fuel from light water reactors in the Republic of Korea. However, the combination of microstructural sensitization, residual tensile stress, and corrosive environments can induce chloride-induced stress corrosion cracking (CISCC) for stainless steel canisters. Suppressing one or more of these three variables can effectively mitigate CISCC initiation or propagation. Surface-modification technologies, such as surface peening and burnishing, focus on relieving residual tensile stress by introducing compressive stress to near-surface regions of materials. Overlay coating methods such as cold spray can serve as a barrier between the environment and the canister, while also inducing compressive stress similar to surface peening. This approach can both mitigate CISCC initiation and facilitate CISCC repair. Surface-painting methods can also be used to isolate materials from external corrosive environments. However, environmental variables, such as relative humidity, composition of surface deposits, and pH can affect the CISCC behavior. Therefore, in addition to research on surface modification and coating technologies, site-specific environmental investigations of various nuclear power plants are required.

Hydride analysis is required to assess the mechanical integrity of spent nuclear fuel cladding. Image segmentation, which is a hydride analysis method, is a technique that can analyze the orientation and distribution of hydrides in cladding images of spent nuclear fuels. However, the segmentation results varied according to the image preprocessing. Inaccurate segmentation results can make hydride difficult to analyze. This study aims to analyze the segmentation performance of the Otsu algorithm according to the morphological operations of cladding images. Morphological operations were applied to four different cladding images, and segmentation performance was quantitatively compared using a histogram, betweenclass variance, and radial hydride fraction. As a result, this study found that morphological operations can induce errors in cladding images and that appropriate combinations of morphological operations can maximize segmentation performance. This study emphasizes the importance of image preprocessing methods, suggesting that they can enhance the accuracy of hydride analysis. These findings are expected to contribute to the advancements in integrity assessment of spent nuclear fuel cladding.

A procedure for minimizing the environmental burden and maximizing the efficiency of storage sites used for the final disposal of spent fuel has been proposed. In this procedure, fission products (highly mobile and producing heat) are collected, and uranium and TRU-RE (transuranium-rare earth) oxide are independently stored. The possibility and applicability of radiation measurement for monitoring the nuclear materials effectively throughout the process has been simulated and evaluated. For the simulation, the properties of the chemical processes were analyzed, the major radiation emitters were determined, and the production of nuclear materials by chemical reactions were evaluated. In each process, the content of nuclear material was changed by up to 20% to represent abnormal conditions. The results showed that the plutonium peak was matched with the change in the TRU content and the measured signal was changed linearly with respect to the content change of the plutonium. From the neutron measurement, a linear response of the TRU content variation was obtained. In addition, a logic diagram was developed for the nuclear monitoring. The integration of radiation detections is recommended for monitoring the process effectively and efficiently.

Dynamic responses of nuclear power plant structure subjected to earthquake loads should be carefully investigated for safety. Because nuclear power plant structure are usually constructed by material of reinforced concrete, the aging deterioration of R.C. have no small effect on structural behavior of nuclear power plant structure. Therefore, aging deterioration of R.C. nuclear power plant structure should be considered for exact prediction of seismic responses of the structure. In this study, a machine learning model for seismic response prediction of nuclear power plant structure was developed by considering aging deterioration. The OPR-1000 was selected as an example structure for numerical simulation. The OPR-1000 was originally designated as the Korean Standard Nuclear Power Plant (KSNP), and was re-designated as the OPR-1000 in 2005 for foreign sales. 500 artificial ground motions were generated based on site characteristics of Korea. Elastic modulus, damping ratio, poisson’s ratio and density were selected to consider material property variation due to aging deterioration. Six machine learning algorithms such as, Decision Tree (DT), Random Forest (RF), Support Vector Machine (SVM), K-Nearest Neighbor (KNN), Artificial Neural Networks (ANN), eXtreme Gradient Boosting (XGBoost), were used t o construct seispic response prediction model. 13 intensity measures and 4 material properties were used input parameters of the training database. Performance evaluation was performed using metrics like root mean square error, mean square error, mean absolute error, and coefficient of determination. The optimization of hyperparameters was achieved through k-fold cross-validation and grid search techniques. The analysis results show that neural networks present good prediction performance considering aging deterioration.

Climate change has led to a significant increase in jellyfish populations globally, causing various problems. For power plants that use nearby seawater for cooling, the intrusion of jellyfish into intake systems can block the flow, leading to reduced output or even shutdowns. This issue is compounded by other small marine organisms like shrimp and salps, making it urgent to develop solutions to prevent their intrusion. This study addressed the problem using the BioSonics DT-X 120 kHz scientific fish finder to conduct preliminary tank experiments. We also deployed underwater acoustic and camera buoys around the intake of nuclear power plant, utilizing a bidirectional communication system between sea and land to collect data. Data collection took place from July 31, 2023 to August 1, 2023. While harmful organisms such as jellyfish and salps were not detected, we successfully gathered acoustic data on small fish measuring backscattering strength (SV). Analysis showed that fish schools were more prominent in the evening than during the day. The highest fish distribution was observed at 3:30 AM on July 31 with an SV of -44.8 dB while the lowest was at 12:30 PM on the same day with an SV of –63.4 dB. Additionally, a solar-powered system was used to enable real-time data acquisition from sea buoys with smooth communication between the land server and the offshore buoy located 1.8 km away. This research developed an acoustic-based monitoring system for detecting harmful organisms around the intake and provided foundational data for preventing marine organism intrusion and planning effective measures.

For the OPR1000, a standard power plant in Korea, an analytical model of the containment building considering voids and deterioration was built with multilayer shell elements. Voids were placed in the vulnerable parts of the analysis model, and the deterioration effects of concrete and rebar were reflected in the material model. To check the impact of voids and deterioration on the seismic performance of the containment building, iterative push-over analysis was performed on four cases of the analytical model with and without voids and deterioration. It was found that the effect of voids with a volume ratio of 0.6% on the seismic performance of the containment building was insignificant. The effect of strength reduction and cross-sectional area loss of reinforcement due to deterioration and the impact of strength increase of concrete due to long-term hardening offset each other, resulting in a slight increase in the lateral resistance of the containment building. To determine the limit state that adequately represents the seismic performance of the containment building considering voids and deterioration, the Ogaki shear strength equation, ASCE 43-05 low shear wall allowable lateral displacement ratio, and JEAC 4601 shear strain limit were compared and examined with the analytically derived failure point (ultimate point) in this study.

Because most spent nuclear fuel storage casks have been designed for low burnup fuel, a safety-significant high burnup dry storage cask must be developed for nuclear facilities in Korea to store the increasing high burnup and damaged fuels. More than 20% of fuels generated by PWRs comprise high burnup fuels. This study conducted a structural safety evaluation of the preliminary designs for a high burnup storage cask with 21 spent nuclear fuels and evaluated feasible loading conditions under normal, off-normal, and accident conditions. Two types of metal and concrete storage casks were used in the evaluation. Structural integrity was assessed by comparing load combinations and stress intensity limits under each condition. Evaluation results showed that the storage cask had secured structural integrity as it satisfied the stress intensity limit under normal, off-normal, and accident conditions. These results can be used as baseline data for the detailed design of high burnup storage casks.

To non-destructively determine the burnup of a spent nuclear fuel assembly, it is essential to analyze the nuclear isotopes present in the assembly and detect the neutrons and gamma rays emitted from these isotopes. Specifically, gamma-ray measurement methods can utilize a single radiation measurement value of 137Cs or measure based on the energy peak ratio of Cs isotopes such as 134Cs/137Cs and 154Eu/137Cs. In this study, we validated the extent to which the results of gamma-ray measurements using cadmium zinc telluride (CZT) sensors based on 137Cs could be accurately simulated by implementing identical conditions on MCNP. To simulate measurement scenarios using a lead collimator, we propose equations that represent radiation behavior that reaches the detector by assuming “Direct hit” and “Penetration with attenuation” situations. The results obtained from MCNP confirmed an increase in measurement efficiency by 0.47 times when using the CZT detector, demonstrating the efficacy of the measurement system.

The 300 concrete silo systems installed and operated at the site of Wolsong nuclear power plant (NPP) have been storing CANDU spent nuclear fuel (SNF) under dry conditions since 1992. The dry storage system must be operated safely until SNF is delivered to an interim storage facility or final repository located outside the NPP in accordance with the SNF management policy of the country. The silo dry storage system consists of a concrete structure, liner steel plate in the inner cavity, and fuel basket. Because the components of the silo system are exposed to high energy radiation owing to the high radioactivity of SNF inside, the effects of irradiation during long-term storage must be analyzed. To this end, material specimens of each component were manufactured and subjected to irradiation and strength tests, and mechanical characteristics before and after irradiation were examined. Notably, the mechanical characteristics of the main components of the silo system were affected by irradiation during the storage of spent fuel. The test results will be used to evaluate the long-term behavior of silo systems in the future.

Earthquakes of magnitude 3.0 or greater occur in Korea about 10 times on average yearly, and the number of earthquakes occurring in Korea is increasing. As many earthquakes have recently occurred, interest in the safety of nuclear power plants has increased. Nuclear power plants are equipped with many cabinet-type control facilities to regulate safety facilities, and function maintenance is required during an earthquake. The seismic performance of the cabinet is divided into structural and functional performances. Structural performance can be secured during the design procedure. Functional performance depends on the vibration performance of the component. Therefore, it is necessary to confirm the seismic performance of the components. Generally, seismic performance is confirmed through seismic simulation tests. When checking seismic performance through seismic simulation tests, it is difficult to determine the effect of frequency and maximum acceleration on an element. In this paper, shaking table tests were performed using various frequencies and various maximum accelerations. The seismic performance characteristics of the functions of electrical equipment components were confirmed through tests.