국내 핵의학은 1959년에 갑상선 질환 환자에서 131I를 이용하여 섭취 및 배출을 측정하면서 시작된 이후, 지난 60여 년간 괄목할 만한 발전을 이루어 왔다. 1961년에 도입된 핵의학 진단영상 검사는 감마카메라를 이용한 감마카메라영상 및 양전자단층 촬영(positron emission tomography, PET)을 이용한 PET/computed tomography (CT)가 현재 주요 검사로 자리잡고 있다. 감마 카메라와 PET/CT에 활용되는 방사성동위원소는 발생기 (generator)와 사이클로트론(cyclotron)을 통해 생산되며, 이러한 방사성동위원소는 표적 장기에 선택적으로 섭취되는 화합물에 표지되어 방사성의약품으로 조제된다. 국내에서 췌장담도 질환 환자에 주로 사용되는 핵의학 진단영상검사용 방사성의약품으 로는 전신뼈스캔에 사용되는 99mTc-dicarboxypropane diphosphonate (DPD)와 99mTc-methylene diphosphonate (MDP), 99mTc-hydroxymethylene diphosphonate (HMDP)가 있으며, 간담도스캔에는 99mTc-bromotriethyliminodiacetic acid (BrIDA 또는 mebrofenin)가 있다. 또한 18F-fluorodeoxyglucose (18F-FDG)와 18F-2-fluoro-3,4-dihydroxyphenylalanine (18F-FDOPA), 111Inpentetreotide (octreotide), 68Ga-1,4,7,10-tetraazacyclododecane- 1,4,7,10-tetraacetic acid0-Tyr3-octreotide (DOTA-TOC)는 주로 췌장담도계 종양의 진단과 치료 방침 결정에 유용하게 활용 되고 있다. 핵의학 진단영상검사로 인한 환자의 의료 피폭은 국내 자연 방사선으로 의한 방사선량과 비교하여 수용 가능한 수준으로 여겨진다. 임상의가 핵의학 진단영상검사의 특성을 충분히 이해하고 이를 환자와 효과적으로 소통할 경우, 신뢰 관계 형성은 물론 진료의 질 향상에도 크게 기여할 수 있을 것이다.

For risk assessment of spent nuclear fuel (SNF) transportation, it is necessary to calculate the damage ratio of SNF rods loaded in the cask. Due to the complexity in the geometry and material properties of SNF, it is impractical to analyze the detailed behavior of every fuel rod and assembly in a single cask model. This study presents a framework for performing cask-level analysis by sequentially simplifying the fuel rods and spent fuel assemblies for fuel damage ratio (FDR) calculation. Using the simplified fuel rod model developed in previous studies, we constructed a CE 16×16 fuel assembly model and presented a methodology to simplify the CE 16×16 assembly model into cuboids. Cask drop analyses were performed to validate the similarity of the detailed CE 16×16 model and the simplified model. Using the proposed simplified models, a procedure for quantifying the bending load and pinch load applied to the fuel rods during the drop impact is presented. The FDR can then be calculated by comparing the quantified loads with their respective failure criteria. Through a case study, the feasibility of the developed framework for systematic and accurate FDR calculation was effectively demonstrated.

Visfatin, an adipokine secreted by cells, is crucial for intracellular nicotinamide adenine dinucleotide+ biosynthesis. Extracellularly, visfatin plays diverse roles in inflammatory conditions, including obesity, which is closely linked to osteoclastogenesis. We previously showed that visfatin enhances receptor activator of nuclear factor kappa-B ligand (RANKL)-induced osteoclastogenesis in bone marrow-derived macrophages. However, its enzymatic activity during this process is poorly understood. Here, we investigated visfatin’s effects on RANKL-induced osteoclast differentiation. Our results demonstrate that visfatin promotes this differentiation, an effect inhibited by FK866, an inhibitor of visfatin’s enzymatic activity. Furthermore, FK866 also inhibited RANKL-induced osteoclast differentiation. These findings suggest that inhibiting visfatin’s enzymatic activity modulates osteoclast differentiation. Thus, visfatin plays an important role in osteoclastogenesis, both intracellularly and extracellularly, and FK866 has therapeutic potential for diseases characterized by imbalanced osteoclast formation, such as osteoporosis and periodontitis.

Securing the safeguardability of a reprocessing process for spent nuclear fuels (SNFs) is imperative. Particularly, the quantity of special nuclear materials inside SNFs must be estimated with the highest achievable precision. Unlike aqueous reprocessing, pyro-processing involves handling input materials in a solid state. Hence, partially extracted samples analyzed by destructive assay (DA) should maintain an acceptable level of representativeness. In this study, a representative sampling method widely applied in the pharmaceutical industry was adopted for homogenization in the head-end process of pyro-processing. By employing representative sampling, specifically based on the mechanism of the rotary riffler, the overall process of homogenization prior to DA analysis was simplified, and less probable hold-up that could contribute to materials unaccounted for (MUF) would be expected. The resulting Pu sampling uncertainty was confirmed to be less than 1% (for ≥ 1,000 μm particle size and ≤ 5 kg sample mass), ensuring sufficient control of Pu accounting uncertainty at a reasonably low level (≤ 1%). Thus, representative sampling can be a competitive alternative to previously suggested methodologies.

This study aims to optimize the orifice diameter to reduce pressure hunting in the pilot valves of positioners used in nuclear power plant control systems. Computational Fluid Dynamics (CFD) analysis using ANSYS CFX was conducted to create 3D models with varying orifice diameters (1 mm, 1.5 mm, 2 mm, 2.5 mm, and 3 mm). To enhance the accuracy of the analysis, boundary layer meshing techniques (Inflation) were applied, and the SST k-ω turbulence model was employed. The analysis of pressure variation and pressure hunting over time revealed that larger orifice diameters resulted in reduced pressure hunting, with a 3 mm orifice diameter achieving 0% pressure hunting. Additionally, it was observed that larger orifice radii slightly increased the average outlet pressure. Based on the findings, a 3 mm orifice diameter is recommended to effectively mitigate pressure hunting in pilot valves, contributing to improved system stability in nuclear power plants. Future studies will explore the design of slanted orifices to further analyze fluid flow characteristics.

The reliability of control valves is critical in nuclear power plants to ensure precise fluid regulation and prevent risks associated with overheating or decreased efficiency. Recently, the supply of imported control valves used in these plants has been discontinued, making the development of domestic alternatives an urgent necessity. This study focuses on the design of an orifice in the pilot valve pipe of a positioner to reduce hunting, a key issue that compromises control stability. Fluid analysis was conducted using ANSYS CFX to investigate the fluid behavior in the pipe with the orifice. The analysis methods included enhanced meshing techniques, turbulence models, and residual values to improve convergence and accuracy. To meet the operational requirements of nuclear power plants (outlet pressure: 3.2 bar, inlet pressure: 7 bar), the inlet fluid velocity was determined. The pressure and pressure hunting were analyzed. Results showed that the selected inlet velocity satisfied the operational conditions, and pressure hunting values were measured and analyzed. The findings provide a basis for further optimizing orifice shapes to achieve the target pressure hunting value of 0.5%.

본 연구는 Tuned Mass Damper(TMD)가 적용된 원자력 발전소 파이핑 시스템의 동적 응답 저감 효과를 평가하기 위해 수행되었다. ABAQUS를 활용하여 실제 크기의 파이핑 시스템 유한요소 모델을 개발하고, 실험 데이터를 통해 모델의 적합성을 검증하 였다. 이후, 확장된 수치해석을 통해 국부 손상 발생 시 TMD의 응답 저감 효과를 분석하였다. 연구 결과, TMD는 무손상 상태에서 가속도와 변위 응답을 각각 최대 20%와 30% 저감하는 효과를 보였으며, 특정 국부 손상(30%, 50%, 70%)에서도 저감 효과가 유지됨 을 확인하였다. 이는 국부 손상이 시스템의 주파수 특성에 미치는 영향이 제한적임을 시사한다. 그러나 손상의 위치와 응답 특성에 따라 저감 효과에는 차이가 있었으며, 최대 응답 위치에서 TMD의 효과가 보다 두드러졌다. 본 연구는 선형 해석에 초점을 맞췄으며, 향후 비선형 재료 특성과 다양한 지진 조건을 고려한 추가 연구가 필요함을 제안한다.

이 연구에서는 김정은 시대 북한의 핵 관련 법에 나타난 핵전략과 핵 태세를 분석하였다. 북한은 미국의 대북 적대시 정책에 대한 대응으로 핵을 개발하였다고 주장하였으나, 김정은 시대 북한은 핵 관련 법을 제 정하고 핵을 무기화하였으며 핵능력을 고도화하였으며 당 제8차 대회 이 후에는 전술핵능력까지 구비하였다. 연구를 통해 북한이 핵을 군사적으 로 사용할 능력을 구비하고 고도화하면서 핵전략과 핵태세가 공세적으로 변화하였음을 확인할 수 있었다. 북한의 핵능력이 고도화되면서 핵전략 은 선언적 수준의 최소억지전략에서 대남 제한억지전략과 대미 최소억지 전략으로 발전하였고, 핵태세 또한 현재는 가장 공세적인 비대칭확장태 세 유형을 보여주고 있다. 북한의 핵·미사일 능력 고도화는 계속될 것이 다. 북한의 핵능력증가에 따른 대응 방안 마련이 제한되는 우리로서는 궁극적으로 북한의 핵 위협이 증가하지 않도록 하는 것이 중요하다.

본 논문은 29개국의 ISSP 환경 설문조사 데이터를 이용해 핵발전 위 험에 대한 성별 인식 차이를 계층선형모형(HLM)을 사용하여 개인수준과 국가수준으로 나누어 분석하였다. 개인 수준에서는 기존에 핵발전소 위 험인식에 영향을 미쳤던 변수들의 효과성이 확인되었으며 국가 차원에서 는 국가 차원의 성불평등이 중요한 예측 변수로 나타났다. 구체적인 국 가차원의 변수를 특정하자면 모성 사망률과 청소년 출산율이 핵심 요인 으로 작용하며, 이는 여성의 정치적 대표성이나 고등교육 수준보다 더 큰 변수로 작용한다. 따라서 핵발전소 위험 인식에 있어서는 여성들의 재생산 권리와 건강권에 따른 젠더 격차가 환경인식에 더 중요한 영향을 끼침을 보여준다. 이러한 결과는 핵에너지에 대한 대중 인식을 이해하기 위해 개인적 요인뿐만 아니라 구조적 불평등까지 고려해야 하는 중요성 을 강조하며, 성별에 따른 사회적 건강 격차가 위험 인식 형성에 어떻게 영향을 주는지에 대한 추가 연구의 필요성을 제기한다.

The transportation of spent nuclear fuel between management stages is expected, and the transportation workers may be exposed to radiation. When transporting spent nuclear fuel, the ALARA principle must be observed for the workers. The objective of this study is to assess a radiation dose for workers transporting spent nuclear fuel using metal overpacks. For this objective, the cask to be handled was selected and the radiation source term was set. Then, the radiation exposure scenario for the transportation workers was defined. Finally, the dose rates for each location of operation were assessed using Monte Carlo simulations, and collective doses were derived for each operation considering the radiation exposure scenario. Each worker performed 11 operations to transport spent nuclear fuel to other facilities and was exposed to a total of 1.138 man-mSv. The operation of removing the bottom shield ring resulted in the highest radiation exposure at 0.503 man-mSv. In contrast, the operation of installing the impact limiter resulted in the lowest radiation exposure at 0.0009 man-mSv. The results of this study can be used to strengthen radiation protection measures for workers transporting spent nuclear fuel in dry storage facilities using metal overpacks.

A multi-barrier can be applied for the deep geological disposal of high-level radioactive waste. The multi-barrier comprises an engineered barrier and the natural barrier of the host rock. In the engineered barrier, the bentonite buffer is the key component for the disposal container, and the bentonite buffer thickness is given important consideration when designing the engineered barrier. This study reviewed the safety functions of bentonite buffers. Subsequently, the requirements and factors necessary to determine the thickness of the bentonite buffer, including criteria for radiological safety and the thermal stability of the disposal system, were identified. Additionally, the bentonite buffer thicknesses required for the top, bottom, and side of the disposal container were calculated. A double-layered emplacement method is also proposed for the bentonite buffer to improve disposal efficiency in terms of thermal management. Based on radiological safety and thermal stability analyses, an optimal thickness of 0.36 m was found to be appropriate for the bentonite buffer surrounding the disposal container. The thickness of the bentonite buffer above the disposal container can be determined based on the excavation damaged zone depth. The study findings can be used as a reference when designing deep geological disposal systems.

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.