Recently, the status of North Korea’s denuclearization has become an international issue, and there are also indications of potential nuclear proliferation among neighboring countries. So, the need for establishment of nuclear activity verification technology and strategy is growing. In terms of ensuring verification completeness, sample collection-based analysis is essential. The concepts of Chain of Custody (CoC) and Continuity of Knowledge (CoK) can be defined in the process of sample extraction as follows: CoC is interpreted as the ‘system for managing the flow of information subjected by the examinee’, and CoK is interpreted as the ‘Continuity of information collection through CoC subjected by the inspector’. In the case of sample collection process in unreported areas for nuclear activity verification, there are additional risks such as worker exposure/kidnapping or sample theft/tampering. Therefore, the introduction of additional devices might be required to maintain CoC and CoK in the unreported area. In this study, an Environmental Geometrical Data Transfer (EGDT) was developed to ensure the safety of workers and the CoC/CoK of the samples during the collection process. This device was designed for achieving both mobility and rechargeability. It is categorized into two modes based on its intended users: sample mode and worker mode. Through the sensors, which is positioned in the rear part of device, such as radiation, gyroscope, light, temperature, humidity and proximity sensors, it can be easily achievable various environmental information in real-time. Additionally, GPS information can also be received, allowing for responsiveness to various hazardous scenarios. Moreover, the OLED display positioned on the front gives us for checking device information such as the current status of the device such as the battery level, the connectivity of wifi, and etc. Finally, an alarm function was integrated to enable rapid awareness during emergency situations. These functions can be updated and modified through Arduino-based firmware, and both the device and the information collected through it can be remotely controlled via custom software. Based on the presented design conditions, a prototype was developed and field assessments were conducted, yielding results within an acceptable margin of error for various scenarios. Through the application of the EGDT developed in this study to the sample collection process for nuclear activity verification purposes, it is expected to achieve a stable maintenance of CoC/CoK through more accurate information transmission and reception.

In the case of nuclear projects, when developing a new reactor type, it is necessary to confirm the reactor type, secure the safety, and especially obtain the construction permit approval of the licensing authority for construction. Schedule management is necessary to carry out nuclear projects, and progress rate management of project progress management is largely composed of three elements: scope management, cost management, and resource management. However, in the case of the small modular reactor (SMR) project currently being carried out, it is difficult to calculate the progress rate including budget and resources due to the nature of the project. Therefore, in the SMR project, it took two years from the beginning to prepare the integrated project master schedule (IPMS) to prepare the draft, and then two revisions were made over a year and a half. In this SMR project, we will consider the entire construction period such as design, purchase and production, construction, commissioning, and operation in terms of scope management. The entire document list was created using the document review and approval sheet created at the beginning of the design. In the PMIS (Project Management Information System), the number of approved documents was calculated by comparing the list of engineering documents. In the purchase production part, the main core equipment such as the primary system nuclear steam supply system (NSSS), the secondary system turbine and condenser, and the man machine interface system (MMIS) are managed. Purchasing and manufacturing management shall be managed so that major equipment can be delivered in a timely manner in accordance with the schedule for delivery of equipment in the IPMS. In order to prevent delays in the start of production, it is necessary to minimize the waiting time for work through advance management tasks such as insurance of drawing, stocking of materials, availability of production facilities, etc. In this way, we decided to carry out the schedule management for the design, purchase and manufacturing part in the SMR project first, and the installation, construction and commissioning part will be prepared for the future schedule management.

As remote sensing measures, satellite imagery has played an essential role in verifying nuclear activities for decades. Starting with the first artificial satellite, Sputnik 1, in 1957, thousands of satellites are currently missioning in space. Since the 2000s, the level of detail in pixels of an image (spatial resolution) has been significantly improving, thereby identifying objects less than one meter, even tens of centimetres. The more things are identifiable, the wider regions become targets for observation. With the increasing number of satellites, computer vision technology is required to explore the applicability of algorithm-based automation. This paper aims to investigate the R&D publications worldwide from the 1990s to the present, which have tried to apply algorithms to verify any clandestine nuclear activities or detect anomalies at the site. The versatile open-source publications, including the IAEA, ESARDA, US-DOE national laboratories, and universities, are extensively reviewed from the perspective of nuclear nonproliferation (or counter-proliferation). Thus, target objects for applications are essentially located in nuclearrelated sites, and the source type of satellite sensors focuses on electro-optical images with high spatial resolution. The research trend over time by groups is discussed with limitations at the time in order to contemplate the role of algorithms in the field and to present recommendations on a way forward.

Korea has an agreement for cooperation with 31 countries, including the United States, Canada, Australia, and Japan. Under the agreement, the obligated items must be used for peaceful purposes, comply with nuclear non-proliferation and international safeguards, and obtain prior consent of shipment in case of enrichment, reprocessing, retransfer. Among them, the United States, Canada, and Australia have signed Administrative Arrangements of Cooperation Agreements (Supplementary Arrangements in Canada) for the international transfer and annual reports of obligated items. When operators submit an annual report, the government compiles and make the annual report based on the data. Ideally, the final report is submitted by the operator should be the national annual report, but in practice, discrepancies occur between sum of the operator’s and goverment’s. In order to resolve these problems and strengthen the linkage between exports contrpol and safeguards, our institute has begun the project to develop an ‘Obligation Tracking System for internationally controlled items (OTS)’. It is believed that obligated items which are unnecessarily included or omitted in annual report could be managed properly by developing OTS for life cycle of the items such as import, disposal/ termination or transfer to other countries. In case of nuclear material, especially, the characteristics of the facilities (e.g., bulk-handling facilities) must be considered and principles of fungibility, equivalence, and proportionality should be applied to materials. In order to computerize these procedures, we would like to propose to adopt the format of Code 10 for obligated item management. Code 10 is the form of the annex to the Korea-IAEA safeguards agreement which includes all records of inventory changes, import/export, and domestic movement of nuclear materials. It is expected to minimize discrepancies between operators’ annual reporting data and national annual reporting and further contribute to enhancing national trust and nuclear transparency.

To mark the 70th anniversary of the alliance between South Korea and the United States, President Yoon Seok-youl of South Korea and President Joseph R. Biden of the United States convened at the White House, adopting the pivotal “Washington Declaration.” This significant act paved the way for the establishment and institutionalization of the ROK-US Nuclear Consultative Group (NCG). The NCG is envisioned as a mechanism to address North Korea’s nuclear threat, striving for nuclear sharing and a nuclear defense system, thereby alleviating concerns about nuclear security. The NCG is perceived as a crucial advancement in the realm of ‘tailored extended deterrence’ on the Korean Peninsula. However, its operational scope and efficacy remain subjects of debate within South Korea. A comparative analysis with other consultative entities, such as NATO’s Nuclear Planning Group (NPG) and Extended Deterrence Strategy Consultative Group (EDSCG), raises questions about NCG’s unique contributions and potential functional overlaps. Furthermore, the establishment of the NCG represents a notable progression in the strengthened ROK-US alliance. This progression coincides with the resumption of large-scale joint nuclear security military exercises under the new administrations of both nations. Anticipated future operations within the NCG framework encompass the continual deployment of strategic assets and the execution of nuclear simulation exercises. Such actions serve not merely as a deterrent message to North Korea but also aim to instill confidence in the US’s commitment to extended deterrence among the South Korean populace. This study aims to highlight the significance and implications of the ROK-US Nuclear Consultative Group (NCG) through an exhaustive comparative analysis of existing nuclear security consultative bodies and pertinent nuclear security policies. Moreover, this research emphasizes strategies to boost the NCG’s effectiveness, the necessity for policy enhancements to foster South Korea’s nuclear security autonomy, and the importance of raising nuclear security awareness among the general public.

The use of nuclear materials for nuclear power generation is increasing worldwide, and the International Atomic Energy Agency (IAEA) has signed an agreement with countries using nuclear materials to prevent using military purpose through the Non-Proliferation Treaty (NPT) for the management of nuclear materials. Accordingly, all member countries manage nuclear material and equipment facilities under the treaty and are obligated to conduct safety measures such as inspection, containment, and surveillance in accordance with safety standards. The equipment used in the inspection basically consists of a Scintillator type and a semiconductor detector type, and is mainly used for portable equipment to ensure the integrity of the equipment. In general, the operating environment of the detector guaranteed by the manufacturer is -10 degrees to 40 degrees due to poor resolution and electrical problems. However, in the case of an outdoor environment other than a laboratory environment, it is difficult to maintain the above temperature conditions. In particular, the internal temperature of the vehicle used for transport rises to more than 50 degrees in Korea, making the detector stored therein vulnerable. In this study, a storage chamber for extreme environments was developed. The developed chamber compared the internal temperature by heating the external temperature. In addition, the performance before and after heating was compared by heating the radiation detectors HPGe, CZT, and NaI from -20 to 70 degrees Celsius while using the storage chamber. Our proposed chamber can play a key role in applications with good performance in complex environmental adaptability in their design.

Nuclear safety, security, and safeguards (nuclear 3S) are essential components for establishing robust nuclear environments. Nuclear safety is to protect public and environments from radioactive contamination, which can be caused in accidents. Nuclear security is to protect nuclear facilities from terrorism or sabotage, which related to physical a ttacks or insider threats. And nuclear safeguards is to protect nuclear materials from extortion by a state with a purpose of weaponizing activities. When a new nuclear facility is introduced, it is possible to save abundant amount of resources by considering nuclear 3S in an early stage (design phases). Initially, the international atomic energy agency (IAEA) recommended safeguards-by-design (SBD) approach. The concept of SBD gradually expands to nuclear 3S-by-design (3SBD). Though there are differences in purpose and target subject, each nuclear ‘S’ is closely related with others. When introducing a certain technology or equipment in order to enhance one ‘S’, another ‘S’ also get affected. The effect can be synergies or conflicts. For instance, confidential information in nuclear security is required for a safeguards activity. On the contrary, inspection equipment for safeguards can be used for security. Pyroprocessing is a technology for managing used nuclear fuels. As pyroprocessing is a backend fuel cycle technology, a sensitive nuclear technology, safeguards has taken a large portion of nuclear 3S research in an effort to achieve international credibility and nuclear transparency. As mentioned, there are both synergies and conflicts in integrating nuclear 3S. In this study, we investigate potential challenges in applying nuclear 3S integration to pyroprocessing by addressing synergies and conflicts. This approach will suggest required supplementary methods to build the reliable pyroprocessing environment.

The CTBTO is the Comprehensive Test Ban Treaty Organization to ban all forms of nuclear testing (underwater, air, and underground) worldwide and was adopted at the UN’s 50th annual general meeting in September 1996. As of September 2023, 187 out of 196 countries signed and 178 ratified. The Republic of Korea signed it in 1996 and ratified it in 1999. Several major Annex 2 countries still need to ratify it, and certain countries have not even signed it, so it has not come entry into force. The CTBTO has three verification systems for nuclear tests and consists of the International Monitoring System (IMS), the International Data Center (IDC), and On-Site Inspections (OSI). IMS consists of seismic, hydroacoustic, infrasound, and radionuclide monitoring. The measured data are delivered to IDC, analyzed by CTBTO headquarters, distributed raw data, and analyzed forms to member states. The final means of verification is in the field of OSI and will be operated when CTBT takes effect. Based on the IMS data, inspectors will be dispatched to the Inspected State Party (ISP) to check for nuclear tests. KINAC is attending the Working Group B, OSI technology development verification along with KINS and KIGAM. Since OSI is a means for final verification, integrated capabilities such as seismic and data interpretation and nuclides detection are required. CTBTO continues its efforts to foster integrated talent and modernize OSI equipment. Types of equipment include measurement, flight simulation equipment, and geographic information monitoring systems etcetera. KINAC is also developing equipment to detect contaminated areas using drones and probes. Development equipment is the nuclides detection and measurement of contaminated areas, and it is the equipment that prepares a control center and drops probes into suspected contamination areas to find a location of the radiation source. The probe can be used to track the location where the dose is most substantial through Bayesian estimation and source measurement.

The Fukushima-Daiichi accident in 2011 revealed the limitations of Zr-alloys in accident scenarios where severe steam oxidation led to the liberation of heat and hydrogen and the destruction of the reactor core. In response to this accident, there has been a concerted effort by industry, national laboratories, and universities to develop cladding and fuel materials for lightwater reactors (LWRs) that are more accident tolerant. The near-term approach has been to develop coatings for Zr-alloys that would provide additional safety and operational margin by virtue of its excellent corrosion/oxidation resistance at both normal and accident conditions. The designs being considered for implementation by major nuclear fuel suppliers include a thin Cr or a ceramic coating on the conventional LWR fuel cladding. For improved economics, the industries are also considering ATF coated cladding with high enrichment fuel (up to 8%) to achieve high burnup (> 75 GWd/MTU). While the development of ATF concepts (i.e., the front end of the fuel cycle), including coated claddings and doped fuels have progressed at an accelerated pace, relatively less attention has been devoted to the used fuel disposition of ATF fuels (i.e., the backend of the fuel cycle). For accelerated deployment of the ATF designs in the current LWR fleet, it is necessary to investigate technical aspects of the ATF used nuclear fuel (UNF) management in transportation, storage, and disposal. This presentation will provide a brief overview of state-of-the-art ATF developments and list out potential considerations to apply the fuels into back-end fuel cycle. New test plan should be planned to compare the characteristics of current LWR used nuclear fuels with those of the new fuel designs. For example, research focus can be understanding of ATF used fuel particulate size and quantity (at high burnup condition) and mechanical integrity of coated cladding under normal and off-normal conditions during transportation and long-term storage. Finally, the impacts of CRUD on the new fuel cladding, increased container weight, temperature, and radiation level to the back-end fuel cycle activities need to be investigated.

In pyroprocessing, the residual salts (LiCl containing Li and Li2O) in the metallic fuel produced by the oxide reduction (OR) process are removed by salt distillation and fed into electrorefining. This study undertook an investigation into the potential viability of employing a separate LiCl salt rinsing process as an innovative alternative to conventional salt distillation techniques. The primary objective of this novel approach was to mitigate the presence of Li and Li2O within the residual OR salt of metallic fuel, subsequently facilitating its suitability for electrorefining processes. The process of rinsing the metallic fuel involved immersing it in a LiCl salt environment at a temperature of 650°C. During this immersion process, the residual OR salt contained within the fuel underwent dissolution, thereby reducing the concentrations of Li2O and Li generated during the OR process. Furthermore, the Li and Li2O dissolved within the LiCl salt were effectively consumed through chemical reactions with ZrO2 particles present within the salt. Importantly, even after the metallic fuel had been subjected to rinsing in a conventional LiCl salt solution, the concentration of Li and Li2O within the salt remained consistent with its initial levels, due to the utilization of ZrO2. Moreover, it was observed that the Li- Li2O content within the metallic fuel was significantly diluted as a result of the rinsing process.

Given the situation in the Republic of Korea that all nuclear power plants are located at the seaside, the interim storage facility is also likely to be located at seaside and the maritime transportation of Spent Nuclear Fuel is considered inevitable. The Republic of Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radionuclides from a submerged transportation cask in the sea. Therefore, there is a need to develop a technology that can assess the impact of immersion accidents and establish a regulatory framework for maritime transportation accidents. The release rate of radionuclides should be calculated from the flow rate through a flow path in the breached containment boundary. According to the cask design criteria, it is anticipated that even under severe accident conditions, the flow path size will be very small. Previous studies have evaluated fluid flow passing through micro-scale channel by integrating internal and external flows within and around a transport cask. As part of the evaluation, a comprehensive “Full-Field Model” incorporating external flow fields and a localized “Local-Field Model” with micro-scale flow paths were constructed. Sub-modeling techniques were employed to couple the flow field calculated by the two models. The aforementioned approach is utilized to conduct the evaluation of fluid flow passing through micro-scale flow paths. This study aims to evaluate fluid flow passing through micro-scale flow paths using the aforementioned CFD (Computational Fluid Dynamics) method and aims to code the findings. The Gaussian Process Regression technique, a machine learning model, is utilized for developing a mathematical metamodel. The selected input parameters for coding are organized and their respective impacts are analyzed. The range of these selected parameters is tailored to suit domestic environments, and computational experiments are planned through Design of Experiments. The flow path size is included as an input parameter in the coded model. In cases where the flow path size becomes extremely small, making it impractical to use CFD techniques for calculations, Poiseuille’s law is employed to calculate the release rate. In this study, a model is developed to evaluate the release rate of radionuclides using CFD and mathematical equations covering the whole possible range of flow path size in a lost cask in the deep sea. The model will be used in the development of a maritime transportation risk assessment code suitable for the situation and environment in Korea.

The types of fuel loaded and burned in domestic nuclear power plants are WH-type and OPR/ APR-type nuclear power plants, with a total of 19 types. In the case of spent nuclear fuel released in Korea, the low combustion level of 45,000 MWD/MTU or less accounts for about 75%. In terms of fuel type, WH 17×17 and CE 16×16 fuels account for about 85% of all spent nuclear fuels. The thickness of the oxide film of the fuel cladding can make the fuel rod vulnerable during reactor operation, directly affecting the integrity of the fuel rods. so, it is a very important design factor in design. Therefore, the fuel rod design code that predicts and evaluates this has also been developed to accurately predict fuel rod corrosion. And it’s being applied to the design. In this study, the ECT probe measured by inserting it between fuel rods. The thickness of the fuel cladding oxide film was measured for spent nuclear fuel. When reloading operational nuclear fuel, the IAEA recommends an oxide film thickness of up to 100 micrometers. In this study, it was confirmed that spent nuclear fuels keeping integrity burned for 2-3 cycles were sufficiently maintained within the limit. However, the difference could be confirmed according to the characteristics of the coating material, the combustion cycle, and the use of poison rods. For the reliability of the data, symmetrical to the quadrant fuels were selected, and the fuel burned at the same period was measured. The method of selecting the target fuel can produce meaningful results.

In Korea, Kori Unit 1 and Wolsong Unit 1, have been permanently shut down in 2017 and 2019, and more nuclear power plants are expected to be permanently shut down after continued operation successively. Spent fuel has been generated during operation and stored in spent fuel pools. Due to the expected saturation of spent fuel pools within the next several decades, transportation of a huge amount of spent fuel is anticipated to interim storage facilities or final disposal facilities, even though the specific location is not decided. The U.S. Nuclear Regulatory Commission (NRC) states that every environmental report prepared for the licensing stage of a Pressurized Water Reactor shall contain a statement concerning risk during the transportation of fuel and radioactive wastes to and from the reactor. Thus, the licensee should ensure that the radiological effects in accidents, as well as normal conditions in transport, do not exceed certain criteria or be small if cannot be numerically quantified. Specific conditions that a full description and detailed analysis of the environmental effects of transportation of fuel and wastes to and from the reactor are exempted are specified in 10 CFR Part 51. Since there are no official requirements for radiological dose assessment for workers and public during the transportation of spent fuel in Korea, the margin when applying the U.S. regulatory criteria to the environmental impact assessment during the transport of spent fuel generated from domestic nuclear power plants is evaluated. A different approach would be needed due to the difference in the characteristics of spent fuel and geographical features.

In the establishment of procedures for managing spent fuel, the development of an information system for data management is an indispensable prerequisite. Given the prolonged period of spent nuclear fuel management, marked by numerous personnel changes and the anticipation of vast data retention, addressing this matter appropriately is imperative, particularly in the specialized field of spent nuclear fuel operations. Recognizing the need for a method to mitigate these challenges, we endeavored to apply semantic technology to the information system. To achieve this, we constructed the ontology of spent nuclear fuel and conducted research to transform it into a relational database. As a result, the information system, developed by the application of semantic technology, has attained the capability to comprehend and perceive relationships among information itself. Through this research, the system not only addresses previously identified concerns but also enhances its versatility, enabling it to perform functions previously unattainable within existing information systems.

In nuclear facilities, a graded approach is applied to achieve safety effectively and efficiently. It means that the structures, systems, and components (SSCs) that are important to safety should be assured to be high quality. Accordingly, SSCs that consist of nuclear facilities should be classified with respect to their safety importance as several classes, so that the requirements of quality assurance relevant to the designing, manufacturing, testing, maintenance, etc. can be applied. Guidance for the safety classification of SSCs consisting of nuclear power plants and radioactive waste management facilities was developed by U.S.NRC and IAEA. Especially, in guidance for nuclear power plants, safety significance can be evaluated as following details. The single SSC that mitigates or/and prevents the radiological consequence or hazard was assumed to be failure or malfunction as the initiating event/accident occurred and the following radiological consequence was evaluated. Considering both the consequence and frequency of the occurrence of the initiating event/accident, the safety significance of each SSC can be evaluated. Based on the evaluated safety significance, a safety class can be assigned. The guidance for the safety classification of the spent nuclear fuel dry storage systems (DSS) was also developed in the United States (NUREG/CR-6407) and the U.S.NRC acknowledges the application of it to the safety classification of DSS in the United States. Also, worldwide including the KOREA, that guidance has been applied to several DSSs. However, the guidance does not include the methodology for classifying the safety or the evaluated safety significance of each SSC, and the classification criteria are not based on quantitative safety significance but are expressed somewhat qualitatively. Vendors of DSS may have difficulties to apply this guidance appropriately due to the different design characteristics of DSSs. Therefore, the purpose of this study is to evaluate the safety significance of representative SSCs in DSS. A framework was established to evaluate the safety significance of SSCs performing safety functions related to radiation shielding and confinement of radioactive materials. Furthermore, the framework was applied to the test case.

Hydride reorientation is widely known as one of the major degradation mechanisms in Zirconium cladding during dry storage. Some previous theoretical models for hydride reorientation used assumption of an ideal radial basal pole orientation for HCP structure of Zirconium cladding. Under this assumption, circumferential hydride was considered to precipitate in the basal plane while radial hydride was considered to precipitate in the prismatic plane, thereby giving energetical penalty on thermodynamical precipitation of radial hydrides. However, in reality, reactor-grade Zirconium cladding exhibits average 30° tilted texture, adding complexity to the hydride precipitation mechanism. In this study, reactor-grade Zirconium cladding was charged with hydrogen and hydride reorientation -treated specimens were fabricated. Microstructural characterization of hydrides was conducted via following three methods in terms of interface and stored energy. And this study aimed to compare these characteristics between circumferential and radial hydrides. Using Electron Back Scattered Diffraction (EBSD), the interface was investigated assuming that interface lies parallel to the axial axis of the tube. These were further validated with Transmission Electron Microscope (TEM). In addition, Differential Scanning Calorimetry (DSC) analysis was conducted to calculate the stored energy. This investigation is expected to establish fundamental understanding of how hydrides precipitate in Zirconium cladding with different orientations. And it will also increase the predictability of radial hydride formation and help understanding the mechanical behavior of Zirconium cladding with radial hydrides.

The Spent Nuclear Fuel (SNF) cladding serves as the first barrier that prevents the release of radioactive materials. It is very important to maintain cladding integrity in SNF management. It is known that the pinch load applied to the cladding can lead to Mode-3 failure and the cladding becomes more vulnerable to this failure mode with the existence of radial hydrides and other forms of mechanical defects. In this study, a numerical analysis process was proposed to scientifically and systematically evaluate the fracture resistance of cladding with reoriented hydrides under pinch load. The mechanical behavior and fracture of the irradiated cladding under pinch load can be evaluated by Ring Compression Test (RCT). Under the stress field generated by RCT, the cracks propagate more easily through radial hydrides than circumferential hydrides. The δ-hydride which form within the α-zirconium matrix causes a large expansion strain due to the volume difference and voids form at the interface between the hydride and the zirconium matrix. Chan demonstrated that the load needed to form voids and separate the hard hydride precipitates from the Zr matrix is considerably lower than that which initiates brittle fracture of hydrides using a micro-cantilever test. Therefore, we propose a microstructure crack propagation analysis method based on Continuum Damage Mechanics (CDM) that can simulate fracture of hydride, zirconium matrix, and Zr/hydride interface. CDM is possible to simulate the hydride, zirconium matrix, and interface cracking in a continuum model based on cladding deformation. The RCT simulation model was constructed from the microscopic images of irradiated cladding. A pixel-based finite element model was created by separating the hydride, zirconium matrix, and interface using the image segmentation method on a morphology operation basis. The appropriate element size was selected for the efficiency of the analysis and crack propagation using CDM. The force-displacement curves and strain energy from RCT were compared and analyzed with the simulation results of different element sizes. The finalized RCT simulation model can be used to evaluate the fracture resistance of the irradiated cladding under the quantified pinch load and to establish the failure criterion of fuel rods under pinch load. The advantages and limitations of the proposed process are discussed.

Zircaloy-4 is utillzed in nuclear fuel rod cladding due to its strength and corrosion resistance. However, it can undergo deformation over time, known as creep, which poses a safety risk in reactors. Furthermore, hydrogen absorption during reactor operation can alter its properties and affect creep rates. Previous research suggests a trend in which hydrogen concentration corelates unidirectionally with creep rates, either increasing or decreasing as the concentration rises. This trend can also be observed in EPRI’s creep model, EDF-CEA Model-3. However, recent literature has suggested that creep behavior may vary depending on the state of hydrogen presence. Therefore, it has become evident that creep behavior can be influenced not only by hydrogen concentration but also by the state of hydrogen presence, whether it is in a solid solution state or precipitated as hydrides. Our study aimed to compare creep behavior in specimens with hydrogen concentrations below and above solubility limits. We fabricated Zircaloy-4 plate specimens with varying hydrogen concentrations and conducted creep tests. The results revealed that specimens below the solubility limit exhibited decreasing creep rates as hydrogen concentration increased, while those above the limit displayed increasing creep rates. This investigation confirms that the state of hydrogen presence significantly impacts creep behavior within Zircaloy-4 cladding. As part of our additional research plans, we intend to conduct creep tests on the material based on its orientation, whether it is in the rolling direction (RD) or the transverse direction (TD). We also plan to perform creep tests on ring specimens. Additionally, for the ring specimens, we aim to evaluate how creep behavior differs between the cold-worked stress-relieved (CWSR) condition and the recrystallized annealed (RXA) condition achieved through high-temperature heat treatment.

In the case of dry storage facilities, slipping of the cask or tip-over are dangerous phenomena. For this reason, in dry storage facilities, measures against slipping and tip-over or related safety evaluations are important. Accidental conditions that can cause cask slippage and tip-over in dry storage facilities include natural phenomena such as floods, tornadoes, tsunamis, typhoons, earthquakes, and artificial phenomena such as airplane crashes. However, among natural phenomena, earthquakes are the most important natural phenomenon that causes tip-over. Also, many people had the stereotype that Korea is an earthquake-safe zone before 2016. However, earthquakes become a major disaster in Korea due to the 2016 Gyeongju earthquake and the 2017 Pohang earthquake, followed by the Goesan earthquake in October 2022. In this paper, seismic analysis was performed based on dry storage facilities including multiple casks. Design variables for the construction of an analysis model for dry storage facilities were investigated, and seismic analysis was performed. To evaluate tip-over accident during earthquake, seismic load was used from 0.2 g PGA to 0.8 g PGA and these earthquakes were followed Design Response Spectrum (DRS) in RG 1.60. The friction coefficient of concrete pad was used from 0.2 to 1.0. As a result of the analysis, tip-over accident could not find in the analysis from 0.2 g to 0.6 g. However, tip-over was appeared at friction coefficients of 0.8 and 1.0 at 0.8 g PGA. Tip-over angular velocity of cask was derived by seismic analysis and was compared with formula and tip-over analysis results. As a result, a generalized dry storage facility analysis model was proposed, and dry storage facility safety evaluation was conducted through seismic analysis. Also, tip-over angular velocity was derived using seismic analysis for tip-over analysis.

Korea Hydro & Nuclear Power (KHNP) is currently developing a vertical concrete dry storage module for the dry storage of used nuclear fuel within nuclear power plants. This module is designed with a structure consisting of cylinders, which can block the ingress of external air, thereby preventing Chloride-Induced Stress Corrosion Cracking (CISCC). However, due to the presence of these cylinder structures, unlike conventional dry storage systems, it cannot directly dissipate heat to the external atmosphere, making thermal evaluation an important issue. The SF dry storage module being developed by KHNP is a massive concrete structure of approximately 20 m × 10 m × 7 m in size, employing a vertical storage system. To demonstrate the safety of such a large structure, there is no alternative to conducting experiments with scaled-down models. Furthermore, according to NUREG-2215 Section 5.5.4, it is explicitly mentioned that design-verification testing can be performed using scaled-down models. In this paper, a 1/4 scaled-down model was constructed to perform thermal performance verification experiments, and the effectiveness of this model was analyzed using Computational Fluid Dynamics (CFD) methods. The analysis results indicated that there was not a significant difference in terms of maximum concrete temperature and air outlet temperature. However, a considerable difference was observed in the canister surface temperature. Therefore, it is concluded that careful consideration of natural convection heat transfer is necessary for the full application of the scaled-down model.