It is known that ZrCl4 can be used in the chlorination process of spent nuclear fuel. However, its solubility in high temperature molten salt is very small, making it difficult to dissolve a large amount of ZrCl4. Therefore, in this study, a flange-type sealed reactor was manufactured to observe the reaction characteristics of LiCl-KCl salt and ZrCl4. LiCl-KCl salt and ZrCl4 were placed in each alumina crucible, the reactor was sealed, and heated. The temperature at the reactor surface was above 500°C and maintained at that temperature for 48 hours. After completion of the reaction, the reactor was opened and the reaction products were recovered from each alumina crucible. The crystal structure of the reaction product was identified through XRD analysis, and the concentration of Zr was analyzed using ICP. Reaction characteristics were observed according to the molar ratio of ZrCl4 added to the number of moles of KCl in LiCl-KCl salt. The molar ratios of ZrCl4 to KCl were 0.5, 1, 2, and 3, respectively. As a result of each experiment, more than 95% of the injected ZrCl4 was vaporized and there was almost no residue in the ZrCl4 crucible. In the LiCl- KCl crucible, the weight increased in proportion to the amount of ZrCl4 added. As a result of XRD analysis, K2ZrCl6 was confirmed in all LiCl-KCl salt product. When the ZrCl4/KCl molar ratio was 2 and 3, LiCl-KCl could not be confirmed. Additionally, when the ZrCl4/KCl molar ratio was 1, LiCl was identified, but KCl was not found. Almost all of the KCl appears to have reacted with ZrCl4. ICP analysis results showed that the Zr concentration was proportional to the amount of ZrCl4 added in each LiCl-KCl salt, and exceeding the number of moles of reaction with KCl in the LiCl-KCl salt was observed. Therefore, these experimental results showed that ZrCl4 can be dissolved in LiCl-KCl salt at a maximum concentration higher than its solubility.

The separation efficiency of nuclides in molten salt systems was investigated, with a focus on the influence of apparatus configuration and experimental conditions. A prior study revealed that achieving effective Sr separation from simulated oxide fuel required up to 96 hours, reaching a separation efficiency of approximately 90% using a static dissolution reaction in a porous alumina basket. In this study, we explored the impact of agitation on improving Sr separation efficiency and dissolution rates. The simulated oxide fuel composition consisted of 2wt% Sr, 3wt% Ba, 2wt% Ce, 3wt% Nd, 3wt% Zr, 2wt% Mo, and 89wt% U. To quantify the Sr concentration in the salt, we utilized ICP analysis after salt sampling via a dip-stick technique. Furthermore, we conducted ICPOES analysis over a 55-hour duration to assess the separated nuclides. Complementing these analyses, SEM and XRD investigations were performed to validate the crystal structure and morphology of the oxide products.

Korea Atomic Energy Research Institute’s Post Irradiated Examination Facility safely stores spent nuclear fuel using a wet storage method to conduct research. Here, in order to remove the radioactivity released into the water, the stored water is passed through an ion exchange resin tower, and the radionuclides are exchanged with the bead-shaped ion exchange resin filled inside to lower the radioactivity concentration. At this time, because the stored water passes in one direction, clogging of the ion exchange resin occurs. If this phenomenon continues, the flow rate of the water treatment process decreases and operation efficiency decreases, so a backwashing process is necessary to re-mix the ion exchange resin and secure the flow rate again. In this study, the flow rate reduction trend according to the lifespan of the ion exchange resin and the flow rate recovery according to the backwash process operation amount were analyzed. The flow rate reduction trend of the ion exchange process was analyzed immediately after the backwashing process was started. In addition, the amount of flow recovery according to the backwash process operation amount was evaluated by the amount of waste generated during the backwash process and the number of days of operation until the backwash process was needed again. As a result, the flow rate of the ion exchange process decreased rapidly right after the backwash process until the position of the ion exchange resins was stabilized, and then stabilized. After that, it gradually decreased and reached the point where the backwash process was necessary. However, the decline trend was analyzed to be the same regardless of the lifespan of the ion exchange resin. In addition, the amount of waste generated during the operation of the backwash process was increased in the order of 400 L, 600 L, 1,100 L, 1,400 L, 3,500 L, and 4,200 L to increase the amount of operation of the backwash process. As a result, the number of days of ion exchange resin operation was 285 days, 338 days, and 342 days, was analyzed as 422 days, 322 days, and 720 days. Based on this study, it was confirmed that the flow rate reduction trend is the same regardless of the lifespan of the ion exchange resin, and as the backwash process operation increases, the number of days the ion exchange process can be operated increases, but there is a turning point where the waste treatment cost exceeds the number of days of operation.

Molten salt reactor (MSR) uses fluoride or chloride based molten salt as a coolant of the system, and fuel materials are dissolved in the molten salt, therefore it can be act as both coolant and nuclear fuel. A few issues have arisen from early-stage research and development program of MSR from Oak Ridge National Laboratory, including corrosion of structural materials and fission product management. For investigating the effect of additives on corrosion of structural materials, Mg(OH)2 and MgCl2*6H2O are added into the NaCl-MgCl2 eutectic salt. Prepared chloride salt is injected into the autoclave in the glove box, as well as corrosion coupons for candidate structural materials for molten chloride salt reactor, SS316, Alloy 600, and C-276 are also prepared. The temperature is set as 700°C. After 500 h corrosion experiment, the samples are taken out from the autoclave, and they are analyzed with scanning electron microscopy (SEM) and energy-dispersive X-ray spectroscopy (EDS). SS316 samples show weight loss with all salt conditions, while Alloy 600 and C-276 show weight gain after the corrosion experiment.

In KNF, fuel performance analysis modules were developed to predict the overall behavior of a fuel rod under normal operating conditions. Their main focus is to provide information on initial conditions prior to dry storage. Potential degradation mechanisms that may affect sheath integrity of spent CANDU 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 modules 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, identify and extend the ranges of all modules to required operating ranges. The 300°C spent CANDU fuel sheath temperature metric for dry storage ensures spent CANDU fuel element integrity from the failure mechanisms of creep rupture, oxidation and stress corrosion cracking at a failure probability of 2×10-5 for a dry storage time of 100 years. The 300°C sheath temperature metric for dry storage has relatively a lower failure rate than the target criteria for dry storage of spent LWR fuel. Although different modes of failure were treated separately for simplicity, ignoring possible synergistic effects, these results are conservative because of the conservative assumptions that have been made for evaluating spent fuel element conditions, and because of the inherent conservatism of the applied models. Additional conservatism of the model comes from the fact that isothermal conditions do not prevail in actual storage conditions. Further R&D being considered includes acquisition of new functional models to implement overall fuel behavior evaluation and cover spent CANDU fuel in dry storage, and upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The developed modules provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for spent CANDU fuel.

Various disposal methods for spent nuclear fuels (SNFs) are being researched, and one of these methods involves separating high heat-generating nuclear isotopes such as Strontium-90 (90Sr) and Cesium-137 (137Cs) for deep disposal. These isotopes has relatively short half-lives and substantial decay energies. Especially, 90Sr undergoes decay through Yttrium-90 to Zirconium-90, emitting intense heat with beta radiation. Therefore, the removal of these high heat-generating isotopes will significantly contribute to reducing disposal site area. To remove 90Sr from SNFs, molten salt was utilized in KAERI. During this process, it was discovered that 90Sr dissolves in the molten salt in the form of SrCl2 and/or Sr4OCl6. Afterwards, it is crucial to recover 90Sr in the form of oxide from the salt to create immobilized forms for disposal. This can be achieved by reactive distillation with K2CO3. However, the amount of 90Sr within the SNFs is only 0.121wt%, and even if all the 90Sr in the SNFs were to leach into the molten salt, the quantity of 90Sr in the molten slat would still be very small. Therefore, adding K2CO3 to the molten salt for reactive distillation could result in significant possibilities of side reactions occurring. In this study, a two-step process was employed to mitigate the side reactions: the 1st step involves evaporating the all molten salts and the 2nd step includes adding K2CO3 to make oxides through solid-solid reaction. Eutectic LiCl-KCl, which is the most commonly used salt, was employed. The eutectic LiCl-KCl with SrCl2 was heated at 850°C for 2 h to evaporate the salts under a vacuum (> 0.02 torr). However, after examining the distillation product before the solid-solid reaction, it was observed that SrCl2 reacted with KCl in the salt, resulting in the formation of KSr2Cl5. It means that salts containing KCl are not suitable candidates for reactive distillation aimed at producing immobilized forms. As an alternative, MgCl2 could be a highly promising candidate because it is inert to SrCl2 and according to a recent study in KAERI, MgCl2 exhibited the most efficient separation of Sr among various salts. Therefore, we plan to proceed with the two-step reactive distillation using MgCl2 for the future work.

It has been investigated on the management of Strontium-90 in KAERI. It is needed to separate the solute from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. A vacuum distillation technology was used for the separation of strontium from the molten salt in our previous study. Strontium chloride was successfully carbonated by reactive distillation of SrCl2 – K2CO3 – LiCl – KCl system. In this study, it was tried to develop another route to recover strontium from the salt solution by a solid-solid reaction for avoiding the entrainment of product and the salt-K2CO3 reaction. Reactive distillation experiments were carried out for SrCl2 - K2CO3 – LiCl – KCl system. The carbonation temperature and pressure were 520°C and 0.8 bar. After the carbonation reaction, the temperature was elevated to 820°C to remove KCl from the reaction product. SrCO3 and KCl peaks were found in the XRD analysis of the residual product. It could be concluded that SrCl2 can be successfully carbonated after salt removal by the solid-solid reaction.

This study investigated the effectiveness of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, the effects of using MgCl2, NH4Cl, and Cl2 as a single chlorinating agent or applying MgCl2, NH4Cl, and Cl2 sequentially for spent fuel chlorination were evaluated Furthermore, in this study, assuming the actual process operation situation, where only a part of the semi-volatile nuclides is removed during the heat treatment process, and including the process of precipitating the molten salt from the chlorination process with K3PO4 and K2CO3 precipitants, the percentage distribution of 50 nuclides in the light water reactor spent fuel into each process stream was quantitatively calculated using the simulation function of the HSC program and tabulated for intuitive viewing. Compared to a single chlorinator, sequential chlorination more effectively separated the heat and radioactivity of the spent fuel from the uranium-dominated product solids. Specifically, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore that sequential chlorination can be an effective method for spent fuel partitioning, either as a standalone approach or in combination with other partitioning processes such as pyroprocessing.

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.

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.

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.

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.

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.

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.

An Internal Compliance Program (ICP) is a system through which enterprise internally manage their own export control processes to ensure compliance with domestic export control laws. Around the world, ICPs are actively utilized as a means of export control for strategic items. However, they are not mostly applied to the Trigger List Items. However, advanced countries such as the United States and the Nuclear Suppliers Group (NSG) have been actively researching the potential application of ICPs to the Trigger List Items recently. This paper suggests additional considerations that should be taken into account when applying an ICP to the Trigger List Items. The key elements of classical ICP include Top-level management commitment to compliance; Risk analysis; Organizational structure/chain of responsibilities; Human and technical resources allocated to the management of exports; Workflow management and operational procedures; Record -keeping and documentation; Selection of staff; training and awareness-raising; Process-/Systemrelated controls (ICP audit)/Corrective Measures; Physical and technical security. An ICP for Trigger List Items must encompass all these core elements. Additionally, as the nuclear industry often involves collaborative projects participating with various companies, the effectiveness of the ICP could be enhanced through the operation of consultation groups among participating companies. Furthermore, enterprises must take into account the unique characteristics of Trigger List Items that differ from other strategic items, when making requirements of the ICP establishment. First, export requirements related to safety measures and physical protection should be reviewed to export the Trigger List Items. The procedure and obligations in aspects of internationally controlled items should also be reviewed. Moreover, active support from enterprises for GTGA procedures should also be included, since the Government to Government Assurance (GTGA) procedure is additionally required for the export of Trigger List Items, in contrast to other strategic items. Additionally, for materials categorized within Trigger List Items, such as deuterium and heavy water, should be controlled based on their end-use and cumulative quantity, which Government cannot effectively manage without enterprise supports. Therefore, enterprises must establish an internal material management system based on the end-use and cumulative quantity of these materials under ICP.

The Democratic People’s Republic of Korea (DPRK) has been exporting weapons of mass destruction (WMD) to the volatile Middle East and Africa. It is expecting that military illicit activities would isolate DPRK economically, as it has been placed on multiple sanctions lists, including UN sanctions, multilateral export control regime sanctions, and country-specific sanctions. However, DPRK funds its WMD programs through various sanctions evasion activities. DPRK’s primary sanctions evasion activities include obtaining foreign currency, acquiring dualuse or restricted technology, smuggling, and money laundering, which are global in scope. This study analyzes the sanctions evasion activities used by DPRK to acquire economic, material, and technological resources for its WMD program and devises ways to disrupt these evasions effectively. First, the international community should strengthen export controls by encouraging states with weak export control regimes to join international organizations and conventions to limit DPRK arms and technology exports. Second, states should improve their intelligence gathering and analysis capabilities by sharing information on DPRK’s evasion activities and working together. This will help identify and counter DPRK’s evasion techniques and networks. Third, the international community should strengthen cooperation on DPRK’s evasion efforts. This can be done by strengthening cooperation with states and entities that enforce international sanctions and by working with relevant agencies such as customs, immigration, and police to track and interdict the movement of funds and assets used for evasion. Fourth, publicize DPRK’s illicit activities and apply diplomatic pressure. Diplomatic pressure can lead to more states and entities to enforce sanctions. In conclusion, these strategies are expected to deter DPRK’s illicit activities; but to sanction DPRK effectively, it is essential to continue to adjust and refine the strategies in response to DPRK’s evolving sanctions evasion efforts. The results of this research are expected to prevent WMD proliferation through DPRK by blocking or reducing the risk of sanctions evasion.

As drones expand beyond military purposes to the private sector, the level of use of drones in various fields is increasing. However, the world was shocked by the attempt to attack with a drone equipped with a C4 bomb in the US and the attempt to assassinate a head of state using a drone in Venezuela. Drone threats to domestic nuclear power plants are also increasing due to the expansion of drone use, terrorist threats, and North Korea’s invasion of drones. Overseas, various drone threats to nuclear power plants have occurred. In October 2014, French electricity company Electricite de France confirmed that it had observed unauthorized drones over seven nuclear power plants across France. A drone threat occurred at the Savannah River Site (SRS), a U.S. Department of Energy facility that processes and stores nuclear materials. In 2016, eight drones were observed by security personnel. In 2016, a drone flew over the cooling tower of the Liebstadt nuclear power plant in Switzerland, and publicly shared the filmed video on YouTube. In July 2018, Greenpeace activists intentionally crashed a drone into the outer wall of the spent fuel building in Boughey, France. In January 2019, they used drones to drop smoke bombs and release videos at Orangeo’s nuclear facility containing irradiated fuel. In January 22, Sweden saw drones flying over three nuclear power plants. Drone was also seen at the Forsmark nuclear power plant on Friday and at two other Swedish nuclear power plants in Oskarshamn and Ringhals on Monday. Anti-drone technology to counter the threat of drone terrorism is also developing. Anti-drone technology detects, tracks, and identifies illegal drones to neutralize them. Various technologies such as radar, EO/IR cameras, Lidar, sensor, and RF scanners are being developed for drone detection. Depending on the detection technology, it has advantages such as detection distance and remote control drone detection. However, there are also disadvantages, such as obstacles, weather condition, and the inability to detect drones that do not transmit signals. Methods such as jammer, capture, and destruction have been developed for incapacitation technology. While it has advantages such as stability and portability, it has disadvantages such as limited use and damage to the surroundings. Accordingly, it is necessary to draw realistic measures to defend against the threat of nuclear power plants by paying constant attention to the various detection, identification, and neutralization anti-drone systems that continue to evolve.

The Republic of Korea (ROK), as a member state of the IAEA, is operating the State’s System of Accounting for and Control (SSAC) and conducting independent national inspections. Furthermore, an evaluation methodology for the material unaccounted for (MUF) is being developed in ROK to enhance capabilities of national inspection. Generally, physical and chemical changes of nuclear material are unavoidable due to the operating system and structure of facilities, an accumulation of material unaccounted for (MUF) has been issued. IAEA developed statistical MUF evaluation method that can be applied to all facilities around the world and it mainly focuses on the diversion detection of nuclear materials in facilities. However, in terms of the national safeguard inspection, an evaluation of accountancy in facilities is additionally needed. Therefore, in this research, a new approach to MUF evaluation is suggested, based on the Guide to the Expression of Uncertainty in Measurement (GUM) that an evaluation of measurement uncertainty factors is straightforward. A hypothetical list of inventory items (LII) which has 6,118 items at the beginning and end of the material balance period, along with 360 inflow and outflow nuclear material items at a virtual fuel fabrication plant was employed for both the conventional IAEA MUF evaluation method and the proposed GUM-based method. To calculate the measurement uncertainty, it was assumed that an electronic balance, gravimetry, and a thermal ionization mass spectrometer were used for a measurement of the mass, concentration, and enrichment of 235U, respectively. Additionally, it was considered that independent and correlated uncertainty factors were defined as random factors and systematic factors for the ease of uncertainty propagation by the GUM. The total MUF uncertainties of IAEA (σMUF) and GUM (uMUF) method were 37.951 and 36.692 kg, respectively, under the aforementioned assumptions. The difference is low, it was demonstrated that the GUM method is applicable to the MUF evaluation. The IAEA method demonstrated its applicability to all nuclear facilities, but its calculated errors exhibited low traceability due to its simplification. In contrast, the calculated uncertainty based on the GUM method exhibited high reliability and traceability, as it allows for individual management of measurement uncertainty based on the facility’s accounting information. Consequently, the application of the GUM approach could offer more benefits than the conventional IAEA method in cases of national safeguard inspections where factor analysis is required for MUF assessment.

This study aims to classify R&D activities related to the nuclear fuel cycle using the deep learning methodology. First, R&D data of the Republic of Korea were collected from the National Science & Technology Information Service (NTIS) for the years 2021, 2022, and 2023. We use keywords such as ‘nuclear,’ ‘uranium,’ ‘plutonium,’ and ‘thorium’ to find nuclear-related R&D projects in the NTIS database. Among the numerous R&D projects found through keyword searches, overlapping and medical-related R&D projects were excluded. Finally, 495 R&D projects conducted in 2021, 430 R&D projects conducted in 2022, and 296 R&D projects conducted in 2023 were obtained for analysis. After that, Safeguards experts determine whether the R&D projects are subject to declaration under the AP. The values of the content validity index (CVI) and content validity ratio (CVR) were used to verify whether the experts’ judgments were valid. The 1,218 collected and labeled data were then divided 8:2 into training and test datasets to see if deep learning could be applied to classify nuclear fuel cycle-related R&D activities. We use the Python and TensorFlow packages, including RNN, GRU, and CNN methods. First, the collected text information was preprocessed to remove punctuation marks and then tokenized to make it suitable for deep learning. After 20 epochs of training to classify the nuclear fuel cycle-related R&D activities, the RNN model achieved 97.30% accuracy and a 5.85% error rate on the validation dataset. The GRU model achieved 96.53% accuracy and a 9.06% error rate on the validation dataset. In comparison, the CNN model achieved 94.61% accuracy and a 2.57% error rate on the validation dataset. When applying the test dataset to each model, the RNN model had a test accuracy of 83.20%, the GRU test accuracy of 82.79%, and the CNN model had a test accuracy of 85.66% for the same dataset. This study applied deep learning models to labeled data judged by various experts, and the CNN model showed the best results. In the future, this study will continue to develop an optimum deep learning model that can classify nuclear fuel cycle-related R&D activities to achieve the purpose of safeguards measures from open-source data such as papers and articles.

Understanding the dispersion of xenon isotopes following a nuclear test is critical for global security and falls within the remit of both the Comprehensive Nuclear-Test-Ban Treaty (CTBT) and the International Noble Gas Experiment (INGE). This paper aims to show if it is possible to discriminate the source of xenon releases based on the atmospheric dispersion of xenon isotopes using HYSPLIT. Using ORIGEN and SERPENT simulations, four released scenarios are defined with four different fractionation times (i.e., 1 hour, 1 day, 10 days, and 30 days) after a 1kt TNT equivalent 235U explosion event. These time-delayed release scenarios were selected to certify the possibility of mis-determining xenon release source. We use the Lagrangian dispersion model for atmospheric dispersion to predict the concentration distribution of xenon isotopes under each scenario. The model allows us to better understand how these isotopes would distribute over time and space, offering valuable data for real-world detection efforts. To our knowledge, there have been no researches on the analysis of xenon isotopic ratios considering atmospheric dispersion. In this work, we focused on the atmospheric dispersion using HYSPLIT to characterize the xenon isotopic ratios from nuclear tests. In addition, we compared the xenon isotopic ratios obtained from the atmospheric dispersion with those from ORIGEN calculations, which would be helpful to discriminate the source of the xenon releases.