In order to ensure the long-term safety of a deep geological repository, the performance assessment of the Engineered Barrier System (EBS) considering a thermal process should be performed. The maximum temperature at the side wall of a disposal canister for the technical design requirement should not exceed 100°C. In this study, the thermal modelling was conducted to analyze the effects of the thermal process from a disposal canister to the surrounding near-field host rock using the PFLOTRAN code. The mesh was generated using the LaGriT code and the material properties were assigned by applying the FracMan code. Initial conditions were set as the average geothermal gradient (25.7°C/km) and an average surface temperature (14.7°C) in Korea. The highest temperature was observed at the middle of the canister side wall. The temperature of the buffer was lower than that of the canister, and the temperature increase of the deposition tunnel and the host rock was insignificant due to the lower effect of the heat source. The result of the thermal evolution of the EBS represented the highest thermal effects in the vicinity of the canister. In addition, the thermal effects were largely decreased after 10 years of the entire simulation period. It demonstrated that the model took 3 years to heat up the buffer around the canister. The temperature at the canister side wall increased until 3 years and then decreased after that time. This is because that the radioactive decay heat from the heat source was emitted enough to raise the overall temperature of the EBS by 3 years. However, the decay heat rate of the canister decreased exponentially with the disposal time and then its decay heat was not emitted enough after 3 years. In conclusion, the peak temperature results of the EBS were lower than 70°C to meet the technical design requirement.

Due to the necessity of isolating spent nuclear fuel (SNF) from the human life zone for a minimum of 106 years, deep geological disposal (DGD) has emerged as a prominent solution for SNF management in numerous countries. Consequently, the resilience of disposal canisters to corrosion over such an extended storage period becomes paramount. While copper exhibits a relatively low corrosion rate, typically measured in millimeters per million years, in geological environment, special attention must be directed towards verifying the corrosion resistance of copper canister welds. This validation becomes inevitable during the sealing of the disposal canister once SNFs are loaded, primarily because the weld zone presents a discontinuous microstructure, which can accelerate both uniform and localized corrosion processes. In this research, we conducted an in-depth analysis of the microstructural characteristics of copper welds manufactured by TIG-based wire are additive manufacturing, which is ideal for welding relatively large structures such as a disposal canister. To simulate the welds of copper canister, a 12 mm thick oxygen-free plate was prepared and Y and V grooves were applied to perform overlay welding. Both copper welding zones were very uniform, with negligible defects (i.e., void and cracks), and contained relatively large grains with columnar structure regardless of groove types. For improving microstructures at welds with better corrosion resistance, the effect of preheat temperature also investigated up to 600°C.

The mobility of uranium (U) in various disposal environments of a deep geological repository is controlled by various geochemical conditions and parameters. In particular, oxidation state of uranium is considered as a major factor to control the mobility of uranium in most of geological environments. In this study, therefore, we investigated the geochemical behaviors of uranium in grounwater samples from natural analogue study sites located in the Ogcheon Metamorphic Belt (OMB). Groundwater samples were taken using a packer system from Boeun Hoenam-myun site and Geumsan Suyoung-ri site where several boreholes were dilled with various depths. The geochemical properties and parameters such as temperature, pH, Eh, EC, and DO were directly measured in the site using an in-line measurement method. The concentrations of major cations and anions in the groundwater samples were measured by using ICP-OES (Inductively Coupled Plasma-Optical Emission Spectrometry) and IC (Ion Chromatography), respectively. The concentrations of trace elements including U and Th were measured by using ICP-MS (Inductively Coupled Plasma-Mass Spectrometry) The concentrations of U in the groundwater samples are very low for the Hoenammyun site (0.03~0.69 ppb) and Suyoung-ro site (0.39~1.74 ppb) even though the two sites are uranium deposits and redox conditions are weakly oxidizing. The speciation, saturation index (SI), pH-Eh (Poubaix) diagram were calculated using the Geochemist’s Workbench (GWB 9.0) program and the recent OECD/NEA thermochemical database for U. Calculation results for U speciation in the groundwater samples show that major dissolved uranium species in the groundwater samples are mainly as calcium uranyl carbonate complexes such as Ca2UO2(CO3)3(aq) and CaUO2(CO3)3 2- for almost all groundwater samples. The calculated results for SI and Poubaix diagram also show that the dominant uranium solid phase is a uranyl silicate mineral, uranophane (Ca(H2O)(UVIO2)2 (SiO2)2(OH)6), not uraninite (UIVO2). Since the determination of Eh values for natural groundwater samples is very difficult and uncertain work, we analyzed and discussed the effect of Eh on the geochemical behaviors of U in the groundwater. However, these calculation results are not consistent with the observation for U minerals in rock samples using electron microscopic techniques. Thus, we need further studies to explain the discrepancy between calculation and observation results.

This study aimed to provide better understanding of the bedrock aquifer bacterial communities and their functions in deep geological repository (DGR) environment. Two study sites of uranium deposits in the Ogcheon Metamorphic Belt were selected: Boeun and Guemsan. From two study sites, six groundwater samples were obtained with different boreholes and depths: OB1 (Boeun, 25 m), OB3 (Boeun, 80 m), GS1 (Guemsan, 25 m), GS2 (Guemsan, 85-90 m), GS3-I (Guemsan, 32- 38 m), GS3-II (Guemsan, 70-74 m). The physicochemical properties of groundwater were analyzed by multi-parameter sensors, ion chromatography (IC), and inductively coupled plasma optical emission spectroscopy (ICP-OES). Illumina Miseq sequencing was performed to investigate bacterial community in six groundwater samples. In addition, the number of sulfate-reducing bacteria (SRB) was quantified by a quantitative PCR (qPCR). Bacterial community composition varied in response to boreholes and depths. A total of 14 different phyla and 36 classes were detected from six groundwater samples. Overall, Proteobacteria, Actinomycetota, and Bacteroidota were dominant in the phylum level. SRB and iron-reducing bacteria (IRB) were detected in all groundwater samples even though organic carbon sources were not abundant (0.7-3.3 mg-total organic carbon/L). This result shows a potential to immobilize uranium in DGR environment. In particular, SRB, Desulfosporosinus fructosivorans and Humidesulfovibrio mexicanus were mainly detected in GS1 and GS2 groundwater samples, which attributed to higher dissimilatory sulfite reductase functional gene copy number in GS1 and GS2 groundwater samples. Statistical analysis was performed to understand the correlation between environmental factors and core bacterial species. Dissolved oxygen (DO), Fe, and Mn concentrations were positively correlated with Curvibacter fontanus while Undibacterium rivi had a negative correlation with pH. These results indicate that bacterial community could be changed in response to environmental variation. Further study with a greater number of samples is necessary to obtain statistically reliable and meaningful results for a safe DGR system.

Buffer materials play an important role in preventing the leakage of radionuclides from the residue. The mineralogical properties of these buffer materials are critical in repository design. This study presents the fundamental properties of Na-type MX80 and a novel Ca-type Bentonil- WRK. The CaO to MgO ratio in Bentonil-WRK was approximately 1:1, and the CaO to Na2O ratio was approximately 2.8:1. These results suggest that Bentonil-WRK demonstrates a lower swelling index compared to Gyeongju bentonite due to its CaO-to-MgO ratio’s proximity to 1:1, despite having a higher montmorillonite content than Gyeongju bentonite. The results of this research can provide useful foundational data for the evaluation of the thermal-hydraulic-mechanical-chemical behavior of buffer materials.

In this study, a third metal layer with a higher corrosion potential than copper was introduced between the copper and cast iron layer to strengthen the corrosion resistance of the copper layer which is considered as a corrosion resistant barrier in the disposal container for spent nuclear fuel. Three types of corrosion-resistant metals, silver, nickel, and titanium, were selected as the intermediate insertion layer, and the galvanic specimens of two bonded metals were exposed to KURT (KAERI Underground Research Tunnel) groundwater and a high voltage of 1.0 V was applied to corrode the specimens at electrochemically accelerated condition. Corrosion of copper part was confirmed in Cu-Ti, Cu-Ni, and Cu-Ag galvanic specimens, but copper part was not corroded in Cu-Fe galvanic specimen. If the corrosion-resistant intermediate layer proposed in this study works properly, the local corrosion problem of copper disposal canister is expected to be some degree solved, which can apply to a welding part or a stress concentrated part.

The thermal evaluations for the conceptual design of the deep geological repository considering the improved modeling of the spent fuel decay heat were conducted using COMSOL Multiphysics computational program. The maximum temperature at the surface of a disposal canister for the technical design requirement should not exceed 100°C. However, the peak temperature at the canister surface should not exceed 95°C considering the safety margin of 5°C due to several uncertainties. All thermal evaluations were based on the time-dependent simulation from the emplacement time of the canister to 100,000 years later. In particular, the heat source condition was set to the decay heat rate and axial decay heat profile of the PLUS7 fuel with 4.0wt% U-235 and 45 GWD/MTU. The thermal properties of the granitic rock in South Korea were applied to the host rock region. For the reference design case, the cooling time of the SNF was set to 40 years, the distance between the deposition holes 8 meters and that between the deposition tunnels 30 meters. However, the peak temperature at the canister surface at 10 years was 95.979°C greater than 95°C. This design did not meet the thermal safety requirement and needed to be modified. For the first modified case, when the distance between the deposition tunnels was set to 30 meters, three cooling time cases of 40, 50 and 60 years and five distances of 6, 7, 8, 9 and 10 meters between the deposition holes were considered. The design with the distances of 9 and 10 meters between the deposition holes for the cooling time of 40 years and all five distances for 50 and 60 years were less than 95°C. For the second modified case, when the distance between the deposition holes was set to 8 meters, three cooling time cases of 40, 50 and 60 years and five distances of 20, 25, 30, 35 and 40 meters between the deposition tunnels were considered. The design with the distances of 35 and 40 meters between the deposition tunnels for the cooling time of 40 years, the distances of 25, 30, 35 and 40 meters for 50 years and all five distances for 60 years were less than 95°C. As a result, the peak temperature at the canister surface decreased as the cooling time and the distance between the deposition holes and the tunnels increased.

The WRK (Waste Repository Korea bentonite) compacted bentonite medium has been considered as the appropriate buffer material in the Korean SNF (Spent nuclear fuel) repository site. In this study, hydraulic properties of the WRK compacted bentonite core (4.5 cm in diameter and 1.0 cm in length) as the buffer material were investigated in laboratory experiments. The porosity and the entry pressure of the water saturated core at different confining pressure conditions were measured. The average velocity of water flow in the WRK compacted bentonite core was calculated from results of the breakthrough curves of the CsI aqueous solution and the hydraulic conductivity of the core was also calculated from the continuous flow core experiments. Because various gases could be generated by continuous SNF fission, container corrosion and biochemical reactions in the repository site, the gas migration property in the WRK compacted bentonite core was also investigated in experiments. The gas permeability and the average of gas (H2) in the core at different water saturation conditions were measured. Laboratory experiments with the WRK Compacted bentonite core were performed under conditions simulating the DGR environment (confining pressure: 1.5- 20.0 MPa, injection pressure: 1.0-5.0 MPa, water saturation: 0-100%). The WRK Compacted bentonite core was saturated at various confining pressure conditions and the porosity ranged from 27.5% to 43.75% (average: 36.75%). The calculated hydraulic conductivity (K) of the core using experimental results was 8.69×10-11 cm/s. The gas permeability of the core when the water saturation 0~58 % was ranged of 19.81~3.43×10-16 m2, representing that the gas migration in the buffer depends directly on the water saturation degree of the buffer medium. The average gas velocity in the core at 58% of water saturation was 9.8×10-6 m/s, suggesting that the gas could migrate fast through the buffer medium in the SNF repository site. Identification of the hydraulic property for the buffer medium, acquired through these experimental measurements is very rare and is considered to have high academic values. Experimental results from this study were used as input parameter values for the numerical modeling to simulate the long-term gas migration in the buffer zone and to evaluate the feasibility of the buffer material, controlling the radionuclide-gas migration in the SNF repository site.

Raman characteristics of various minerals constituting natural rocks collected from uranium deposits in Okcheon metamorphic zone in Korea are presented. Micro-Raman spectra were measured using a confocal Raman microscope (Renishaw in Via Basis). The focal length of the spectrometer was 250 mm, and a 1800 lines/mm grating was installed. The outlet of the spectrometer was equipped with a CCD (1,024256 pixel) operating at -70°C. Three objective lenses were installed, and each magnification was 10, 50, and 100 times. The diameter of the laser beam passing through the objective lens and incident on the sample surface was approximately 2 m. The laser beam power at 532 nm was 1.6 mW on the sample surface. Raman signal scattered backward from the sample surface was transmitted to the spectrometer through the same objective lens. To accurately determine the Raman peak position of the sample, a Raman peak at 520.5 cm-1 measured on a silicon wafer was used as a reference position. Since quartz, calcite, and muscovite minerals are widely distributed throughout the rock, it is easy to observe with an optical microscope, so there is no difficulty in measuring the Raman spectrum. However, it is difficult to identify the uraninite scattered in micrometer sizes only with a Raman microscope. In this case, the location of uraninite was first confirmed using SEM-EDS, and then the sample was transferred to the Raman microscope to measure the Raman spectrum. In particular, a qualitative analysis of the oxidation and lattice conditions of natural uraninite was attempted by comparing the Raman properties of a micrometer-sized natural uraninite and a laboratory-synthesized UO2 pellet. Significantly different T2g/2LO Raman intensity ratio was observed in the two samples, which indicates that there are defects in the lattice structure of natural uraninite. In addition, no uranyl mineral phases were observed due to the deterioration of natural uraninite. This result suggests that the uranium deposit is maintained in a reduced state. Rutile is also scattered in micrometer-sizes, similar to uraninite. The Raman spectrum of rutile is similar in shape to that of uraninite, making them confused. The Raman spectral differences between these two minerals were compared in detail.

Uranium (U) is a hazardous material that can lead to both chemical and radiological toxicity, including kidney damage and health issues associated with radiation exposure. In South Korea. In Korea, where shallow weathered granitic aquifers are widespread, several previous studies have reported high levels of radioactivity in shallow groundwater. This ultimately led to the closure of 60 out of 4,140 groundwater production wells in South Korea. In this study, we examined aquifers currently dedicated to drinking water supply and investigated a dataset of 11,225 records encompassing 103 environmental parameters, based on the random forest classifier. This dataset comprises 80 physical parameters associated with the hydraulic system and 23 chemical parameters linked to water-rock interactions. Among the hydraulic parameters, the presence of a coarse loamy texture in the subsoil displayed a notable positive relationship with the concentration of uranium, implying that it plays a significant role in forming redox conditions for the leaching of uranium from host rocks. Fluorine (F), a major product of water-rock interaction in granitic aquifers, exhibited a positive correlation with the distribution of uranium concentrations. The positive relationship between F concentration and uranium levels suggests that the dissolved uranium originates from groundwater interacting with granites. In conclusion, our findings indicate that two key factors, namely the infiltration capacity of soil layers and the aqueous speciation in groundwater resulting from interactions with local solids, play important roles in determining uranium concentrations in granitic aquifers.

Spent nuclear fuels (SNFs) are stored in nuclear power plants for a certain period of time and then transported to an interim storage facility. After that, SNFs are finally repackaged in a disposal canister at an encapsulation plant for final disposal. Finland and Sweden have already completed the design of the spent nuclear fuel encapsulation plant. In particular, Finland has begun the construction of the encapsulation plant and is on the verge of completion. Korea Radioactive Waste Agency (KORAD) is conducting a conceptual design of a deep geological repository for SNFs. Conceptual design of the encapsulation plant is part of the research activity. It is highly required to draft an operation process of the encapsulation plant before an actual design activity. As part of the activity, Finnish design concept of the encapsulation plant and experience were thoroughly reviewed. Finally a preliminary concept of the operation process was proposed considering Korean unique situations such as the volume of SNFs estimated to be disposed of, types of transportation cask and other considerations.

Recently, as carbon-neutral energy sources become increasingly important worldwide, SMRs (Small Modular Reactors), which offer significantly enhanced safety, versatility, and mobility compared to conventional nuclear reactors, are gaining attention as a viable alternative. SMR generally refers to small modular reactors with a power output of 300 MWe or less. Unlike conventional reactors, SMRs are characterized by an all-in-one design where peripheral systems and equipment are all integrated into the reactor itself, leading to enhanced reliability and durability. Additionally, the nuclear fuel reloading cycle is significantly extended compared to traditional reactors, resulting in a substantial reduction in maintenance difficulty and costs. Researchers have taken note of these characteristics of SMRs, particularly the extended fuel reloading cycle. Therefore, we have initiated the initial design of an ultra-small Micro Modular Reactor with an electricity generation capacity of 10 MWe and a fuel cycle of up to 55 years, with the goal of using it as a propulsion power source for various transportation modes, especially ships. Our design of MMR, called ‘ARA,’ is primarily distinguished by its use of U233 and Th232 fuels instead of conventional UO2 fuel. Due to various features of ‘ARA,’ including different fuel compositions, ARA is predicted to exhibit several characteristic features compared to conventional PWRs. In this study, among these characteristics, we focused on predicting changes in material composition within the fuel rod during the extended cycle operation of high-enriched fuel, rather than short-cycle operation using low-enriched fuel, unlike conventional reactors. The primary goal of this research is to observe the behavior of the composition of the materials used in the fuel cycle of the MMR, which utilizes U233 and Th232 fuels instead of UO2. Considering the difficulties in the spent nuclear fuel disposal process, many different trials were made to minimize the fission products of ARA, which differs from conventional reactors in terms of fuel type, size, and fuel cycle, in relation to waste generation.

Concrete structures of spent nuclear fuel interim storage facility should maintain their ability to shield and structural integrity during normal, off-normal and accident conditions. The concrete structures may deteriorate if the interim storage facility operates for more than several decades. Even if deterioration occurs, the concrete structures must maintain their own functions such as radiation shielding protection and structural integrity. Therefore, it is necessary to establish an analysis methodology that can evaluate whether the deteriorated concrete structure maintains its integrity under not only normal or off-normal condition but also accident condition. In this study, dynamic material testing was conducted on concrete cores extracted from HANARO exterior wall during seismic reinforcement construction. HANARO was constructed at the Korea Atomic Energy Research Institute in 1995, following strict nuclear quality assurance standards. In order to conduct the dynamic material testing of the extracted concrete cores, self-disposal had to be performed because the concrete cores were extracted and stored in a radiation controlled area. A self-disposal application was prepared and submitted based on the radionuclide analysis results, and it was finally approved in April 2023. Then, a test was performed by processing a specimen for dynamic property testing using a self-disposed concrete core. The concrete cores were processed to create specimens for dynamic material testing and the dynamic material testing was performed to obtain stress-strain diagrams according to the strain rate.

One of released radioactive gases from a spent fuel is cesium (137Cs) as semi-volatile fission products and reticulated ceramic foam could be used for capturing the gaseous cesium. It has threedimensional open-pore structures and consumes cesium above 600°C to form cesium species including Cs-nepheline (CsAlSiO4) and pollucite (CsAlSi2O6) phases. Kaolinite-based foam filter is a favorable ceramic filter because they exhibit superior capture characteristics compared to other aluminosilicate minerals and other shape filters. However, for usage in special conditions, structural limitations such broken struts must be improved. Here, recoating by using centrifugation, followed by a pre-sintering cycle was conducted for covering the cracks and voids, resulting from the burnout of the polyurethane sponge as a sacricial template. The slurry including additives was chosen by considering viscous behavior of slurries for the centrifugation. The microstructure and strength was improved by the recoating.

Reticulated foams have a continuous skeleton network consisting of aluminosilicates and are used for capturing gaseous cesium released from spent nuclear fuel at high temperature. It has high stability to high temperature and good capturing performance. Homogeneous cell distribution and modified surface structures are indispensable conditions for stable operation and handling. In particular, triangularly shaped holes inside the struts were generated during the pyrolysis of polyurethane sponge as a sacricial template, which lead to limite the strength of the reticulated foam as well as cracks. However, several attempts have been focused on the increasing the strut thickness. Here, we have prepared ceramic foams by the polyurethane sponge replication method with roller squeezing. Ceramic slurry including additives was determined with consideration of its viscous behavior. After pre-sintering, infiltration under vacuum was conducted. Metakaolin slurry was filled partially into the triangular void. As a result, the compression strength was improved by structure modification without composition change.

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.

Strategic item export control aims to maintain international peace and safety and serves as a significant nuclear non-proliferation regime that directly impacts a nation’s security. Therefore, establishing an autonomous export control system at the state level is crucial, and one of the most efficient methods to achieve this is by enhancing an export company’s management system. Accordingly, many advanced countries, such as the United States, Europe, and Japan, have operated their own internal compliance programs (CP or ICP) to manage and screen the export of strategic items as a corporate social responsibility and risk mitigation measure. In Korea, which has a high dependence on trade, the need for CP was continuously confirmed, but the system was introduced in 2004, relatively late compared to other advanced countries. So far, the Korean government has made steady efforts to develop and establish the system and is actively encouraging businesses to obtain Compliance Program certification to autonomously manage strategic items. Major technologically advanced countries utilize technology transfer as a tool for economic sanctions, trade security, and strategic technology management, and they continue to strengthen their control regimes. In these countries, CP certification is considered a standard practice for export control among mid-sized and large enterprises. It serves as a vital risk management system that protects companies from unforeseen incidents. However, in Korea, the application of CP under the Foreign Trade Act is limited to dual-use items and does not extend to the nuclear export control system. Therefore, this paper analyzes international cases and CP requirements in countries like the United States, Japan, Europe, and Singapore. As a result of the review, the application of CP into Korea’s nuclear export control could be a coexistence means that can strengthen supply chain control as well as provide benefits not to impede technical research, international trade, and exchanges.

In compliance with the amended export control of strategic items and technology in Jan. 2014, KAERI should pay attention to the export control of ITT (Intangible Technology Transfer). To control an ITT (Intangible Technologies Transfer) effectively and efficiently, the Korean government encourages the R&D institute and universities obtaining the ICP (Internal Compliance Program) from the relevant authority, MOTIE. This means that the exporters can control the ITT by themselves, because the exporters know very well the counterparts of the trading and the exporting items and technologies. In fact, ICP is for export control of dual-use items and technology in Korea. However, KAERI has tried to obtain a license from the authority, MOTIE. In an effort to do so, KAERI completed enacting a new internal self–regulation for export controls in 2016, and proceeded to apply for an ‘AA’ license of ICP in 2017 and obtained the ICP license in 2018 and re-obtained the license in 2021 from the MOTIE. In light of KAERI’s case, to obtain the ‘AA’ license of ICP is one of the best methods to increase the ability of export controls. As of now, there is no R&D institutes sponsored by the Korean government to obtain the ‘AA’ license of ICP except KAERI. KAERI can provide the actual methods as a standard case to the R&D institutes in Korea for obtaining an ‘AA’ license of ICP. According to the internal regulation of KAERI for export control, KAERI implemented an inner self-audit for export control in Nov. 2022. This is the first real self-audit for export control at KAERI. The main purpose of the self-audit is to check the transfer management of ITT and the relationship of relevant office through the interview of the staffs in the ICP organization. KAERI self-audit planed specifically and implemented for the achievement of the basic principal of selfaudit. The specific contents of this self-audit is as follows - The interview of the relevant offices: physical protection office, manpower planning office, manpower management office, nuclear education and training center, technology transfer office and international cooperation office, nuclear control and management office - Building the self-audit checklist considering the characteristics of each office - The confirmation of the inner procedure and the status of management on the export controls Through the interview of the relevant office, KAERI checked the inner procedure and the status of management on the export controls and tried to provide the supplementary measures of each relevant offices. The followings are the main results of the inner self-audit implemented in Nov. 2022. - Generally, the staffs know the meaning and relevant regulation such as foreigner’s management and the intangible technology transfer - Each office reflects the necessities of export controls on the relevant regulation and procedures and make DB for the proper duty. However, there is no indication for export controls on the DB - In the case of foreigner’s temporary visit for simple work and site tour, there is a difficult situation not to be able to check all the visitors by checking the denial lists - If necessary, KAERI may build the TFT (Task force Team) for the efficiency of export controls - Others

Over the past decades, particle physics has made significant progress in characterizing neutrinos even if neutrinos have extremely small cross-section (~10-44 cm2), allowing them to penetrate any object. More recently, neutrino detection and analysis have indeed become valuable tools in various aspects of nuclear science and technology. Neutrinos are detected using various methods, including Inverse Beta Decay (IBD), Neutrino-electron scattering, and Coherent Neutrino-Nucleus Scattering (CNNS). For the detection of anti-neutrinos from nuclear reactor, the Inverse Beta Decay (IBD) is commonly considered with scintillators. Notable experiments in Korea, such as RENO and NEOS, have been conducted using the IBD method at the Hanbit Nuclear Power Plant since 2006. Additionally, the NEON experiment, which employs CNNS, which has a significantly larger reaction cross-section than IBD but its low-energy signal detection difficulty, has been ongoing since 2021. Based on the results of NEOS (2015-2020) the signal to noise is ~30 and IBD detection rate is ~2000 counts per day. The IBD event in nuclear power plants provides valuable information about reactor behavior. IBD count rates are in good agreement with the thermal power of the reactor. Furthermore, the neutrino energy spectrum can be used to estimate the fission isotope ratio of the reactor core, showing promise for obtaining reactor core information from antineutrino detection techniques. Neutrino detection in nuclear facilities provides valuable information about reactor behavior. However, as a surveillance technology neutrino detection faces challenges due to the very low cross-section, requiring efforts to overcome limitations related to detector size and signal acquisition time. In 2008, the International Atomic Energy Agency (IAEA) included neutrino detection in its Research and Development (R&D) program for reactor safeguards. In January 2023, the IAEA organized a “Technical Meeting on Nuclear Data Needs for Antineutrino Spectra Applications” to discuss the latest developments and research results in this field. In summary, the use of neutrino detection in the nuclear field, particularly for reactor monitoring and safeguarding, has advanced significantly. Ongoing research and collaboration are expected to enhance our understanding of neutrinos and their applications in nuclear science and technology.

Regulators conduct inspections and issue non-compliance notice, and it is necessary to examine whether this is equivalent a corrective order. A corrective order imposes binding obligations on a business, and violations of it can result in sanctions such as fines, license revocation, or suspension. Therefore, if it is a corrective order, it must go through procedures such as prior notification under the Administrative Procedure Act. However, so far, no such procedures have been followed when issuing non-compliance notice. There are three possible interpretations of it: 1) the issuance of a non-compliance notice is not a corrective order but a part of an inspection, 2) the issuance of a non-compliance notice is a corrective order but not a disposition, and 3) the process of hearing opinions and prior notification was carried out during the inspection. However, if it is a part of an inspection, it should be issued by KINAC or KINS, which is entrusted with the inspection, but it is issued by the Nuclear Safety and Security Commission, and it is a disposition because it makes specific demands, and the corrective orders themselves have not gone through the procedure of hearing opinions. Therefore, in order for a non-compliance notice to be enforceable unlike a recommendation and to be issued by the Nuclear Safety and Security Commission instead of the inspection agency, the law should be amended and the procedure of prior notice and hearing opinions required by the Administrative Procedure Act should be carried out at the issuance stage.