Surface water temperature of a bay (from the south to the north) increases in spring and summer, but decreases in autumn and winter. Due to shallow water depth, freshwater outflow, and weak current, the water temperature in the central to northern part of the bay is greatly affected by the land coast and air temperature, with large fluctuations. Water temperature variations are large in the north-east coast of the bay, but small in the south-west coast. The difference between water temperature and air temperature is greater in winter and in the south-central part of the bay than that in the north to the eastern coast of the bay where sea dykes are located. As the bay goes from south to north, the range of water temperature fluctuation and the phase show increases. When fresh water is released from the sea dike, the surrounding water temperature decreases and then rises, or rises and then falls. The first mode of empirical orthogonal function (EOF) represents seasonal variation of water temperature. The second mode represents the variability of water temperature gradient in east-west and north-south directions of the bay. In the first mode, the maximum and the minimum are shown in autumn and summer, respectively, consistent with seasonal distribution of surface water temperature variance. In the second mode, phases of the coast of Seosan~Boryeong and the east coast of Anmyeon Island are opposite to each other, bordering the center of the deep bay. Periodic fluctuation of the first mode time coefficient dominates in the one-day and half-day cycle. Its daily fluctuation pattern is similar to air temperature variation. Sea conditions and topographical characteristics excluding air temperature are factors contributing to the variation of the second mode time coefficient.

Some of the metal waste generated from KEPCO NF is being disposed of in the form of ingots. An ingot is a metal that is melted once and then poured into a mold to harden, and it is characterized by a uniform distribution of radioactive material. When measuring the uranium radioactivity in metal ingot with HPGe detector, 185.7 keV of U-235 is used typically because most gamma rays emitted at U-235 are distributed in low-energy regions below 200 keV. To analyze radioactivity concentration of U-235 with HPGe detector more accurately, self-attenuation due to geometrical differences between the calibration source and the sample must be corrected. In this study, the MCNP code was used to simulate the HPGe gamma spectroscopy system, and various processes were performed to prove the correlation with the actual values. First an metal ingottype standard source was manufactured for efficiency calibration, and the GEB coefficient was derived using Origin program. And through the comparison of actual measurements and simulations, the thickness of the detector’s dead layers were defined in all directions of Ge crystal. Additionally instead of making an metal ingot-type standard source every time, we analyzed the measurement tendency between commercially available HPGe calibration source (Marinelli beaker type) and the sample (metal ingot type), and derived the correction factor for geometry differences. Lastly the correction factor was taken into consideration when obtaining the uranium radioactivity concentration in the metal ingot with HPGe gamma spectroscopy. In conclusion, the U-235 radioactivity in metal ingot was underestimated about 25% of content due to the self-attenuation. Therefore it is reasonable to reflect this correction factor in the calculation of U-235 radioactivity concentration.

The engineered barrier system (EBS) for deep geological disposal of high-level radioactive waste requires a buffer material that can prevent groundwater infiltration, protect the canister, dissipate decay heat effectively, and delay the transport of radioactive materials. To meet those stringent performance criteria, the buffer material is prepared as a compacted block with high-density using various press methods. However, crack and degradation induced by stress relaxation and moisture changes in the compacted bentonite blocks, which are manufactured according to the geometry of the disposal hole, can critically affect the performance of the buffer. Therefore, it is imperative to develop an adequate method for quality assessment of the compacted buffer block. Recently, several non-destructive testing methods, including elastic wave measurement technology, have been attempted to evaluate the quality and aging of various construction materials. In this study, we have evaluated the compressive wave velocity of compacted bentonite blocks via the ultrasonic velocity method (UVM) and free-free resonant column method (FFRC), and analyzed the relationship among compressive wave velocity, dry density, thermal conductivity, and strength parameter. We prepared compacted bentonite block specimens using the cold isostatic pressure (CIP) method under different water content and CIP pressure conditions. Based on multiple regression analysis, we suggest a prediction model for dry density in terms of manufacturing conditions. Additionally, we propose an empirical model to predict thermal conductivity and unconfined compressive strength based on compressive wave velocity. The database and suggested models in this study can contribute to the development of quality assessment and prediction techniques for compacted buffer blocks used in the construction of a disposal repository.

Since high-level radioactive wastes contain long-lived nuclides and emit high energy, they should be disposed of permanently through a deep geological disposal system. In Korea, the first (2016.07) and the second (2021.12) basic plans for the management of high-level disposal systems were proposed to select sites for deep geological disposal facilities and to implement business strategies. Leading countries such as Finland, Sweden and France have developed and applied safety cases to verify the safety of deep geological disposal systems. By examining the regulatory status of foreign leading countries, we analyze the safety cases ranging from the site selection stage of the deep geological disposal system to the securing of the permanent disposal system to the investigation, analysis, evaluation, design, construction, operation, and closure. Based on this analysis, we will develop safety case elements for long-term safety of deep geological disposal systems suitable for domestic situation. To systemically analyze data based on safety cases, we have established a database of deep geological disposal system regulations in leading foreign countries. Artificial intelligence text mining and data visualization techniques are used to provide database in dashboard form rather than simple lists of data items, which is a limitation of existing methods. This allows regulatory developers to understand information more quickly and intuitively and provide a convenient interface so that anyone can easily access the analyzed data and create meaningful information. Furthermore, based on the accumulated bigdata, the artificial intelligence learns and analyzes the information in the database through deep learning, and aims to derive a more accurate safety case. Based on these technologies, this study analyzed the legal systems, regulatory standards, and cases of major international leading countries and international organizations such as the United States, Sweden, Finland, Canada, Switzerland, and the IAEA to establish a database management system. To establish a safety regulation base suitable for the domestic deep geological disposal environment, the database is provided as data to refer to and apply systematic information management on regulatory standards and regulatory cases of overseas leading countries, and it is expected that it will play a key role as a forum for understanding and discussing the level of safety of deep geological disposal system among stakeholders.

The engineered barrier system (EBS) is an indispensable element of a deep geological repository (DGR) designed to prevent the discharge of radioactive materials into the environment. The buffer material is a vital component of the EBS by creating a physical and chemical barrier that prevents the migration of radioactive materials. In the disposal environment, gases can be generated from the corrosion of the canister. When the gas generation rate exceeds the diffusion rate, the buffer material’s performance can deteriorate by the physical damage induced by the increase in pore pressure. Therefore, understanding the EBS’s behavior under gas generation conditions is crucial to guarantee the longterm safety and performance of the DGR. Lab-scale and field-scale experiments have been conducted to examine the stability of the buffer material concerning gas generation and movement by the previous researchers. To evaluate long-term stability for more than 100,000 years, it is essential to assess stability using a numerical model verified by these experiments. This study investigated the effect of interfacial characteristics on the numerical modeling accuracy of experimental simulation while verifying a numerical model through field-scale experimental results. The findings of this study are expected to furnish fundamental data for establishing numerical analysis guidelines for the longterm stability assessment of disposal systems.

Since spent nuclear fuel (SNF) should be isolated from the human life zone for at least 106 years, deep geological disposal (DGD) is considered a strong candidate for SNF management in many countries. Therefore, a disposal canister should be nearly immune to corrosion in such a long-term storage environment. Even though copper has a low corrosion rate of a few millimeters per million years in geological environments, the corrosion resistance of the copper welds must be preferentially validated, which inevitably occurs during the sealing of the disposal canister after the SNF is loaded. This is because the weld zone is a discontinuous area of microstructure, which can accelerate uniform and localized corrosion. In this study, the microstructural characteristics of copper welds in different welding conditions such as friction stir welding, electron beam welding, cold spray, were analyzed, focusing on the formation of microstructure, which affects resistance to corrosion. In addition, the microstructure and corrosion properties of the copper weld zone manufactured by recent wire-based additive manufacturing (AM) technology were experimentally evaluated. From this preliminary test result, it was found that the corrosion characteristics of the welds produced by the AM process using wire are comparable to those of the conventional forged copper plate.

Geologic disposal at deep depth is an acceptable way to dispose of high-level radioactive waste and isolate it from the biosphere. The geological repository system comprises an engineered barrier system (EBS) and the host rock. The system aims to delay radionuclide migration through groundwater flow, and also, the flow affects the saturation of the bentonite in the EBS. The thermal conductivity of bentonite is a function of saturation, so the temperature in the EBS is directly related to the flow system. High-temperature results in the two-phase flow, and the two-phase flow system also affects the flow system. Therefore, comprehending the influencing parameters on the flow system is critical to ensure the safety of the disposal system. Various studies have been performed to figure out the complex two-phase flow characteristics, and numerical simulation is considered an effective way to predict the coupled behavior. DECOVALEX (DEvelopment of COupled models and their VALidation against EXperiments) is one of the most famous international cooperating projects to develop numerical methods for thermo-hydro-mechanicalchemical interaction, and Task C in the DECOVALEX-2023 has the purpose of simulating the Fullscale Emplacement (FE) experiment at the Mont-Terri underground research laboratory. We used OGS-FLAC, a self-developed numerical simulator combining OpenGeoSys and FLAC3D, for the simulation and targeted to analyze the effecting parameters on the two-phase flow system. We focused on the parameters of bentonite, a key component of the disposal system, and analyzed the effect of compressibility and air entry pressure on the flow system. Compressibility is a parameter included in the storage term, defining the fluid storage capacity of the medium. While air entry pressure is a crucial value of the water retention curve, defining the relation between saturation and capillary pressure. From a series of sensitivity analyses, low compressibility resulted in faster flow due to low storage term, while low air entry pressure slowed flow inflow into the bentonite. Low air entry pressure means the air easily enters the medium; hence the flow rate becomes lower based on the relativity permeability definition. Based on the sensitivity analysis, we further investigate the effect of shotcrete around the tunnel and excavation damaged zone. Also, long-term analysis considering heat decay of the radioactive waste will be considered in future studies.

Backfill is one of the key elements of deep geological disposal. The backfill material is used to fill disposal tunnels and is mainly composed of swellable clay, preventing the migration of nuclide and structurally supporting the tunnel. The selection and application of backfill material are critical for the stable and efficient disposal of spent fuel. Therefore, it is essential to secure various candidate materials for backfill and to comprehensively understand the properties and behavior of these materials. Recently, the Korea Atomic Energy Research Institute has selected a candidate material called Bentonil-WRK and is evaluating its applicability. To utilize this material as backfill, the safety function of a mixed backfill concept, consisting of sand and Bentonil-WRK, was assessed. The swelling pressure was measured as a function of dry density for a bentonite/silica sand mix ratio of 3/7. The results showed that the swelling pressure ranged from 0.15 to 0.273 MPa, depending on the dry density, with higher dry densities resulting in higher swelling pressures. The measured swelling pressure met the target performance criteria suggested by SKB and Posiva (i. e., 0.1 MPa), but did not meet the design requirement for swelling pressure (i. e., 1 MPa). This indicate the need for further research after increasing the mass fraction of bentonite (e. g., mix ratio 4/6 or more). The results of this study are expected to be used in the selection of candidate backfill materials and the establishment of design guidelines for engineered barrier backfill.

A methodology is under development to reconstruct and predict the long-term evolution of the natural barrier comprising the site of radioactive waste disposal. The natural barrier must protect the human zone from radionuclides for a long time. So for this, we need to be able to restore the evolution of the bedrock constituting the natural barrier from the past to the present and to predict from the present to the future. A methodology is being studied using surface outcrop, tunnel face of KURT (KAERI Underground Research Tunnel), and drill core at KAERI (Korea Atomic Energy Research Institute). Among them, drill core is an essential material for identifying deep geological properties, which could not be confirmed near the surface when considering the geological condition of the repository in the deep part. In this study, we selected several qualitative and quantitative analyses to construct a deep lithological model from the disposal perspective. These were applied to drill core samples around the KURT. There are the dikes presumed the Cretaceous were intruded by Jurassic granitoids in the study area. Analyzing trace elements of each rock type in the study area classified through geochemical characteristics and microstructure in previous studies made it possible to obtain qualitative information on the petrogenetic process. In addition, synthesizing the quantitative numerical age allows for grasping the evolution of bedrock, including intrusion and cutting relationships. LAICPMS was used for determining the age of zircons in plutonic rocks. The highly reliable 40Ar-39Ar method was selected for volcanic rocks because it can correct the loss of Ar gas and obtain the values of two types of Ar isotopes in a single sample. As a result, it was possible to infer the formation environment of rocks through anomalies in specific trace element content. And according to the numerical ages, it was possible to support the known separated rock type found in previous studies or to present a quantitative precedence relation for unclassified rocks. These methods could be applied to reconstruct the long-term evolution of bedrock within natural barriers.

The purpose of this study was to examine whether galvanic corrosion of copper occurs by inserting a third barrier layer with a higher corrosion potential than copper between copper and cast iron when the copper layer is locally perforated by pitting or partial corrosion. A triple layer composed of copper, inserted metal, and carbon steel was manufactured by cold spray coating of inserting metal powders such as Ag, Ni, and Ti on carbon steel plate followed by Cu coating on it. First, the corrosion properties were evaluated electrochemically for each metal coating. As a result of Tafel plot anaylsis in KURT groundwater condition, the corrosion potential of Fe (-567 mV) was much lower than that of Cu (-91 mV), and the corrosion potential of Ni (-150 mV) was also lower than that of Cu. Therefore, Ni was likely to corrode before Cu. However, the corrosion current of Ni was lower than that of the Cu. In the galvanic specimen where the copper and inserting metal were exposed together, Cu-Fe was much lower corrosion potential of -446 mV, and the corrosion potential of Cu-Ti, Cu-Ni, and Cu-Ag were slightly higher than that of Cu. Therefore, it seemed that Ag, Ni, and Ti all might promote galvanic corrosion of surrounding copper when the copper layer was perforated to the inserted metal layer. If the metal insertion presented in this study operates properly, the disposal container does not need to worry about the partial corrosion or non-uniform corrosion of external copper layer.

In Korea, borated stainless steel (BSS) is used as a storage rack in spent fuel pools (SFP) to maintain the nuclear criticality of spent fuels. As the number of nuclear power plants and the corresponding amount of spent fuels increased, the density in SFP storage rack also increased. In this regard, maintaining subcriticality of spent nuclear fuels became an issue and BSS was selected as the structural material and neutron absorber for high density storage rack. Since it is difficult to replace the storage rack, corrosion resistance and neutron absorbency are required for long period. BSS is based on stainless steel 304 and is specified in the ASTM A887-89 standard depending on the boron concentration from 304B (0.20-0.29% B) to 304B7 (1.75-2.25% B). Due to the low solubility of boron in austenitic stainless steel, metallic borides such as (Fe, Cr) 2B are formed as a secondary phase. Metallic borides could cause Cr depletion near it, which could decrease the corrosion resistance of the material. In this paper, the long-term corrosion behavior of BSS and its oxide microstructures are investigated through accelerated corrosion experiment in simulated SFP conditions. Because the corrosion rate of austenitic stainless steel is known to be dependent on the Arrhenius equation, a function of temperature, the corrosion experiment is conducted by increasing the experimental temperature. Detail microstructural analysis is conducted using a scanning electron microscope, transmission electron microscope and energy dispersive spectrometer. After oxidation, a hematite structure oxide film is formed, and pitting corrosion occurs on the surface of specimens. Most of the pitting corrosion is found at the substrate surface because the corrosion resistance of the substrate, which has low Cr content, is relatively low. Also, the oxidation reaction of B in the secondary phase has the lowest Gibbs free energy compared to other elements. Furthermore, oxidation of Cr has low Gibbs free energy, which means that oxidation of B and Cr could be faster than other elements. Thus, the long-term corrosion might affect the boron content and the neutron absorption ability of the material. Using boron’s high cross-section for neutrons, the neutron absorption performance of BSS was evaluated through neutron transmission tests. The effect of the corrosion behavior of BSS on its neutron absorption performance was investigated. Samples simulated to undergo up to 60 years of degradation before corrosion through accelerated corrosion testing did not show significant changes in the neutron shielding ability before and after corrosion. This can be explained in relation to the corrosion behavior of BSS. Boron was only leached out from the secondary phase exposed on the surface, and this oxidized secondary phase corresponds to about 0.17% of the volume of the total secondary phase. This can be seen as a very small proportion compared to the total boron content and is not expected to have a significant impact on neutron absorption performance.

Korean MMTT project has been launched in order to clarify the vibration and shock loads under normal condition and transportation (NCT) in Korean geological and transportation conditions and to evaluate the integrity of SNF under such a transportation load. To evaluate the integrity of the SNF during normal land and sea transport tests, a representative SNF that represents the entirety of the different types of SNFs stored in the spent fuel pool of the power plant should be selected. And, it is necessary to make the test assembly to have a statically and dynamically similar behavior with the actual SNF. Therefore, in this project, we selected two types of fuel assembly that are expected to exhibit relatively conservative behavior under NCT, and these assemblies are being fabricated into surrogate test assemblies to have a similar characteristic as actual SNF based on the accumulated data from the poolside examination and the hot cell test so far. Tests were conducted for NCT conditions. In addition, a fatigue test was performed to integrity of the nuclear fuel rods under NCT conditions. Nuclear fuel assemblies are transported while being laid inside the cask under NCT, and are exposed to external shocks and vibrations. At this time, the fuel rod between the grid and grid is exposed to bending motion by this external force. For this simulation, a fixture was developed and used for static bending tests and bending fatigue tests. To simulate spent nuclear fuel rod specimens, hydrogen reorientation Zry-4 cladding was used and simulated pellets made of stainless steel were applied. And also, it was bonded using epoxy to give bonding conditions between the inside and the pellet. As a result of the test, cracks occurred due to the concentrated load between the pellets, resulting in damage to the fuel rod. The fatigue results showed a similar trend compared to the results performed by ORNL, and the lower bound fatigue curve presented by NUREG-2224 was also satisfactory.

Spent Fuel Pool Island (SFPI) is a spent nuclear fuel storage pool that operates independently of existing nuclear facilities to safely manage SNF and minimize maintenance costs during the nuclear decommissioning process. Since the radiation controlled area can be dismantled before transporting SNF to a dry storage facility, the overall decommissioning period can be shortened, and the risk of occupational exposure during dismantling is reduced. In the US, various nuclear power plants have introduced SFPI for this reason. In this paper, to analyze the economic feasibility of application of SFPI to nuclear power plants to be decommissioned, several scenarios are established in consideration of the decommissioning plan and schedule, SFPI and dry storage facility application schedule. Cost and benefit list (SFPI application cost, SNF management cost, SNF dry storage cask cost, etc.) for each alternative were derived, and economic analysis was conducted by applying the Net Present Value (NPV). As a result of the analysis, it is found that the application of SFPI during decommissioning is economically effective as the NPV showed a positive number even when uncertainty was taken into account.

The current storage capacity of the spent nuclear fuel at the Kori unit 2 has reached approximately 94% saturation, excluding emergency core capacity. As South Korea has not yet constructed any interim storage facilities to store spent nuclear fuel, one of possible options is to install high density storage racks in spent fuel pool at the reactor site to increase its capacity. The high density storage rack is a technology developed to allow the storage to have more spent nuclear fuel than originally planned, while still ensuring safety. It achieves this by reducing the storage gap between the spent nuclear fuel. The purpose of this study is to investigate an appropriate storage capacity for spent fuel pool in the Kori unit 2 and the factors to be considered during the high density storage rack installation. Given that the Kori unit 2 is planning continued operation and Korea Hydro & Nuclear Power (KHNP) is preparing to install a temporary storage facility for spent nuclear fuel at the Kori nuclear site, it is reasonable to consider the installation of high density storage racks in the spent fuel pool as a storage solution for these issues. When evaluating the capacity of the spent fuel pool, the amount of spent nuclear fuel generated by other reactors in Kori nuclear site and the amount of spent nuclear fuel generated by continued operation of the Kori unit 2 should be taken into account. This study aims to consider these factors and evaluate the capacity of the spent fuel pool. Furthermore, when installing high density storage rack for the spent nuclear fuel, it should be noted that the existing storage racks at the Kori unit 2 are welded to the liner plate, which may require additional cutting work. Therefore, it is necessary to review the suitable method for the cutting work. Additionally, assuming that divers need to access the area near the storage racks or cutting & welding devices require radiation protection in the area, it is essential to analyze the expected radiation level with computational code and propose appropriate measures to limit work time or establish a work zone. Thus, this study evaluates appropriate capacity of spent fuel pool and work methods for the installation of high density storage rack in the spent fuel pool at the Kori unit 2. It is expected that this paper contributes to install high density storage racks in SFP safely.

Flow-induced vibration can lead to fretting wear damage of fuel rods and spacer grids in nuclear reactors. Similarly, during the transport of spent nuclear fuel assemblies, continuous vibration and intermittent impact might also result in fretting wear due to dynamic interaction. Therefore, it is important to evaluate fuel rod damage due to fretting wear under such transport conditions. This study examines spent nuclear fuel rod specimens fabricated with hydride cladding tubes and simulated pellets, with hydrogen content ranging from 200 to 700 ppm and oxide film thickness ranging from 25 to 100 micrometers. Tests were conducted under a worst-case scenario, assuming continuous exposure to that condition during the expected transport time. Results indicate that the wear depth of all rod specimens occurred within the oxide film, suggesting a high resistance to fretting wear during transportation.

There have been a variety of issues related to spent nuclear fuel in Korea recently. Most of the issues are related to intermediate storage and disposal of spent nuclear fuel. However, recently, various studies have been started in advanced nuclear countries such as the United States to reduce spent nuclear fuel, focusing on measures to reduce spent nuclear fuel. In this study, a simple preliminary assessment of the thermal part was performed for the consolidation storage method which separates fuel rods from spent nuclear fuel and stores them. The preliminary thermal evaluation was analyzed separately for storing the spent fuel in fuel assembly state and separating the fuel rods and storing them. The consolidation storage method in separating the fuel rods was advantageous in terms of thermal conductivity. However, detailed evaluation should be performed considering heat transfer by convection and vessel shape when storing multiple fuel bundles simultaneously.

The DPRK had been upgrading its nuclear weapons capabilities from the past to the sixth nuclear test in 2017, and Kim Jong-un has been in power since the death of Kim Jong-il in 2011, striving to upgrade and diversify four nuclear tests and firing means. In 2022, in particular, DPRK launched more than 40 ballistic and cruise missiles and provoked them in various ways, such as developing solid rocket engines, flying fighter jets, and invading drones. In addition, reprocessing facility activities have been observed again since 2021 at the Yongbyon Nuclear Research Complex. Operational activities such as continuous activities of the 5MWe Yongbyon reactor and the additional construction of new buildings are observed. DPRK’s recent activities could result in nuclear weapons in all except conventional weapons provocations. DPRK has researched and developed nuclear weapons since the 1950s. It has been preparing to operate nuclear weapons, operating nuclear power, and modernizing nuclear power simultaneously. Given the number of nuclear weapons using DPRK’s nuclear materials and various means of missiles, an offensive transition is expected to enable restrictive deterrence strategies that can be used first use and on assured retaliation. In addition, based on the nuclear strategy, which is interpreted as Vipin Narang’s nuclear doctrine and nuclear posture, DPRK is also capable of assured retaliation and asymmetric escalation posture. In particular, considering the continuous activities of the Yongbyon Nuclear Research Complex, which has recently diversified the investment vehicle, and the delegation of nuclear weapon use, it is expected to move differently from the previous one based on the changed nuclear strategy. However, there are clear limitations to interpreting it as a completely assured retaliation and asymmetric expansion nuclear strategy. First, there is a lack of development of atmospheric reentry vehicles that can avoid ICBM interception for assured retaliation capabilities against the United States. Second, there are limitations in the open capacity of nuclear weapons due to the absence of SSBN capabilities. However, delegation to operations at strategy force suggests the possibility of asymmetric expansion strategies. The previous analysis of DPRK’s nuclear strategy and limitations is valid in that the U.S. nuclear umbrella guarantees the Republic of Korea in a strong alliance between the Republic of Korea and the U.S. If the Republic of Korea lacks a nuclear umbrella due to the weakening of the alliance or limits U.S. intervention by having more than dozens of ICBMs, it is considered that DPRK can use a definite confirmation retaliation and asymmetric escalation nuclear strategy. As a response to this, it is the first way to verify and strengthen the validity of the three-axis system (Kill Chain, KAMD, KMPR), second to strengthen the Korea-U.S. alliance, and finally to appeal to the international community and increase consensus. In particular, it is possible to form a consensus of sanctions and condemnation DPRK by expressing concerns about nuclear dominoes caused by nuclear proliferation and arms competition to the international community.