The decommissioning of nuclear power plants will generate a lot of low and intermediate-level radioactive waste (LILW), and preliminary radioactive evaluation for these wastes should be carried out before decommissioning work. Mainly, Concrete, Carbon Steel, Stainless Steel-304 (SS304) and Inconel are used in many parts of nuclear power plants and considered as main resource of nuclear wastes. Depending on the material location, the number of neutrons irradiated to material varies, which can range from self-disposal waste to LILW. In this paper, activation analysis was performed to compare the radiation dose according to the presence or absence of impurity elements present in SS304. For the calculation, SS304 composition and impurity elements were used as described in the report of NUREG-3474. This report lists 41 impurity elements for SS304 and other materials. Calculation code is used ORIGEN-S module in SCALE 6.1 code. Neutron flux is used as arbitrary value that around 1E+11 level and irradiation time is set as 30 year with 10-year cooling time. In the ORIGEN-S calculation, 1g of SS304 is used for easy calculation of specific activity. The ORIGEN-S calculation results are as follows. All impurity elements contained case calculated 9.32E+07 Bq activity. In the absence of all impurity elements case and most cases shows that total Becquerel value after 10-year cooling time around 9.11E+07 Bq, and Co impurity case had larger result. The calculation was performed again by increasing the amount of impurity substances by 100 times to perform the sensitivity evaluation more reliably. Representatively, Li, N, Co, and Ba impurity elements cases were calculated to have a particularly high Becquerel. Especially Co impurity element case, a total Becquerel of 3.03E+08 was calculated. Accordingly, evaluation of impurities mixed in SS304 must be considered, and in particular, the inclusion rate for Co must be considered.

Advanced countries in the field of nuclear research and technology are currently examining the feasibility of deep geological disposal as the most appropriate method for the permanent management of high-level radioactive waste, with no intention of future retrieval. Deep geological disposal involves the placement of such waste deep underground within a stable geological formation, ensuring its permanent isolation from the human environment. To guarantee the enduring isolation and retardation of radionuclides with half-lives spanning tens of thousands to millions of years from the broader ecosystem, it is imperative to comprehend the long-term evolution of deep disposal systems, especially the role of natural barriers. These natural barriers, typically consisting of bedrock, encase the repository and undergo long-term evolutions due to tectonic movements and climate variations. For the effective disposal of high-level radioactive waste, a thorough assessment of the site’s long-term geological stability is essential. This necessitates a comprehensive understanding of its tectonic evolution and development characteristics, including susceptibility to seismic and magmatic events like earthquakes and intrusions. Furthermore, a detailed analysis of alterations in the hydrogeological and geochemical environment resulting from tectonic movements over extended time frames is required to assess the potential for the migration of radionuclides. In this paper, we have examined international evaluation methodologies employed to elucidate the predictive long-term evolution of natural barriers within disposal systems. We have extracted relevant methods from international case studies and applied a preliminary scenario illustrating the long-term evolution of the geological environment at the KURT (KAERI Underground Research Tunnel) site. Nevertheless, unlike international instances, the scarcity of quantitative data limits the depth of our interpretation. To present a dependable scenario in the future, it is imperative to develop predictive technologies aimed at comprehensively studying the geological evolution processes in the Korean peninsula, particularly within the context of radioactive waste disposal.

For the performance and safety assessments of deep geological disposal, developing scenarios, which represent possible long-term changes in the surface environment, is required. These scenarios are formulated using a list of FEPs (Features, Events, and Processes) that describes characteristics of disposal system components. In this study, using international FEP (IFEP) list from OECD/NEA, the individual FEPs related to uplift-subsidence and erosion-deposition were analyzed, and the correlation between each FEP was evaluated. From the IFEP list, the elements related to uplift-subsidence and erosion-deposition processes that cause long-term changes in the surface environment were identified. Uplift-subsidence, erosion - deposition, and the long-term change factors caused by them were analyzed and a correlation diagram was produced according to their interactions. Basis for the integrated analysis of long-term changes in the surface environment and the construction of long-term change scenarios were established considering the evaluation of the factors that cause uplift-subsidence and erosiondeposition, and their correlation with the hydrology-hydrogeology, topography and local climate of the affected surface. The results of this study will be used for systematically formulating scenarios of long-term changes in the surface environment due to uplift-subsidence and erosion-deposition based on natural phenomena. And, it may be necessary to modify and supplement the correlation of domestic FEPs based on the correlation diagram of IFEPs in order to analyze long-term changes in the surface environment in an integrated manner.

In the evaluation of the stability of radioactive waste disposal, it is imperative to take into account the concept of the redox front. Initially, this front is typically observed near the surface. However, if the hydraulic gradient increases due to the construction of a disposal facility, the redox front can potentially transport deeper into the geological environment through groundwater flow. This transport triggers changes in the geochemical characteristics, potentially diminishing the natural buffering capacity of the bedrock. Consequently, it is necessary to characterize both the unsaturated and saturated zones in the disposal site. In this context, a tracer test is a useful method to identify the characteristics of the site from the surface to the deep geological environment where the disposal facility can be located. Therefore, this study also aims to establish a methodology enabling a comprehensive understanding of the hydrogeochemical characteristics through the tracer test that can be applied to future sites for research URL (Underground Research Laboratory) or radioactive waste disposal in Korea. For the tracer test, a UNIT (UNsaturated zone Insitu Test facility) was built within the KAERI and five wells with a depth of 24 m were installed in 2022. Before conducting the test, to determine the geochemical background characteristics of the site, topsoil and soils at depths of 30 cm, 60 cm, and 90 cm were collected. Additionally, a groundwater sample was obtained from the newly installed well. Soil samples were analyzed for soil texture, moisture content, total and exchange cations, anions, and heavy metals. Similarly, the groundwater sample was analyzed for cations, anions, and trace elements. The outcomes of these comprehensive analyses will serve as the baseline values in the hydrogeochemical changes after the tracer test. This includes changes in soil composition, water quality, precipitation/dissolution processes, and mineral phases. Furthermore, these results will be provided as input parameters for surface-underground interface models in future studies.

The seven-year research project entitled “Development of workflow for integrated 3D geological site descriptive modeling” is being carried out from 2023. This research is funded by Ministry of Trade, Industry, and Energy (MOTIE). Progress of the research is discussed here. The integrated 3D geological SDM (site descriptive model; GSDM hereafter) consists of three part; 1) three dimensional representation of geologic elements, 2) database for material properties and modeling results from SDMs of other disciplines (e.g., rock mechanics), and 3) a visualization tool for geology, material properties and modeling results. The GSDM is comparable to the GDSMs of SKB and POSIVA in its representation of geology by volume of geologic elements. However, our GSDM is different in that extra information of material properties and an extra tool for visualization is included in the GDSM. The rationale for incorporating material properties and a visualization tool into the GSDM is to expedite the development of the GSDM and SDMs of other disciplines by allowing single institution to integrate database and visualization with the GSDM. SKUA-GOCAD is used for representation of geologic surfaces for ductile and brittle shear zones, and also for surfaces for delineation of volumes of rock units. We have adopted SKUAGOCAD because the program offers powerful functions of interpolation including borehole data and geophysical prospecting. So far, we have tested the program for five different geologies, including sedimentary, high-grade metamorphic, and intrusive igneous geology. The test results are promising. Incorporation of data and modeling results for the SDMs of other disciplines is at conceptual stage. The working conceptual model involves the following steps, 1) to provide the modeler of other disciplines with surface information representing geologic elements, 2) the modeler returns not only material properties but the results of numerical analysis, and 3) incorporation of material properties and modeling results into database. Since the numerical codes in other disciplines adopt different types of formats for 3D geology, we plan to adopt the widely used FEM format prepared by Gmsh. The visualization tool will also adopt Gmsh for graphical representation of 3D geology as well as database for material properties and modeling results. When the working model of GSDM becomes available, rapid and significant progress is expected in the SDMs of other disciplines and related areas, for example, geotechnical investigation for deep geological repository.

The effectiveness of a crystalline natural barrier in providing sealing capabilities is based on the behavior of numerous fractures and their intersections within the rock mass. It is important to evaluate the evolving characteristics of fractured rock, as the hydro-mechanical coupled processes occurring through these fractures play a dominant role. KAERI is actively developing a true tri-axial compression test system and concurrently conducting hydro-mechanical experiments using replicated fractured rock samples. This research is focused on a comprehensive examination of coupled processes within fractures, with a particular emphasis on the development of true tri-axial testing equipment. The designed test system has the capability to account for three-dimensional stress conditions, including vertical and both maximum and minimum horizontal principal stresses, realizing the disposal conditions at specific underground depths. Notably, the KAERI-designed test system employs the mixed true tri-axial concept, also known as the Mogi-type, which allows for fluid flow into fractures under tri-axial compression conditions. This system utilizes a hydraulic chamber to maintain constant stress in one direction through the application of oil pressure, while the other two directional stresses are applied using rigid platens with varying magnitudes. Once these mechanical stress conditions are established, control over fluid flow is achieved through the rigid platens in contact with the specimen section. This pioneering approach effectively replicates in-situ mechanical conditions while concurrently observing the internal fluid flow patterns within fractures, thereby enhancing our capacity to study these coupled phenomena. As future research, numerical modeling efforts will be proceeding with experimental data-driven approaches to simulate the coupled behavior within the fractures. In these numerical studies, two distinct fracture geometry domains will be generated, one employing simplified rough-walled fractures and the other utilizing mismatched rough-walled fractures. These investigations mark the preliminary steps in the process of selecting and validating an appropriate numerical model for understanding the hydro-mechanical evolution within fractures.

It is crucial to understand the hydro-mechanical behavior of rock mass to assess the performance of natural barriers. As rock fractures serve as both mechanically weak planes and prominent pathways for hydraulic flow, they significantly influence the hydro-mechanical behavior of the rock mass. Hence, understanding the characteristics of rock fractures is necessary to analyze the long-term behavior of natural barriers. In particular, fracture apertures are crucial parameters directly associated with groundwater flow and consequently hold significant importance in determining the hydro-mechanical behavior of natural barriers. Fracture apertures are defined as mechanical and hydraulic apertures, and various studies have been conducted to measure and analyze them. However, direct measurement of mechanical aperture according to changes in normal stress is known to be a challenging task. For this reason, there has been a scarcity of direct comparative findings between mechanical and hydraulic apertures under various normal stress conditions. This study aims to analyze the characteristics of the mechanical and hydraulic apertures according to changes in normal stress based on experimental results. A digital analysis technique using a pressure film image was applied to analyze the mechanical aperture characteristics of the fracture. This technique can be applied by performing a pressure film compression test and a normal stiffness test on a fracture specimen, and has the advantage of being able to derive mechanical apertures under various normal stress conditions. The hydraulic aperture characteristics of the fracture were analyzed based on Cubic law after measuring the flow rate by performing a constant pressure injection test under triaxial compression conditions. By applying various confining pressures, it was possible to examine the hydraulic apertures according to changes in normal stress conditions. Through the experimental results, the relationship between the mechanical and hydraulic apertures of the fracture was summarized under various normal stress conditions. In addition, the experimental results were used to examine the applicability of various empirical equations for mechanical and hydraulic apertures proposed in previous studies. The characteristics of the fracture aperture resulting from this study are significant because they are required in the hydro-mechanical model of natural barriers. Future studies will entail further experiments, with the objective of establishing novel relationships based on the accumulation of experimental data.

Currently, Korea is considering a disposal system based on Sweden’s KBS-3 model to dispose of high-level waste. The disposal system uses a multi-barrier concept to protect high-level waste with canister, buffer, backfill, and natural rock. In Korea, copper and iron are being considered for external and internal canisters, and bentonite is being considered as a buffer material. This is a similar choice to many overseas disposal systems. However, unlike the rolling, extrusion, and forging manufacturing methods being considered overseas for manufacturing external canister, domestic research is currently underway on manufacturing external copper canister using cold spray coating. The canister manufacturing method may vary depending on unit cost and manufacturing convenience. However, the properties of metal vary slightly depending on the manufacturing method of the metal. In this case, the characteristics of the canister may vary slightly depending on the canister manufacturing method, and eventually the corrosion resistance may also vary slightly. In order to understand how the copper canister manufacturing method affects corrosion resistance, corrosion rates were calculated and compared through electrochemical corrosion experiments at domestic groundwater ion concentration.

Long-term climate and surface environment changes can influence the geological subsurface environment evolution. In this context, a fluid flow pathway developing and connection possibility can be increased between the near-surface zone and deep depth underground. Thus, it is necessary to identify and prepare for the overall fluid flow at the entire geological system to minimize uncertainty on the spent nuclear fuel (SNF) disposal safety. The fluid flow outside the subsurface environment is initially penetrated through the surface and then the unsaturated area. Thus, the previously proved reports, POSIVA in Finland, suggested that sequential research about the fluid infiltration experiment (INEX) and the investigation is necessary. Characterizing the unsaturated zone can help predict changes and ensure the safety of SNFs according to geological long-term evolution. For example, the INEX test was conducted at the upper part of ONKALO, about 50 to 100 m depth, to understand the geochemical evolution of the groundwater through the unsaturated zone, to evaluate the main flow of groundwater that can approach the SNF disposal reservoir, and to estimate the decreasing progress of the buffering capacity along the pathway through the deep geological disposal. In the present study, a preliminary test was performed in the UNsaturated-zone In-situ Test (UNIT) facility near the KAERI underground research tunnel to design and establish a methodology for infiltration experiments consistent with the regional characteristics. The results represented the methodological application is possible for characterizing unsaturated-zone to perform infiltration experiments. The scale of the experiment will be expanded sequentially, and continuous research will be conducted for the next application.

The design and fabrication of suitable waste forms with high thermal and structural stability are essential for the safe management and disposal of radioactive wastes. In particular, the thermal properties and temperature distribution of waste form containing high heat-generating nuclides such as Cs and Sr can be used to evaluate its thermal stability, but also provide useful information for the design of canisters, storage systems, and repositories. In this study, a new program code-based thermal analysis framework has been developed to facilitate the characterization, design, and optimization of the waste form. Matlab was used as a software development platform because it provides powerful mathematical computation and visualization components such as the partial differential equation (PDE) toolbox for solving heat transfer problems using finite element method, the App Designer for graphical user interface (GUI), and the MATLAB Compiler for sharing MATLAB programs as standalone applications and web applications. The thermal analysis results such as temperature distribution, heat flux, maximum/ minimum temperature, and centerline/surface temperature profile are visualized with graphs and tables. To evaluate the effectiveness of the developed program, several design and optimization studies were carried out for the SrTiO3 waste form, selected as a stable form of strontium nuclide.

The nuclide management process for reducing the environmental burden being developed by the Korea Atomic Energy Research Institute is performed in molten salts, resulting in contaminated salt wastes containing fission products such as Cs, Sr, Ba, and rare-earth nuclides. In addition, the spent fuel of a molten salt reactor (MSR) contains a variety of fission products, and a purification process may be required for the reuse of the salt and the separation and disposal of the fission products in the spent nuclear fuel. The melt-crystallization method is a technique used for the purification and separation of chemicals or metals based on the different melting points of the different substances. In a recent study, our group developed a reactive-crystallization method using Li2CO3 precipitation agent to precipitate metal corrosion from the reactor through a chlorination reaction by HCl and Cl2, which may occur in chloride molten salt, and successfully precipitated the metal precipitate and purified and recovered LiCl salt. In this study, reactive-crystallization method has been established for removing fission products and corrosive materials. Using the reactive crystallization method, white LiCl-KCl salt that was not discolored by metal corrosion was recovered through the crystallization plates, and fission products and metal elements were shown to be suppressed to several ppm in the purified salt. Consequently, high-purity salts were recovered with high nuclide and corrosive separation efficiencies. The reactive crystallization procedure can also be applied to other salt waste systems, such as MSR nuclear fuel treatment and molten salt chemistry for the elimination of corrosive substances.

Various types of spent fuel assembly in nuclear power plants have been transported to a post irradiation examination facility (PIEF) in KAERI to examine the mechanical and chemical properties of fuel and cladding. Once the fuel assembly arrive at PIEF, it is dismantled in a pool area to extract the fuel rods. Dismantling of the fuel assembly is performed by cutting the top nozzle. Currently, couple of dismantled assemblies have been stored in a storage pool without the top nozzle in PIEF. These assemblies cannot be handled directly using a gantry crane in the pool, and thus are contained in a special basket to handle. In this research, we developed a restoration method for a dismantled spent fuel assembly, especially for 16×16 Korea Optimized Fuel Assembly (KOFA). After reviewing the original design document and reports of KOFA, two tools are devised; an assembly tool and a tightening tool for a bolt. Since the top nozzle and dismantled KOFA can be re-assembled using a bolt, we follow the original design, size, and materials of the previously used bolt. The bolt to restore the top nozzle of KOFA is made of 321 stainless steel and has a design that fits the guideline of DIN 13-21 international standard. Our procedure can potentially be used to restore and repair the dismantled spent fuel assembly.

Ring Tensile Test (RTT) is mainly performed for comparing tensile strength and total strain between nuclear fuel cladding specimens under various initial conditions. Through RTT, the loaddisplacement (F-D) curve obtained from the uniaxial tensile test can also be obtained. However, the Young’s modulus estimated from the gradient of the straight portion is much lower than general value of materials. The reasons include tensile machine compliance, slack in the fixtures, or elastic deformation of the fixtures and the tooling. Another reason is that the bending of the test part in the ring is stretched with two pieces of tools. Although the absolute value of the Young’s modulus is smaller than the actual value, it is applicable to calculate the ratio of the Young’s moduli of different materials, that is, the relative value. The Young’s modulus, or slope of the linear section, varies slightly depending on which location data is used and how much data is included. In order to obtain a more accurate ratio of Young’s moduli between materials using the RTT results, a post-processing method for the ring tensile test results that can prevent such human errors is proposed as follows. First, the slope of the linear section is obtained using the displacement and load when the load increase is the largest and the displacement and load of the position that is 95% of the maximum load increase. To replace the section where the ring-shaped specimen is stretched at the beginning of the F-D curve, a straight line equal to the slope of the linear section is drawn to the displacement axis from the position of maximum load increase and moved to the origin to obtain the final F-D curve for a RTT. Lastly, the yield stress uses the stress at the point where the 0.2% offset straight line and the F-D curve meet as suggested in the ASTM E8/E8M-11 “Standard test methods for tensile testing of metallic materials”. RTT results post-processing method was coded using FORTRAN language so that it could be performed automatically. In addition, sensitivity analysis of the included data range on the Young’s modulus was performed by using the included data range as 90%, 85%, and 80% of the maximum load increase.

The solid-state chemistry of uranium is essential to the nuclear fuel cycle. Uranyl nitrate is a key compound that is produced at various stages of the nuclear fuel cycle, both in front-end and backend cycles. It is typically formed by dissolving spent nuclear fuel in nitric acid or through a wet conversion process for the preparation of UF6. Additionally, uranium oxides are a primary consideration in the nuclear fuel cycle because they are the most commonly used nuclear fuel in commercial nuclear reactors. Therefore, it is crucial to understand the oxidation and thermal behavior of uranium oxides and uranyl nitrates. Under the ‘2023 Nuclear Global Researcher Training Program for the Back-end Nuclear Fuel Cycle,’ supported by KONICOF, several experiments were conducted at IMRAM (Institute of Multidisciplinary Research for Advanced Materials) at Tohoku University. First, the recovery ratio of uranium was analyzed during the synthesis of uranyl nitrate by dissolving the actual radioisotope, U3O8, in a nitric acid solution. Second, thermogravimetric-differential thermal analysis (TG-DTA) of uranyl nitrate (UO2(NO3)2) and hyper-stoichiometric uranium dioxide (UO2+X) was performed. The enthalpy change was discussed to confirm the mechanism of thermal decomposition of uranyl nitrate under heating conditions and to determine the chemical hydrate form of uranyl nitrate. In the case of UO2+X, the value of ‘x’ was determined through the calculation of weight change data, and the initial form was verified using the phase diagram for the U-O system. Finally, the formation of a few UO2+X compounds was observed with heat treatment of uranyl nitrate and uranium dioxide at different temperature intervals (450°C-600°C). As a result of these studies, a deeper understanding of the thermal and chemical behavior of uranium compounds was achieved. This knowledge is vital for improving the efficiency and safety of nuclear fuel cycle processes and contributes to advancements in nuclear science and technology.

Since the Fukushima nuclear accident in 2011, the development of accident tolerant fuel (ATF) has been actively pursued as an alternative to improve the safety of nuclear power plants. In addition, nuclear power plants containing ATF have recently been included as green energy in the 2022 EU taxonomy bill, receiving a lot of attention. Many countries are considering increasing 235U enrichment from 5 to 10 235U % for higher burnup and long cycle operation with ATF improving safety. To utilize ATF, the applicability of fuel storage systems such as new fuel storage vault, Region 1, and Region 2 must be determined. The purpose of this paper is to confirm the applicability of applying ATF, which is being developed in Korea, to the nuclear fuel storage system of Korean nuclear power plants. The nuclear power plant model used in the analysis is APR-1400, a representative Korean nuclear power plant model, and ATF model used in the analysis is Mo microcell UO2 pellet with CrAl coating, which is being developed in Korea. MCNP 6.2 has been used for multiplication factor calculations, and the TRITON/NEWT and ORIGEN-S modules of the SCALE code have been used for depletion calculations. From the analysis results, solutions and additional analysis would be necessary to satisfy criticality regulatory requirements to utilize ATF with increased enrichment.

In nuclear fuel development research, consideration of the back-end cycle is essential. In particular, a review of an in-reactor performance of nuclear fuel related to the various degradation phenomena that can occur during spent fuel dry storage is an important area. The important factors affecting the degradation of zirconium-based cladding during dry storage are the cladding’s hydrogen concentration and rod internal pressure after irradiation. In this study, a preliminary analysis of the in-reactor behavior of the HANA cladding, which has been developed and is currently undergoing licensing review, was performed, and based on this result, a comparative analysis between nuclear fuel with HANA cladding and current commercial fuel under storage conditions was performed. The results show that the rod internal pressure of nuclear fuel with HANA cladding is not significantly different from that of commercial cladding, and the hydrogen concentration in the cladding tends to reduce due to the increased corrosion resistance, so fuel integrity in a dry storage conditions is not expected to be a major problem. Although the lack of cladding creep data under dry storage conditions, the results from the Halden research reactor test comparing in-reactor creep behavior with Zircaloy-4 showed that there is sufficient margin for degradation due to creep during storage.

Once discharged, spent nuclear fuel undergoes an initial cooling process within deactivation pools situated at the reactor site. This cooling step is crucial for reducing the fuel’s temperature. Once the heat has sufficiently diminished, two viable options emerge: reprocessing or interim storage. A method known as PUREX, for aqueous nuclear reprocessing, involves a chemical procedure aimed at separating uranium and plutonium from the spent nuclear fuel. This separation not only minimizes waste volume but also facilitates the reuse of the extracted materials as fuel for nuclear reactors. The transformation of uranium oxides through dissolution in nitric acid followed by drying results in uranium taking the form of UO2(NO3)2 + 6H2O, which can then be converted into various solid-state configurations through different heat treatments. This study specifically focuses on investigating the phase transitions of artificially synthesized UO2(NO3)2 + 6H2O subjected to heat treatment at various temperatures (450, 500, 550, 600°C) using X-ray Diffraction (XRD) analysis. Heat treatments were also conducted on UO2 to analyze its phase transformations. Additionally, the study utilized XRD analysis on an unidentified oxidized uranium oxide, UO2+X, and employed lattice parameters and Bragg’s law to ascertain the oxidation state of the unknown sample. To synthesize UO2(NO3)2 + 6H2O, U3O8 powder is first dissolved in a 20% HNO3 solution. The solid UO2(NO3)2 + 6H2O is obtained after drying on a hotplate and is subsequently subjected to heat treatment at temperatures of 450, 500, 550, and 600°C. As the heat treatment temperature increases, the color of the samples transitions from orange to dark green, indicating the formation of different phases at different temperatures. XRD analysis confirms that uranyl nitrate, when heattreated at 500 and 550°C, oxidizes to UO3, while the sample subjected to 600°C heat treatment transforms into U3O8 due to the higher temperature. All samples exhibit sharp crystal peaks in their XRD spectra, except for the one heat-treated at 450°C. In the second experiment, the XRD spectra of the heat-treated UO2 consistently indicate the presence of U3O8 rather than UO3, regardless of the temperature. Under an oxidizing atmosphere within a temperature range of 300 to 700°C, UO2 can be oxidized to form U3O8. In the final experiment, the oxidation state of the unknown UO2+X was determined using Bragg’s law and lattice parameters, revealing that it was a material in which UO2 had been oxidized, resulting in an oxidation state of UO2.24.

To ensure the long-term supply and sustainability of uranium fuel, exploring alternative resources is essential, particularly considering that terrestrial reserves of uranium are limited (about 4.6 million tons). Since the amount of uranium dissolved in seawater is approximately 1000 times that of terrestrial reserves (i.e., about 4.5 billion tons), uranium extraction from seawater (UES) can be an alternative resource. However, the ultra-low concentration of uranium in seawater (about 3.3 ppb) poses a significant challenge in achieving economic feasibility for UES. This paper introduces case studies on the cost analysis of systems for recovering uranium from seawater, specifically focusing on braided fiber-based adsorbents developed by JAEA and ORNL. The cost analysis has been conducted based on using the deployment of these adsorbents on the bottom of the sea, which is a passive deployment method, thereby reducing the total costs of recovery. The analysis results can be used to identify R&D areas necessary for reducing cost components, making UES economically feasible.