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.

In 2012, POSIVA selected a bentonite-based (montmorillonite) block/pellet as the backfilling solution for the deposition tunnel in the application for a construction license for the deep geological repository of high-level radioactive waste in Finland. However, in the license application (i.e. SC-OLA) for the operation submitted to the Finnish Government in 2021, the design for backfilling was changed to a granular mixture consisting of bentonite (smectite) pellets crushed to various sizes, based on NAGRA’s buffer solution. In this study, as part of the preliminary design of the deep geological repository system in Korea, we reviewed history and its rationale for the design change of Finland’s deposition tunnel backfilling solution. After the construction license was granted by the Finnish Government in 2015, POSIVA conducted various lab- and full-scale in-situ tests to evaluate the producibility and performance of two design alternatives (i.e. block/pellet type and granular type) for backfilling. Principal demonstration tests and their results are summarized as follows: (a) Manufacturing of blocks using three types of materials (Friedland, IBeco RWC, and MX-80): Cracking and jointing under higher pressing loads were found. Despite adjusting the pressing process, similar phenomena were observed. (b) 1:6 scale experiment: Confirmation of density difference inhomogeneity due to the swelling of block/pellet backfill and void filling due to swelling behavior into the mass loss area of block/pellet. (c) FISST (Full-Scale In situ system Test): Identification of technical unfeasibility due to the inefficient (too manual) installation process of blocks/pellets and development of an efficient granular in-situ backfilling solution to resolve the disadvantage. (d) LUCOEX-FE (Large Underground Concept Experiments – Full-scale Emplacement) experiment: Confirmation of dense/homogeneous constructability and performance of granular backfilling solution. In conclusion, the simplified granular backfill system is more feasible compared to the block/ pellet system from the perspective of handling, production, installation, performance, and quality control. It is presumed that various experimental and engineering researches should be preceded reflecting specific disposal conditions even though these results are expected to be applied as key data and/or insights for selecting the backfilling solution in the domestic deep geological repository.

Rock discontinuities in underground rock behave as weak planes and affect the safety of underground structures, such as high-level radioactive waste disposal and underground research facilities. In particular, rock discontinuities can be a main flow path of groundwater and induce large deformation caused by stress disturbance or earthquakes. Therefore, it is essential to investigate the characteristics of rock discontinuities considering in-situ conditions when constructing highlevel radioactive waste disposal, which needs to assure the long-term safety of the structure. We prepared Hwang-Deung granite rock block specimens, including a saw-cut rock surface, to perform multi-stage direct shear tests as a preliminary study. In the multi-stage direct shear tests, we can exclude possible errors induced by different specimens for obtaining a full failure envelope by using an identical specimen. We applied the initial normal stress of 3 MPa on the specimen and increased the normal stress to 5 and 10 MPa step by step after peak shear stress observation. We obtained the mechanical properties of saw-cut rock surfaces from the experiments, including friction coefficient and cohesion. Additionally, we investigated the effect of filling material between rock discontinuities, assuming the erosion and piping phenomenon in the buffer material of the engineering barrier system. When the filling material existed in the rock surfaces, the shear characteristics deteriorated, and the effect of bentonite was dominant on the shear behavior.

The engineered barrier system (EBS), composed of spent nuclear fuel, canister, buffer and backfill material, and near-field rock, plays a crucial role in the deep geological repository for high-level radioactive waste. Understanding the interactions between components in a thermo-hydro-mechanical -chemical (THMC) environment is necessary for ensuring the long-term performance of a disposal facility. Alongside the research project at KAERI, a comprehensive experimental facility has been established to elucidate the comprehensive performance of EBS components. The EBS performance demonstration laboratory, which installed in a 1,000 m2, consists of nine experimental modules pertaining to rock mechanics, gas migration, THMC characteristics, buffer-rock interaction, buffer & backfill development, canister corrosion, canister welding, canister performance, and structure monitoring & diagnostics. This facility is still conducting research on the engineering properties and complex interactions of EBS components under coupled THMC condition. It is expected to serve as an important laboratory for the development of the key technologies for assessing the long-term stability of engineered barriers

The high-level radioactive waste repository must ensure its performance for a long period of time enough to sufficiently reduce the potential risk of the waste, and for this purpose, multibarrier systems consisting of engineered and natural barrier systems are applied. If waste nuclides leak, the dominating mechanisms facilitating their movement toward human habitats include advection, dispersion and diffusion along groundwater flows. Therefore, it is of great importance to accurately assess the hydrogeological and geochemical characteristics of the host rock because it acts as a flow medium. Normally, borehole investigations were used to evaluate the characteristics and the use of multi-packer system is more efficient and economical compared to standpipes, as it divides a single borehole into multiple sections by installing multiple packers. For effective analyses and groundwater sampling, the entire system is designed by preselecting sections where groundwater flow is clearly remarkable. The selection is based on the analyses of various borehole and rock core logging data. Generally, sections with a high frequency of joints and evident water flow are chosen. Analyzing the logging data, which can be considered continuous, gives several local points where the results exhibit significant local changes. These clear deviations can be considered outliers within the data set, and machine learning algorithms have been frequently applied to classify them. The algorithms applied in this study include DBSCAN (density based spatial clustering of application with noise), OCSVM (one class support vector method), KNN (K nearest neighbor), and isolation forest, of which are widely used in many applications. This paper aims to evaluate the applicability of the aforementioned four algorithms to the design of multi-packer system. The data used for this evaluation were obtained from DB-2 borehole logging data, which is a deep borehole locates near KURT.

The compacted bentonite buffer is a key component of the engineered barrier system in deep geological repositories for high-level radioactive waste disposal. Groundwater infiltration into the deep geological repository leads to the saturation of the bentonite buffer. Bentonite saturation results in bentonite swelling, gelation and intrusion into the nearby rock discontinuities within the excavation damaged zone of the adjacent rock mass. Groundwater flow can result in the erosion and transport of bentonite colloids, resulting in bentonite mass loss which can negatively impact the long-term integrity and safety of the overall engineered barrier system. The hydro -mechanicalchemical interactions between the buffer, surrounding host rock and groundwater influence the erosion characteristics of the bentonite buffer. Hence, assessing the critical hydro-mechanicalchemical factors that negatively affect bentonite erosion is crucial for the safety design of the deep geological repository. In this study, the effects of initial bentonite density, aperture, discontinuity angle and groundwater chemistry on the erosion characteristics of Bentonil WRK are investigated via bentonite extrusion and artificial fracture experiments. Both experiments examine bentonite swelling and intrusion into simulated rock discontinuities; cylindrical holes for bentonite extrusion experiments and plane surfaces for artificial fracture experiments. Compacted bentonite blocks and bentonite pellets are manufactured using a compaction press and granulation compactor respectively and installed in the transparent extrusion cells and artificial fracture cells. The reference test condition is set to be 1.6 g/cm3 dry density and saturation using distilled water. After distilled water or solution injection, the axial and radial expansion of the bentonite specimens into the simulated rock discontinuities are monitored for one month under free swelling conditions with no groundwater flow. Subsequent flow tests are conducted using the artificial fracture cell to determine the critical flow rate for bentonite erosion. The intrusion and erosion characteristics are modelled using a modified hydro-mechanicalchemical coupled dynamic bentonite diffusion model and a fluid-based hydro-mechanical penetration model.

The presence of technological voids in deep geological repositories for high-level radioactive nuclear waste can have negative effects on the hydro-mechanical properties of the engineered barrier system when groundwater infiltrates from the surrounding rock. This study conducted hydration tests along with image acquisition and X-ray CT analysis on compacted Korean bentonite samples, which simulated technological voids filling to investigate the behavior of fracturing (piping erosion) and cracking deterioration. We utilized a dual syringe pump to inject water into a cell consisting of a bentonite block and technological voids at a consistent flow rate. The results showed that water inflow to fill technological voids led to partial hydration and self-sealing, followed by the formation of an erosional piping channel along the wetting front. After the piping channel generated, the cyclic filling-piping stage is characterized by the repetitive accumulation and drop of water pressure, accompanied by the opening and closing of piping channels. The stoppage of water inflow leads to the formation of macro- and micro cracks in bentonite due to moisture migration caused by high suction pressure. These cracks create preferential flow paths that promote longterm groundwater infiltration. The experimental test and analysis are currently ongoing. Further experiments will be conducted to investigate the effects of different dry density in bentonite, flow rate, and chemical composition of injected water.

Since the first operation of the Gori No. 1 nuclear power plant in Korea was started to operate in 1978, currently 24 nuclear power plants have been being operated, out of which 21 plants are PWR types and the rest are CANDU types. About 30% of total electricity consumed in Korea is from all these nuclear power plants. The accumulated spent nuclear fuels (SFs) generated from each site are temporarily being stored as wet or dry storage type at each plant site. These SFs with their high radiotoxicity, heat generating, and long-lived radioactivity are actually the only type of high-level radioactive waste (HLW) in Korea, which urgently requires to be disposed of in deep geological repository. Studies on disposal of HLW in various kind of geological repositories have been carried out in such countries as Sweden, Finland, United States, and etc. with their own methodologies and management policies in consideration of their situations. In Korea long-term R&D research program for safe management of SF has also been conducted during last couple of decades since around 1997, during which several various alternative type of disposal concepts for disposal of SNFs in deep geological formations have been investigated and developed. The first concept developed was KAERI Reference Disposal System (KRS) which is actually very much similar to Swedish KBS-3, a famous concept of direct disposal of SF in stable crystalline rock at a depth of around 500 m which has been regarded as one of the most plausible method worldwide. The world first Finnish repository which is expected to begin to operate sooner or later will be also this type. Since the characteristics of SF discharged from domestic nuclear reactors have been changed and improved, and burnup has sometimes increased, a more advanced deep geological repository system has been needed, KRS-HB (KRS with High Burnup SF) has been developed and in consideration of the dimensions of SNFs and the cooling period at the time point of the disposal time, KRS+, a rather improved disposal concept has also been subsequently developed which is especially focused on the efficient disposal area. Recently research has concentrated on rather advanced disposal technology focused on a safer and more economical repository system in recent view of the rapidly growing amount of accumulated SF. Especially in Korea the rock mass and the footprint area for the repository extremely limited for disposal site. Some preliminary studies to achieve rather higher efficiency repository concept for disposal of SF recently have already been emphasized. Among many possible ones for consideration of design for high-efficiency repository system, a double-layered system has been focused which is expected to maximize disposal capacity within the minimum footprint disposal area. Based on such disposal strategy a rather newly designed performance assessment methodology might be required to show long-term safety of the repository. Through the study some prerequisites for such methodological development has been being roughly checked and investigated, which covers FEP identification and pathway and scenario analyses as well as preliminary conceptual modeling for the nuclide release and transport in nearfield, far-field, and even biosphere in and around the conceptual repository system. Through the study such scenarios and models has been implemented to development of a safety assessment by utilizing GoldSim development tool for a rough quantitative comparison with existing disposal options and simple illustration purpose as well as for showing how to develop and implementation of the model to GoldSim templet.

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.

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.

The concept of deep geological disposal for high-level radioactive waste is based on an engineered barrier system (EBS), including a canister, bentonite buffer and backfill material. The bentonite buffer is key component of the EBS to prevent groundwater infiltration and radionuclide leakage. However, the bentonite buffer can become saturated due to groundwater flow through the excavation damaged zone in the adjacent rock, causing erosion of bentonite buffer and affecting the long-term performance of EBS. While the RH (relative humidity) sensor is commonly used to assess the degree of saturation in the bentonite buffer, it has a critical challenge due to its sensor size, which can disturb the overall integrity of the bentonite buffer during the initial installation process. In contrasts, the electrical resistivity test, widely known as a non-destructive method, is used to predict soil properties such as the degree of saturation and water contents. This method measures the electric resistance of materials using electric current induced by electric potential difference between two electrodes. Notably, there is no study that assess the integrity of bentonite buffer in a nuclear waste repository using electrical resistivity measurement. This study presents the electrical resistance numerical module under steady state using commercial finite element method (FEM), and quantitatively estimate the change of electrical resistance according to saturation and erosion of bentonite buffer. Furthermore, the electric potential and current density distribution formed between two electrodes are analyzed.

The HADES (High-level rAdiowaste Disposal Evaluation Simulator) was developed by the Nuclear Fuel Cycle & Nonproliferation (NFC) laboratory at Seoul National University (SNU), based on the MOOSE Framework developed by the Idaho National Laboratory (INL). As an application of the MOOSE Framework, the HADES incorporates not only basic MOOSE functions, such as multi-physics analysis using Finite Element Method (FEM) and various solvers, but also additional functions for estimating the performance assessment of Deep Geological Repositories (DGR). However, since the MOOSE Framework does not have complex mesh generation and data analyzing capabilities, the HADES has been developed to incorporate these missing functions. In this study, although the Gmsh, finite element mesh generation software, and Paraview, finite element analysis software, were used, other applications can be utilized as well. The objectives of HADES are as follows: (i) assessment of the performance of a Spent Nuclear Fuel (SNF) disposal system concerning Thermal-Hydraulic-Mechanical-Chemical (THMC) aspects; (ii) Evaluation of the integrity of the Engineered Barrier System (EBS) of both general and high-efficiency design perspective; (iii) Collaboration with other researchers to evaluate the disposal system using an open-source approach. To achieve these objectives, performance assessments of the various disposal systems and BMTs (BenchMark Test), conducted as part of the DECOVALEX projects, were studied regarding TH behavior. Additionally, integrity assessments of various DGR systems based on thermal criteria were carried out. According to the results, HADES showed very reasonable results, such as evolutions and distributions of temperature and degree of saturation, when compared to validated code such as TOUGH-FLAC, ROCMAS, and OGS (OpenGeoSys). The calculated data are within the range of estimated results from existed code. Furthermore, the first version of the code, which can estimate the TH behavior, has been prepared to share the contents using Git software, a free and open-source distribution system.

For the sake of future generations, the management of radioactive waste is essential. The disposal of spent nuclear fuel (SNF) is considered an urgent challenge to ensure human safety by storing it until its radioactivity drops to a negligible level. Evaluating the safety of disposal facilities is crucial to guarantee their durability for more than 100,000 years, a period sufficient for SNF radioactivity to become ignored. Past studies have proposed various parameters for forecasting the safety of SNF disposal. Among these, radiochemistry and electrochemistry play pivotal roles in predicting the corrosion-related chemical reactions occurring within the SNF and the structural materials of disposal facilities. Our study considers an extreme scenario where the SNF canister becomes compromised, allowing underground water to infiltrate and contact the SNF. We aim to improve the corrosion mechanism and mass-balance equation compared with what Shoesmith et al. proved under the same circumstances. To enhance the comprehensibility of the chemical reactions occurring within the breached SNF canister, we have organized these reactions into eight categories: mass diffusion, alpha radiolysis, adsorption, hydrate formation, solidification, decomposition, ionization, and oxidation. After categorization, we define how each species interacts with others and calculate the rate of change in species’ concentrations resulting from these reactions. By summing up the concentration change rates of each species due to these reactions, we redefine the mass-balance equations for each species. These newly categorized equations, which have not been explained in detail previously, offer a detailed description of corrosion reactions. This comprehensive understanding allows us to evaluate the safety implications of a compromised SNF canister and the associated disposal facilities by numerically solving the mass-balance equations.

As part of the preparation of a glossary of terminologies related to the disposal of spent nuclear fuel, definitions of potentially issuable terminologies used in domestic regulations were inferred from relevant regulations or comparatively analyzed with foreign definitions. These terminologies are safety assessment and performance assessment, safety function and safety performance, disposal containers and package, isolation and containment, and so on. Their concise and easy-to-understand definitions have been proposed in order to obtain these opinions of stakeholders.

In KAERI, the nuclide management technology is currently being developed for the reduction of disposal area required for spent fuel management. Among the all fission products of interest, Cs, I, Kr, Tc are considered to be significantly removed by following mid-temperature and hightemperature treatment, however, a difficulty of real spent-fuel thermal treatment experiment limits the development of such thermal treatment. The test employing SimFuel (Simulated Spent Fuel) can be an alternative for such condition, however, the fabrication of SimFuel containing semivolatile species such as Cs, I and Re (substitute for Tc) was not achieved for conventional sintering method since such species are easily removed during hot temperature treatment. In this study, for the prevention of volatilization of such species and the inclusion of semi-volatile species in fabrication of SimFuel, argon-based high pressurizing up to Max 100 bar was considered to be applied in high temperature treatment. For this, lab-scale hot-isostatic press applicable up to 1,500°C was fabricated and is being waiting for the approval for high-pressure test. After approval of license, UO2 baesd SimFuel containing CsI will be fabricated and its micro-structure and composition will be evaluated through SEM-EDX and XRD

The PRIDE scale mechanical decladder is decladding apparatus for separating and recovering fuel material and cladding hull by horizontally slitting rod-cut. In order to enhance mechanical decladdng efficiency, the main requirements were considered as follows. Decladding of the fuel rods may be performed by rotation of three circular cutting blades inserted among the rollers arranged at 120° portion. In a mechanical decladder, a slitting assembly as a unit for slitting the cladding tube may include cutting blades for slitting and rollers for guiding extrusion of the cladding tube. Rotation of the cutting blades may be caused by the fuel rods being extruded from a plurality of rollers. Slitting intervals of rod-cuts having different diameters may be controlled by adding or removing a spacing plate between the cutting blade and a ranch bolt for fixing the slitting blade to the slitting assembly. An extrusion velocity with respect to the fuel rods may be controlled by a hydraulic pressure applied to the fuel rods. A force for cutting the fuel rods may be adjusted by controlling steel plates. Forces applied to a plurality of rollers may be generated by the hydraulic cylinder. The hydraulic pressure may be controlled by hydraulic pressure controller. The PRIDE scale mechanical decladder mainly consists of auto feeding module, hydraulic cylinder module and blade module. A load cell was installed between the hydraulic cylinder and the extrusion pin to measure the decladding force and slitting velocity, and a data acquisition system capable of obtaining data by using the RSC 232 was constructed. Also, the control panel can control the forward and backward movement of the extrusion pin, the hydraulic flow rate, and the hydraulic velocity. In the mechanical decladding test, 40 pieces of simulated rod-cuts were loaded in two auto feeding basket and slit by utilizing the 3-CUT blade modules in the housing, and hulls and simulated pellets were collected in the collection container. As a result, 80 pieces of simulated rodcut (brass pellets + Zry4 tube) were slit continuously without any problem. About 35 min was required to slit 80 rod-cuts and average decladding force was 260 kg. The decladding force of the ceramic simulated rod-cuts (castable) requires 25 kg less force than the brass pellets. Therefore, it is estimated that the spent fuel rod-cut can be fully split into three pieces using the mechanical decladder.

Dry storage of nuclear fuel is compromised by threats to the cladding integrity, such as creep and hydride reorientation. To predict these phenomena, spent fuel simulation codes have been developed. In spent fuel simulation, temperature information is the most influential factor for creep and hydride formation. Traditional fuel simulation codes required a user-defined temperature history input which is given by separate thermal analysis. Moreover, geometric changes in nuclear fuel, such as creep, can alter the cask’s internal subchannels, thereby changing the thermal analysis. This necessitates the development of a coupled thermal and nuclear fuel analysis code. In this study, we integrated the 2D FDM nuclear fuel code GIFT developed at SNU with COBRA -SFS. Using this, we analyzed spent nuclear stored in TN-24P dry storage cask over several decades and identified conditions posing threats due to phenomena like creep and hydrogen reorientation, represented by the burnup and peak cladding temperature at the start of dry storage. We also investigated the safety zone of spent nuclear fuel based on burnup and wet storage duration using decay heat.

Post Irradiation Examination Facility (PIEF) is a test facility for nuclear fuel research and development and performance evaluation. From the past to the present, assemblies and fuel rods have been transported from nuclear power plants (NPP) several times, and various destructive and non-destructive tests have been performed. Among these, in the case of the 14×14 Westinghouse STD assemblies that are transported as a whole assembly, the top nozzle is connected to the guide tube by welding. Therefore, the fuel rods could not be removed from the assembly at the NPP, so the assemblies were transported to PIEF as is. Then, after cutting between the top nozzle and the guide tube in the pool, and the fuel rods were extracted and tested. In order to transport the assembly in the future, it is necessary to maintain stability by inserting the dummy rod into the unit cell from which the fuel rod is extracted. However, since the length of the dummy rod is almost 4 m and the diameter is about 10 mm, the dummy rod often bends while passing through the dimple spring of the grid. Additionally, when dummy rods are inserted into unit cells that are continuously empty after the fuel rods are extracted, there may be cases where the dummy rods are not inserted into the desired unit cell but are bent and incorrectly inserted into the next unit cell. The moment the dummy rods are inserted into the dimple spring of grid, a load is applied to the dummy rod due to the tension of the spring. If it can be inserted while offsetting the load, the work can be performed more smoothly. Accordingly, an underwater handling tool was developed that can be inserted while offsetting the tension of the spring. Using this handling tool applies a load to the dummy rod and rotates the dummy rod itself, offsetting the tension of the spring and allowing the dummy rod to be inserted without bending. This handling tool is equipped with a shock absorbing device to protect the dummy rod and spring, and a module to rotate the dummy rod. As a result of inserting the dummy rod using the developed handling tool, it was possible to easily insert the dummy rod into unit cells that were previously impossible to insert.

In general, systems are developed by repeatedly performing the processes of design, analysis, manufacturing, and performance testing. In particular, systems with temperature, pressure, and flow rate often utilize computational fluid dynamics tools at the design stage. In this paper, we aim to verify the reliability of the analysis results of Solidworks Flow Simulation, which is widely used in heat flow analysis at the design stage. A tube furnace was manufactured, various experiments were performed, and a study was conducted to compare the analysis results. The details of the experiment are as follows. First, an experiment was conducted in which the heater was heated to 900°C without insulating the exposed part of the tube. The detailed contents of the experiment are as follows; - Heating heater and measuring temperature without supplying flow inside the tube, - Tube flow supply (25°C, 15 lpm air) and heater heating/temperature measurement. Second, an experiment was performed in which the exposed part of the tube was insulated (thickness 50 mm) and the heater was heated to 900°C. The detailed contents of the experiment are as follows; - Insulate the outside of the tube except for the flanges at both ends of the tube, and heat the heater and measure the temperature without supplying flow inside the tube. - Insulate the outside of the tube except for the flanges at both ends of the tube, supply flow rate inside the tube (25°C, 15 lpm air) and measure heater heating/temperature. - Insulate the flange of the flow supply section, heat the heater and measure temperature without supplying flow inside the tube. - Insulate the flange of the flow supply section, heat the supply air (277°C, 15 lpm) and measure the temperature using a heating gun without heating the heater. - Insulate the flange of the flow supply section, supply heated air (277°C, 15 lpm) and measure heater heating/temperature. - Insulate the flange of the flow supply section and measure temperature according to heater heating (900°C) and supply temperature (25°C, 277°C 15 lpm). The following results were derived from the experimental and analysis results. - When the exposed part of the tube is insulated, the temperature inside the tube increases and the steady-state power decreases compared to non-insulated. - In areas with insulation, the temperature error between experiment and analysis results is not large. - When flow rate is supplied, there is a large temperature error in experiment and analysis results. - The temperature change after the center of the heater is not large for a temperature change of 15 lpm flow rate. From these results, it can be seen that Solidworks Flow Simulation has a significant difference from the experimental results when there is a flow rate in the tube. This was thought to be because the flow rate acts as a disturbance, and this cannot be sufficiently accounted for in the analysis. In the future, we plan to check whether there is a way to solve this problem.