Wolsong Unit 1, a domestic heavy water reactor nuclear power plant, was permanently shut down in December 2019. Accordingly, Wolsong Unit 1 plans to prepare a Final Decommissioning Plan (FDP), submit it to the government by 2024, receive approval for decommissioning, and begin full-scale decommissioning. One of the important tasks in the decommissioning of Wolsong Unit 1 is to determine the decommissioning strategy. It is necessary to decide on a decommissioning strategy considering various factors and variables, secure the technical background, and justify it. The selection of a decommissioning strategy is best achieved through the use of formal decisionmaking assistance techniques, such as considerations related to influencing factors. It is very important to understand the basic decommissioning strategy alternatives and whether sufficient consideration has been given to situations where only a single unit is permanently shut down in a multi-unit site like Wolsong Unit 1, while the remaining units are in normal operation. As a process for selecting a decommissioning strategy, first, all considerations that could potentially affect decommissioning presented in the KINS Decommissioning Safety Review Guidelines were synthesized, influencing factors to be used in the decision-making process were determined, and the concept was defined. In order to select the most appropriate decommissioning strategy by considering various evaluation attributes of possible decommissioning alternatives (immediate dismantling and delayed dismantling), the Wolsong Unit 1 decommissioning strategy was evaluated by reflecting the AHP decision-making technique.

Domestic commercial low- and intermediate-level radioactive waste storage containers are manufactured using 1.2 mm thick cold-rolled steel sheets, and the outer surface is coated with a thin layer of primer of 10~36 μm. However, the outer surface of the primer of the container may be damaged due to physical friction, such as acceleration, resonance, and vibration during transportation. As a result, exposed steel surfaces undergo accelerated corrosion, reducing the overall durability of the container. The integrity of storage containers is directly related to the safety of workers. Therefore, the development of storage containers with enhanced durability is necessary. This paper provides an analysis of mechanical properties related to the durability of WC (tungsten carbide)-based coating materials for developing low- and intermediate-level radioactive waste storage containers. Three different WC-based coating specimens with varied composition ratios were prepared using HVOF (high-velocity oxy-fuel) technique. These different specimens (namely WC-85, WC-73, and WC-66) were uniformly deposited on cold-rolled steel surfaces ensuring a constant thickness of 250 μm. In this work, the mechanical properties of the three different WCbased coaitng materials evaluated from the viewpoints of microstructure, hardness, adheision force between substrate and coating material, and wear resistance. The cross-sectional SEM-EDS (Scanning Electron Microscope-Energy Dispersive X-ray Spectroscopy) images revealed that elements W (tungsten), C (carbon), Ni (nickel), and Cr (chromium) were uniformly distributed within the each coating layers which was approximately 250 μm thick. The average hardness values of HWC-85 and HWC-73 were found to be 1,091 Hv (Vickers Hardness) and 1,083 Hv, respectively, while the HWC-66 exhibited relatively lower hardness value of 883 Hv. This indicates that a higher WC content results in increased hardness. Adhesion force between and substrates and coating materials exceeded 60 MPa for all specimens, however, there were no significant differences observed based on the tungsten carbide content. Furthermore, a taber-type abrasion tester was used for conducting abrasion resistance tests under specific conditions including an H-18 load weight at 1,000 g with rotational speed set at 60 RPM. The abrasion resistance tests were performed under ambient temperatures (RT: 23±2°C) as well as relative humidity levels (RH: 50±10%). Currently, the ongoing abrasion resistance tests will include some results in this study.

Radioactive Oxide is formed on the surface of the coolant pipe of the nuclear power plant. In order to remove the oxide film that is formed on the surfaces of the coolant pipe, chemical and physical decontamination technologies are used. The disadvantage of traditional technologies is that they produce secondary radioactive wastes. Therefore, in this study, the short-pulsed laser eco-friendly technology was used in order to reduce the production of secondary radioactive wastes. It was also used to minimize the damage that was caused to the base material and to remove the contaminated oxide film. The study was carried out using a Stainless steel 304 specimen that was coated with nickel-ferrite particles. Additionally, a transport robot was 3D modeled and manufactured in order to efficiently remove the oxide film from the coolant pipe of the nuclear power plant. The transport robot has a fixed laser head to move inside the horizontal and vertical pipes. The rotating laser head removes the contaminated oxide film on the inner surface of the coolant pipe. In the future, as a condition of the 1064nm short-pulsed laser ablation technique determined by basic analysis, we plan to analyze whether the transport robot is applicable to the radiation contamination site of the nuclear power plant.

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.

APro, a modularized process-based total system performance assessment framework, was developed at the Korea Atomic Energy Research Institute (KAERI) to simulate radionuclide transport considering coupled thermal-hydraulic-mechanicalchemical processes occurring in a geological disposal system. For reactive transport simulation considering geochemical reactions, COMSOL and PHREEQC are coupled with MATLAB in APro using an operator splitting scheme. Conventionally, coupling is performed within a MATLAB interface so that COMSOL stops the calculation to deliver the solution to PHREEQC and restarts to continue the simulation after receiving the solution from PHREEQC at every time step. This is inefficient when the solution is frequently interchanged because restarting the simulation in COMSOL requires an unnecessary setup process. To overcome this issue, a coupling scheme that calls PHREEQC inside COMSOL was developed. In this technique, PHREEQC is called through the “MATLAB function” feature, and PHREEQC results are updated using the COMSOL “Pointwise Constraint” feature. For the one-dimensional advection-reaction-dispersion problem, the proposed coupling technique was verified by comparison with the conventional coupling technique, and it improved the computation time for all test cases. Specifically, the more frequent the link between COMSOL and PHREEQC, the more pronounced was the performance improvement using the proposed technique.

Recently, it is being carried out the project to evaluate the properties of materials harvested from nuclear reactor after the decommissioning of Kori Unit 1. However, it is not sufficient adequate machining equipment and remote machining technique to perform the projects for evaluation of materials harvested from nuclear reactor. Thus, it is required to develop the remote machining technique in hotcell to evaluate the mechanical properties of nuclear reactor materials. The machining technique should be performed inside a hotcell to evaluate mechanical properties of materials harvested from nuclear reactor and is essential to prevent radiation exposure of workers. Also, it is essential to design the apparatus and develop the machining process so that it can be operated with a manipulator and minimize contamination in hotcell. In this research, development of remote specimen machining technique in hotcell such as machining apparatus, technique and process for compact tension specimens of material harvested from nuclear reactor are described. Remote machining technique will be useful in specimen machining to evaluate changes in mechanical properties of materials harvested in high-radioactive reactor. Also, it is expected that various types of specimens can be machining by applying the developed machining technique in the future.

The hydro-mechanical behavior of rock mass in natural barriers is a critical factor of interest, and it is mainly determined by the characteristics of the fractures distributed in the rock mass. In particular, the aperture and contact area of the fractures are important parameters directly related to the fluid flow and significantly influence the hydro-mechanical behavior of natural barriers. Therefore, it is necessary to analyze the aperture and contact area of fractures distributed in potential disposal sites to examine the long-term evolution of the natural barriers. This study aims to propose a new technique for analyzing the aperture and contact area using the natural fractures in KURT (KAERI Underground Research Tunnel), an underground research facility for the deep geological disposal of high-level radioactive waste. The proposed technique consists of a matching algorithm for the three-dimensional point cloud of the upper and lower fracture surfaces and a normal deformation algorithm that considers the fracture normal stiffness. In the matching process of upper and lower fracture surfaces, digital images obtained from compression tests with pressure films are used as input data. First, for the primary matching of the upper and lower fracture surfaces, an iterative closest point (ICP) algorithm is applied in which rotation and translation are performed to minimize the distance error. Second, an algorithm for rotation about the x, y, and z axes and translation in the normal direction is applied so that the contact area of the point cloud is as consistent as possible with the pressure film image. Finally, by applying the normal deformation algorithm considering the fracture normal stiffness, the aperture and contact area of the fracture according to the applied normal stress are derived. The applicability of the proposed technique was validated using 12 natural fractures sampled from KURT, and it was confirmed that the initial apertures were derived similarly to the empirical equation proposed in the previous study. Therefore, it was judged that the distribution of apertures and contact areas according to applied normal stress for laboratory-scale fractures could be derived through the technique proposed in this study.

A tensile test is performed to obtain the mechanical property data of the spent fuel cladding. In general, the elastic modulus, elongation, yield stress, tensile stress, etc. are obtained by axial tensile test of cladding attaching an extensometer. However, due to the limitation in the number of specimens for spent nuclear fuel that can be made, the ring tensile test (RTT) whose required length of the specimen is short is mainly performed. In the case of RTT, an extensometer or strain gauge cannot be attached because the gauge part of the specimen is formed around the cladding and is short. In addition, since a load is applied in the radial direction of the cladding, a curved portion of the circular cladding is spread out and becomes straight, and then the cladding is tensioned. For this reason, it is difficult to obtain the stress-strain curve directly from the RTT results. Isight, which is used to identify the optimization design parameters, was used to build an optimization process that minimizes the difference between the RTT and the analysis to estimate the material property. For this, the elastic modulus, plastic strain, and the radius of the RTT jig were taken as fixed variables. As variables, isotropic hardening data and plastic stress were taken. The objective function was taken as the minimization of the area difference of the load-displacement curve obtained from the tests and analysis, of the difference in the magnitude of the maximum reaction force, and of the difference in the location where the maximum reaction force occurred. Optimization workflow was configured in the following order. First, using the calculator component, plastic stress design variables were created. Next, ABAQUS was placed to perform analysis using design variables, and the reaction force or displacement was calculated. After that, the reaction force was calculated considering the 1/4 symmetry condition using the script component. After that, the data matching component performed quantitative comparison of test and analysis data. Finally, by utilizing the exploration component, the plastic stress design variable that minimizes the difference in the objective function was obtained by automatically changing six optimization algorithms. In this paper, the constructed optimization process and the obtained plastic stress by applying it to the SUS316 RTT results are briefly described. The established optimization process can be utilized to obtain mechanical property from the results of the cladding RTT of spent nuclear fuel or new material.

The buffer is a critical barrier component in an engineered barrier system, and its purpose is to prevent potential radionuclides from leaking out from a damaged canister by filling the void in the repository. No experimental parameters exist that can describe the buffer expansion phenomenon when Kyeongju bentonite, which is a buffer candidate material available in Korea, is exposed to groundwater. As conventional experiments to determine these parameters are time consuming and complicated, simple swelling pressure tests, numerical modeling, and machine learning are used in this study to obtain the parameters required to establish a numerical model that can simulate swelling. Swelling tests conducted using Kyeongju bentonite are emulated using the COMSOL Multiphysics numerical analysis tool. Relationships between the swelling phenomenon and mechanical parameters are determined via an artificial neural network. Subsequently, by inputting the swelling tests results into the network, the values for the mechanical parameters of Kyeongju bentonite are obtained. Sensitivity analysis is performed to identify the influential parameters. Results of the numerical analysis based on the identified mechanical parameters are consistent with the experimental values.

For spent nuclear fuel transferred to dry storage facilities, it is difficult to apply safeguards approaches and long-term integrity verification due to the structural characteristics of the facility. There is a need to check the integrity of the nuclear fuel assembly before transferring it to a dry storage facility and are need to provide information on whether there are any defects. At the Korea Institute of Nuclear Nonproliferation and Control, as a non-destructive testing technology for ensuring Continuity of Knowledge (CoK) of the dry storage facilities, a methodology for reconstructing images by neutron tomographic technique from spent nuclear fuel using a He-4 gas scintillation detector was presented. It is thought that the He-4 gas scintillation detector-based technology can be used to verify the defect of the nuclear fuel assembly. This methodology must be accompanied by accurate neutron measurements. The place where the technique was conducted is surrounded by a concrete wall. Concrete contains water molecules, which can affect neutron measurements. In this study, reconstruction images based on neutron measurements and MCNP simulations are compared to verify the effects of the concrete. Neutron measurements were performed by measuring Cf-252 neutron sources in a 1/10 lab-scale TN- 32 cask with six He-4 gas scintillation detectors as an array. Neutron sources are fixed at each point in the cask, and the He-4 detector array is rotated from 0° to 360° at 10° intervals to reconstruct the image using the filtered back-projection (FBP) method. Also, in MCNP reconstructed images, there are two versions depending on whether concrete wall. The source image and ring shape were found in the measurement-based thermal neutron reconstruction image, which was similar to the simulation image that considering the concrete effects. On the other hand, in the simulation reconstruction image without the concrete, only the shape of the source was found. Thus, the effect of concrete should be considered when performing the neutron tomographic techniques using He-4 gas scintillation detectors.

In this study, a drop analysis of metallic disposal containers for radioactive wastes is performed according to accident scenarios at the disposal site. The weight of the disposal container is about 8 tons, and the ingot-type wastes are loaded in the disposal container. To simulate the floor of the disposal site as the impact target, the reinforced concrete pad is modeled. High impact energy of the disposal container due to their heavy weight and high drop height causes excessive deformation and failure of the concrete target having relatively weak strength. Dynamic growth of cracks due to such failures causes penetration and delamination of concrete. Since the impact force delivered to the container strongly depends on the failure of the concrete pad, it is important to properly simulate the failure of the concrete in the drop analysis. A material erosion method can be used to simulate the concrete failure. In the case of applying erosion based on the finite element method (FEM), the element is deleted when the element exceeds a certain criterion, which causes material and energy loss problem. To solve this problem, mesh-free methods such as smoothed particle hydrodynamics (SPH) can be commonly used, but the mesh-free method has the disadvantage of incurring high numerical cost. Therefore, an adaptive method combining SPH and FEM-based SOLID elements is used for concrete target modeling to simulate excessive deformation and failure of the concrete target. In the adaptive coupling method of SPH and SOLID, the concrete target is first modeled as a solid element. When the damage of concrete exceeds the failure criterion, the solid element is eroded and the SPH element replacing the solid element is activated. Since the activated SPH element continues to participate in the impact, the problem of loss of materials and energy can be effectively solved. In this way, analysis results consistent with actual physical phenomena can be obtained.

Waste that contains or is contaminated with radionuclides arises from a number of activities involving the use of radioactive material. Such activities include the operation and decommissioning of nuclear facilities; the use of radionuclides in medicine, industry, agriculture, research and education. Radioactive waste must be safely disposed in a radioactive waste repository for the protection of public health and the environment. In order to safely dispose of radioactive waste in a repository, it is important to derive an optimal predisposal management scenario because radioactive waste must be processed (i.e. processing (pretreatment, treatment and conditioning), storage and transport) for satisfying waste acceptance criteria (WAC). Optimal scenario of predisposal management of radioactive waste is derived for considering the balancing of exposures of workers and/or those of members of the public, the short term and long term risk implications of different waste management strategies, the technological options available and the costs. However, existing studies for deriving the optimal scenario of predisposal management of radioactive waste have evaluated only the radiation dose of workers and public within given scenarios using fixed value, or have derived optimal single process (i.e. decontamination) of predisposal management using Multi-Attribute Decision Making (MADM) methodology. In this study, optimal predisposal management scenario is derived by evaluating exposures of workers using system dynamics (SD) technique. Radiation dose assessment SD model was modeled using VENSIM® code developed by VENTANA systems Inc.. SD Model has the advantage of being able to respond flexibly when decision makers want to change input data and it has the advantage of being able to track dynamically changing phenomena and visually confirm interdependence. After that, based on the SD model derived from this study, evaluations of exposures of public, cost, and technicality will be added to be utilized when establishing an optimal scenario of predisposal management of radioactive waste considering multi attribute.

Various models have been proposed to describe the swelling behavior of buffer in high level waster repository. One of the most notable models, the Barcelona Basic Model (BBM), is a mechanical model that simulates the behavior of unsaturated ground and is widely applied to soils that undergo large expansion due to water. Among the BBM parameters of Kyeongju bentonite, which is found in Korea, there are no experimental data for parameters that describe the unsaturated state. Such hydromechanical properties should be characterized through experimental programs. However, such experiments are highly complicated and require long periods of time to produce an unsaturated state through different methods according to the suction range. Although there are several studies in which geotechnical parameters were obtained through a back analysis instead of direct experiments, few studies have employed machine learning methods for the identification of geotechnical parameters. In this study, instead of direct experiments, the results of a relatively simple swelling pressure experiment was compared to the numerical analysis results to propose a method of determining some of BBM parameters. Influential factors were identified by a sensitivity analysis and the values of the factors were estimated using an artificial neural network and optimization method. The obtained parameters were applied to the numerical model to estimate the swelling pressure growth, which was subsequently compared to the experimental value. As a result, it was found that there was no significant difference between the two swelling values.

Molten salt immersion technique has been tested with several Sr oxides, SrZrO3, SrMoO4 and U2SrOy, and MgCl2 based molten salts for the Sr nuclide separation. Reaction time, temperature, and salt composition were varied to effectively separate Sr in chloride forms. ICP-OES, XRD, and SEM analysis were conducted for the conversion efficiency and structure and morphology analysis. It is confirmed that all experiments of SrZrO3 with MgCl2 at 800°C for reaction time 5, 10, 20 hours showed higher conversion efficiency than 99% and in LiCl-KCl-MgCl2 and NaCl-MgCl2 molten salts at 500°C or 600°C, conversion efficiency higher than 97% was obtained. SrMoO4 in MgCl2 immersion experiments for 10 hours showed higher conversion efficiency than 99% when the molar ratio of salt/oxide powder is 7. U2SrOy was also tested with MgCl2 molten salt at 800°C and higher efficiency than 99% and mainly MgUO4 were produced as a reaction product.

In this paper, among the W-S-R relationship methods proposed by Lee, et. al., (2020) to produce rain-based rain information in real time, we tried to produce actual rain information by applying machine learning techniques to take into account the effects of wiper operation. To this end, rain sensor proposed the Graded Descent and Threshold Map method for pre-processing the cumulative value of the difference before and after wiper operation by utilizing four sensitive channels for optical sensors developed by Kim Byung-sik (2016) and using rain sensor data produced by five rain conditions in indoor artificial rainfall experiments. This method is the method of producing rainfall information by calculating the average value of the Threshold according to the rainfall conditions and channels, creating a Threshold Map corresponding to the grid 4 (channel) x 5 (thinking of rainfall information) and applying Optima Rainfall Intensity among the big data processing techniques. For the verification of these proposed findings, the application was evaluated by comparing the rainfall observations with the methods presented by Lee, et. al., (2020).