• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1009-1014
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• 2023

Muon tomography is a useful method for monitoring special nuclear materials (SNMs) such as spent nuclear fuel inside dry cask storage. Multiple Coulomb scattering of muons can be used to provide information about the 3-dimensional structure and atomic number(Z) of the inner materials. Tomography using muons is less affected by the shielding material and less harmful to health than other measurement methods. We developed a muon detector for muon tomography, which consists of a plastic scintillator, 64 long wavelength-shifting (WLS) fibers attached to the top of the plastic scintillator, and silicon photomultipliers (SiPMs) connected to both ends of each WLS fiber. The muon detector can acquire X and Y positions simultaneously using a position determination algorithm. The design parameters of the muon detector were optimized using DETECT2000 and Geant4 simulations, and a muon detector prototype was built based on the results. Spatial resolution measurement was performed using simulations and experiments to evaluate the feasibility of the muon detector. The experimental results were in good agreement with the simulation results. The muon detector has been confirmed for use in a muon tomography system.

• Journal of Radiation Protection and Research
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• v.47 no.4
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• pp.226-236
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• 2022

Background: An effective communication strategy for reducing conflicts in South Korea has been designed through the analysis of public perception and communication variables on nuclear power under the conditions of rapidly changing nuclear power policies. Materials and Methods: This study conducted both qualitative research through group discussions based on social psychology and quantitative research through surveys. Results and Discussion: Nuclear power plant (NPP) area residents in favor of nuclear power indicated higher levels of communication, safety perception, and contribution than those against it. NPP area residents trusted the civilian expert groups (18.3%) and local government (17.3%) the most, while metropolitan city residents trusted the Nuclear Safety and Security Commission and the Korea Institute of Nuclear Safety (20.7%) the most. In determining nuclear power policy, both the NPP area residents (18.1%) and metropolitan city residents (17.1%) prioritized safety, health, and the environment. While metropolitan city residents thought that energy security and economic growth (16.4%) were important, NPP area residents thought the current issue of spent fuel rods (14.1%) to be important. Conclusion: It is necessary for the nuclear power industry to have and actively implement communication and conflict resolution strategies based on the patterns obtained in the study results.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.411-417
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• 2023

During the operation of a nuclear power plant (NPP), the generation of radioactive waste, including dry active waste (DAW), concentrates, spent resin, and filters, mandates the implementation of appropriate disposal methods to adhere to Korea's waste acceptance criteria (WAC). In this context, this study investigates the potential use of polymer concrete (PC) as a high-integrity container (HIC) material for solidifying and packaging these waste materials. PC is a versatile composite material comprising binding polymers, aggregates, and additives, known for its exceptional strength and chemical stability. A comprehensive analysis of PC's long-term integrity was conducted in this study. First, its compressive strength, which is crucial for ensuring the structural stability of HICs over extended periods, was evaluated. Subsequently, the resilience of PC was tested under various stress conditions, including biological, radiological, thermal, and chemical stressors. The findings of this study indicate that PC exhibits remarkable long-term properties, demonstrating exceptional stability even when subjected to diverse stressors. The results therefore underscore the potential viability of PC as a reliable material for constructing high-integrity containers, thus contributing to the safe and sustainable management of radioactive waste in NPPs.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1913-1924
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• 2024

This paper presents the design and operation of an autonomous robot for pyroprocess automation, which requires unique approaches beyond those used in industrial applications to achieve the desired performance. Maintaining an extremely dry atmosphere is crucial to handle various materials, including chloride, and an autonomous system ensures this dry environment. The drying room dehumidifier was carefully selected and designed to generate dry air, and different types of dry air conditioning performance were evaluated, including assessing worker accessibility inside the mock-up to determine the system's feasibility. Containers used for process materials were modified to fit the gripper system of the gantry robot for automation. The loading and unloading of process materials in each equipment were automatically performed to connect the process equipment with the robotic system. The gantry robot primarily operated through macro motion to approach waypoints containing process materials, eliminating the need for precise approach motion. The robot's tapered jaw design allowed it to grasp the target object even with imperfect positioning. Robot motions were programmed using a robot simulator for initial positioning and motion planning, and real accuracy was tested in a mock-up facility using the OPC platform.

• Corrosion Science and Technology
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• v.23 no.4
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• pp.266-277
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• 2024

Due to its good corrosion and heat resistance with excellent mechanical properties, 304L stainless steel is commonly used in the fabrication of spent nuclear fuel dry storage canisters. However, welds are sensitive to stress corrosion cracking (SCC) due to residual stress generation. Although SCC resistance can be improved by stress relieving the weld and changing the chloride environment, it is difficult to change corrosion environment for certain applications. Stress control in the weld can improve SCC resistance. Ultrasonic shot peening (USP) needs further research as compressive residual stresses and microstructure changes due to plastic deformation may play a role in improving SCC resistance. In this study, 304L stainless steel was welded to generate residual stresses and exposed to a chloride environment after USP treatment to improve SCC properties. Effects of USP on SCC resistance and crack growth of specimens with compressive residual stresses generated more than 1 mm from the surface were studied. In addition, correlations of compressive residual stress, grain size, intergranular corrosion properties, and pitting potential with crack propagation rate were determined and the improvement of SCC properties by USP was analyzed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.2
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• pp.211-217
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• 2024

The growing significance of sustainable energy technologies underscores the need for safe and efficient management of spent nuclear fuels (SNFs), particularly via deep geological disposal (DGD). DGD involves the long-term isolation of SNFs from the biosphere to ensure public safety and environmental protection, necessitating materials with high corrosion resistance for DGD canisters. This study investigated the feasibility of a Cu-Ni film, fabricated via additive manufacturing (AM), as a corrosion-resistant layer for DGD canister applications. A wire-fed AM technique was used to deposit a millimeter-scale Cu-Ni film onto a carbon steel (CS) substrate. Electrochemical analyses were conducted using aerated groundwater from the KAERI underground research tunnel (KURT) as an electrolyte with an NaCl additive to characterize the oxic corrosion behavior of the Cu-Ni film. The results demonstrated that the AM-fabricated Cu-Ni film exhibited enhanced corrosion resistance (manifested as lower corrosion current density and formation of a dense passive layer) in an NaCl-supplemented groundwater solution. Extensive investigations are necessary to elucidate microstructural performance, mechanical properties, and corrosion resistance in the presence of various corroding agents to simplify the implementation of this technology for DGD canisters.

• Journal of the Korea Concrete Institute
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• v.26 no.1
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• pp.71-78
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• 2014

The long term integrity of concrete cask is very important for spent nuclear fuel dry storage system. However, there are serious concerns about early deterioration of concrete cask from creaking and corrosion of reinforcing steel by chloride ion because the cask is usually located in seaside, expecially by combined deterioration such as chloride ion and heat, carbonation. This study is to investigate the relation between temperature and chloride ion diffusion of concrete. Immersion tests using 3.5% NaCl solution that were controlled in four level of temperature, i.e. 20, 40, 65, and $90^{\circ}C$, were conducted for four months. The chloride ion diffusion coefficient of concrete was predicted based on the results of profiles of Cl- ion concentration with the depth direction of concrete specimens using the method of potentiometric titration by $AgNO_3$. Test results indicate that the diffusion coefficient of chloride ion increases remarkably with increasing temperature, and there was a linear relation between the natural logarithm values of the diffusion coefficients and the reciprocal of the temperature from the Arrhenius plots. Activation energy of concrete in this study was about 46.6 (W/C = 40%), 41.7 (W/C = 50%), 30.7 (W/C = 60%) kJ/mol under a temperature of up to $90^{\circ}C$, and concrete with lower water-cement ratio has a tendency towards having higher temperature dependency.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.207-215
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• 2020

The feasibility of rare earth (RE) removal process via oxidation reactions with UCl₃ was investigated using the HSC Chemistry code to reduce the concentrations of RE in transuranic (TRU) products. The composition and thermodynamic data of TRU and RE elements contained in the reference spent fuel were examined. The reactivity was evaluated by calculating equilibrium data considering oxidation reactions with UCl₃. Both RE removal rate and TRU recovery rate were evaluated for the two cases, wherein TRU products with different RE concentrations were used. When TRU products were reacted with UCl₃, selective oxidation was driven by the difference in the Gibbs free energy of each element. The calculation results imply that the TRU/RE ratio of the final product can be increased by removing RE elements while maintaining the maximum recovery rate of TRU, which is accomplished by controlling the amount of UCl₃ injected. Since the results of this study are based on thermodynamic equilibrium data, there are many limitations to apply to the actual process. However, it is expected to be used as an important data for the process design to supply the TRU product of pyroprocessing to SFR's fuel demanding low RE concentrations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.261-267
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• 2010

The pyrometallurgical nuclear fuel recycle process, called pyroprocessing, has been known as a promising nuclear fuel recycling technology. Pyroprocessing technology is crucial to advanced nuclear systems due to increased nuclear proliferation resistance and economic efficiency. The basic concept of pyroprocessing is group actinide recovery, which enhances the nuclear proliferation resistance significantly. One of the key steps in pyroprocessing is "electrowinning" which recovers group actinides with lanthanide from the spent nuclear fuels. In this study, a vertical cadmium distiller was manufactured. The evaporation rate of pure cadmium in vertical cadmium distiller varied from 12.3 to $40.8g/cm^2/h$ within a temperature range of 773 923 K and pressure below 0.01 torr. Uranium - cadmium alloy was fabricated by electrolysis using liquid cadmium cathode in a high purity argon atmosphere glove box. The distillation behavior of pure cadmium and cadmium in uranium - cadmium alloy was investigated. The distillation behavior of cadmium from this study could be used to develop an actinide recovery process from a liquid cadmium cathode in a cadmium distiller.

• Analytical Science and Technology
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• v.12 no.3
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• pp.177-183
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• 1999

The determination method of free acid in spent U-Al nuclear fuel solutions by Gran plot titration was described. Effect of U(VI) and Al(III) on the alkalimetric titration of nitric acid was investigated in oxalate complexing media as well as in noncomplexing media. Positive biases were observed in both titration media when the end-point was estimated by the Gran plot method. It was found that the cause of the bias was U(VI) in the oxalate complexing media, but Al(III) in the noncomplexing media. The relative error was less than 1% in the titration of 0.1 M $HNO_3$ at a U(VI) : Al(III) : $H^+$ mole ratio of up to 2:12:1 as long as the pH of the oxalate titration media was sustained to be below 5.0 at the beginning of titration. The method was successfully applied to the determination of nitric acid in a solution of HANARO reactor fuel with U:Al mole ratio of 1:6.