Plasma torch melting technology can pyrolyze and melt waste with high-temperature heat (about 1,600°C) using electric arc phenomena such as lightning. Waste that may be treated in a plasma torch melting facility is injected in solid (combustible, non-combustible) and liquid form depending on facility capacity. The 200 L drum type, screw supply type, and nozzle type liquid injection device are applied to MW plasma facilities, and the push rod type and screw supply type are applied to smallcapacity plasma facilities. In consideration of the characteristics of radioactive waste generated from operating and dismantling nuclear power plants, a waste input device suitable for plasma torch facilities was developed and verified through tests. In the future, facility soundness will be confirmed through long-term performance tests, and stability will be secured through continuous improvement.

The high-level nuclear waste disposal system is a structure with a very long life expectancy, and deterioration and cracking of the structure may occur over time. In addition, the high-level nuclear waste disposal system is in complex extreme conditions such as high temperature, groundwater, and radiation. Therefore, we need to develop a highly durable monitoring sensor that can detect the deterioration and crack of structures in extreme conditions. Since the durability of a sensor is closely related to the sensor lifetime, it is essential to predict the sensor lifetime accurately. The sensor lifetime can be predicted through the reliability qualification test. Among them, the accelerated life test conducted under harsh conditions is widely used as a method to shorten the test period. The major factor in carrying out the accelerated life test is to set the appropriate harsh conditions. Therefore, this study experimentally derived the operating limit of the monitoring sensor. It is essential to set the proper harsh conditions when performing the accelerated life test. Through this study, it is judged that it will be helpful in determining the appropriate stress level when performing the accelerated life test for accurate lifetime prediction.

APro, developed by KAERI as a process-based total system performance assessment model, can simulate the radionuclide transport affected by thermal, hydraulic, mechanical and geochemical changes that may occurs in the engineering and natural barriers of a geological disposal system. APro targets a large-scale and heterogeneous 3D system that includes more than 10,000 boreholes located about 500 m underground and hundreds of fractures of different sizes distributed within an area of several km2. Simulating transport and reaction phenomena for such a system through the global implicit approach (GIA) may require considerable computational resources or be intractable in some cases. Therefore, APro adopts the sequential non-iterative approach (SNIA), one of the operator splitting (OS) methods, to separate the mass transport and reaction phenomena into independent problems. By using SNIA, the parallel computation performance in APro with multiple cores is expected to be improved. In this study, the effect of SNIA on the parallel computation performance was analyzed through a simple 1D reactive transport problem. Without SNIA, finite difference equations, discretized from the partial differential equations (PDEs) describing the reactive transport problem, have to be solved at once because all dependent variables are nonlinearly and spatially interconnected through reaction and mass transport terms. When the reaction and mass transport terms are separated through SNIA, the mass transport problem can be converted into independent linear equations for each chemical and the efficient linear system solver can be applied to each linear equation. In particular, since the reaction problem is changed to independent nonlinear equations for each node, the parallel computation performance can be greatly improved. To verify this, the 1D reactive transport problem was implemented in MATLAB, and SNIA and GIA were applied to solve the problem. As a result, there was no significant difference in results between SNIA and GIA for proper spatial and temporal discretization, which verified the accuracy of SNIA. In order to see the parallel computation performance, the calculation times for SNIA and GIA with increasing number of cores were measured and compared. As the number of cores increased, the SNIA calculation speed became faster than that of GIA, which verified that SNIA could improve parallel computation performance in APro. In the future, the effect of SNIA on the parallel computation performance will be verified for the numerical analysis of large-scale geological disposal systems.

The manufactured nuclear fuel assembly is loaded into the nuclear reactor after the core design, and is finally discharged to the wet storage pool after depletion for 3 cycles. The discharged spent nuclear fuel is transported and stored in a dry storage system at the on-site of the nuclear power plant, which is cooled by natural convection, and undergoes final disposal or reprocessing through an intermediate dry storage facility. In this series of processes, the characteristics of the final product, the spent fuel, vary depending on the environmental conditions, so it is essential to manage each history data to verify the long-term integrity of the spent nuclear fuel. In this paper, safety information on spent nuclear fuel is described in order to establish technical requirements that should be considered in each stage of storage, transport, reprocessing, and disposal of spent nuclear fuel. Comprehensive safety information on spent nuclear fuel is basically calculated from basic information that considers characteristic information that can be obtained through the manufacture and design of nuclear fuel assemblies, operation history in a nuclear reactor, and location history in a wet storage pool. It can be divided into secondary production information (SF Burnup, Nuclide Inventory, etc.) and tertiary integrity-related information obtained through cladding inspection during spent fuel storage. KHNP produces this multi-layered information according to the production stage and manages it through the comprehensive management system of the spent nuclear fuel, and safety information with some errors is not only improved through re-verification but also continuously updated. In this paper, the spent nuclear fuel safety information was derived based on various information calculated in the entire process of being discharged and managed in a wet storage pool, including new fuel manufacturing information and depletion history. Such safety information will be used as basic data for long-term safe management of spent nuclear fuel, and will be continuously produced and managed. In the future, additional discussions will be held on the safety information of the spent nuclear fuel through consultation with KORAD and regulatory agencies.

Inaccurate amounts of active ingredients in disinfectants can result in inefficiency or in potential toxicity to the environment and living organisms. This may lead to disinfection failure, and consequently, biosecurity failure. To ensure the efficiency of disinfectants and their ongoing compliance with safety and quality requirements, continuous and systematic post-market surveillance studies are needed. Herein, the content of ten commercial disinfectants purchased in 2021 was analyzed. Selective analytical techniques, such as automated titration, colorimetry, and high-performance liquid chromatography, were used to evaluate the content of several active ingredients present in disinfectants, such as potassium peroxymonosulfate, benzalkonium chloride, glutaraldehyde, phosphoric acid, citric acid, and malic acid. The obtained values were then compared with the label claims; the active ingredient contents of all disinfectants were within the acceptable range of 90–120% of the label claim.