The radioactive Sr-90, which is formed from beta decay, is well known as one of the most commonly detected nuclides in radioactive waste. In 2015, it was reported that Sr-90 was observed in some soil and metal wastes among the 516 drums of radioactive waste transferred from the decommissioning site of the Korea Research Reactor (in Seoul) to the disposal site (in Gyeongju). Decontamination and sequestration of radionuclides, including Sr, from nuclear waste is important because they are hazardous and harmful to the ecological environment. Immobilization of these nuclides using a zeolite framework is suitable and simple method that has been widely studied. Therefore, it is still necessary to continuously explore the thermal stability of various zeolites and environmental changes around adsorbed cations in zeolite pore for effective immobilization of these radionuclides. In this study, we observed the thermal stability in fully Sr-exchanged natrolite (Sr-NAT), one of small-pore zeolite, from room temperature to 350°C using the in-situ synchrotron X-ray powder diffraction and thermogravimetric (TGA) analysis. In addition, we investigated the structural changes in Sr-NAT during temperature increase by Rietveld analysis. Sr-NAT exhibited apparent zero thermal expansions (ZTE) with the thermal expansion coefficients of -3(1) × 10-6 at the initial stage of increasing the temperature due to dehydration process. In the section from 250°C to 300°C, a phenomenon like negative thermal expansion (NTE) occurs in which the unit cell volume of Sr-NAT decreases despite the increase in temperature. Sr-NAT maintained well its crystallinity up to 350°C, and it became amorphous at 350°C. In this study, we provide a fundamental understanding of the structural changes and thermal stability mechanism of Sr-exchaged zeolite natrolite with increasing temperature.

An induction melting facility includes several work health and safety risks. To manage the work health and safety risks, care must be taken to identify reasonably foreseeable hazards that could give rise to risks to health and safety, to eliminate risks to health and safety so far as is reasonably practicable. If it is not reasonably practicable to eliminate risks to health and safety, attention have to be given to minimize those risks so far as is reasonably practicable by implementing risk control measures according to the hierarchy of control in regulation, to ensure the control measure is, and is maintained so that it remains, effective, and to review and as necessary revise control measures implemented to maintain, so far as is reasonably practicable, a work environment that is without risks to health or safety. The way to manage the risks associated with induction melting works is to identify hazards and find out what could cause harm from melting works, to assess risks if necessary – understand the nature of the harm that could be caused by the hazard, how serious the harm could be and the likelihood of it happening, to control risks – implement the most effective control measures that are reasonably practicable in the circumstances, and to review control measures to ensure they are working as planned.

Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, Offgas system should be properly designed since the fission products in off-gas accelerates the corrosion in reactor structure materials and deteriorates the purity of liquid fuel. The design of off-gas system therefore requires the preliminary study of the behavior of evolved fission products in off-gas units and the development of off-gas model is crucial in developing such system. In this study, we corrected the off-gas illustrative model proposed by ORNL (Nuclear Engineering and Design, vol 385(15) 111529, 2021) by employing physically consistent concept of capture rate of fission product and holdup. For the application of the corrected off-gas model to Chloride-based 6 MW Molten Salt Reactor, major fission products were firstly determined from OpenMC based neutronics calculation and chain reaction related to the major fission products were defined. Based on these data, the holdup behavior of fission products in off-gas units (decay tank, caustic scrubber, Halide trap, H2O trap and charcoal bad) were investigated.

The dismantlement of the Kori Unit 1 and Wolsong Unit 1 nuclear power plants is scheduled. Since about 40% of the cost of dismantling nuclear power plants is the cost of disposing of generated wastes, it is important to secure recycling technologies. Among them, low and intermediate level radioactive wastes are made of porous filters and adsorbent materials of ceramic foam to remove nuclides such as C-14, I, and Xe generated during nuclear dismantling. In order to remove a large amount of nuclides, physical properties such as a specific surface area and porosity of a ceramic foam filter are important, however when a heat treatment temperature is increased to increase the strength of the filter, the nuclides removal ability is reduced. In order to remove a large amount of nuclides, physical properties such as a specific surface area and porosity of a ceramic foam filter are important, however when a heat treatment temperature is increased to increase the strength of the filter, the nuclides removal ability is reduced. Therefore, in this study, the foam filter performance was improved by applying a sacrificial material to increase the specific surface area and porosity of the ceramic foam filter. The sacrificial material is burned out with polyurethane (PU) of the green filter before the heat treatment temperature to increase the strength of the ceramic foam filter so that it can be maintained as pores, thereby improving the specific surface area and porosity. The sacrificial materials and melting temperature (Tm) reviewed in this study were anthracite (530~660°C), PMMA (160°C), Cellulose acetate (260~270°C), and aluminum particle (660°C), and their effect on the manufacture of foam filters was studied by applying this. The specific surface part and porosity of the foam filter were improved when anthracite and aluminum particle were added, and PMMA and Cellulose acetate, which are relatively low temperature melting points, were burned out at a temperature lower than PU, and thus their physical properties were not greatly affected. The physical properties and specific surface part and porosity of ceramic foam filters manufactured using various sacrificial materials will be discussed.

Korea Radioactive Waste Agency (KORAD), regulatory body and civic groups are calling for an infrastructure system that can more systematically and safely manage data on the results of radioactive waste sampling and nuclide analysis in accordance with radioactive waste disposal standards. To solve this problem, a study has been conducted on the analysis of the nuclide pattern of radioactive waste on the nuclide data contained in low-and intermediate-level radioactive waste. This paper will explain the optimal repackaged algorithm for reducing radioactive waste based on previous research results. The optimal repackaged algorithm for radioactive waste reduction is comprised based on nuclide pattern association indicators, classification by nuclide level of small-packaged waste, and nuclide concentration. Optimization simulation is carried out in the order of deriving nuclide concentration by small-packaged, normalizing drum minimization as a function of purpose, normalizing constraints, and optimization. Two scenarios were applied to the simulation. In Scenario 1 (generating facilities and repackaged by medium classification without optimization), it was assumed that there are 886 low-level drums and 52 very low-level drums. In Scenario 2 (generating facilities and repackaged by medium classification with optimization), 708 and 230 drums were assigned to the low-level and very low-level drums, respectively. As a result of the simulation, when repackaged in consideration of the nuclide concentration and constraints according to the generating facility cluster & middle classification by small package (Scenario 2) the low-level drum had the effect of reducing 178 drums from the baseline value of 886 drums to 708 drums. It was found that the reduced packages were moved to the very low-level drum. The system that manages the full life-cycle of radioactive waste can be operated effectively only when the function of predicting or tracking the occurrence of radioactive waste drums from the source of radioactive waste to the disposal site is secured. If the main factors affecting the concentration and pattern of nuclides are systematically managed through these systems, the system will be used as a useful tool for policy decisions that can prevent human error and drastically reduce the generation of disposable drums.

Dry head end process is developing for pyro-processing at KAERI (Korea Atomic Energy Research Institute). Dry processes, which include disassembling, mechanical decladding, vol-oxidation, blending, compaction, and sintering shall be performed in advance as the head-end process of pyro-processing. Also, for the operation of the head-end process, the design of the connecting systems between the down ender and the dismantling process is required. The disassembling process includes apparatus for down ender, dismantling of the SF (Spent Fuel) assembly (16×16 PWR), rod extraction, and cutting of extracted spent fuel rods. The disassembling process has four-unit apparatus, which comprises of a down ender that brings the assembly from a vertical position to a horizontal position, a dismantler to remove the upper and bottom nozzles of the spent fuel assembly, an extractor to extract the spent fuel rods from the assembly, and a cutter to cut the extracted spent fuel rods as a final step to transfer the rod-cuts to the mechanical decladding process. An important goal of dismantling process is the disassembling of a spent nuclear fuel assembly for the subsequent extraction process. In order to design the down ender and dismantler, these systems were analyzed and designed, also concept on the interference tools between down ender and dismantler were considered by using the solid works tool.

Based on the results of a review for various precipitation methods phosphorylation (phosphate precipitation) of metal chlorides considered as a proper treatment method for recovering of the fission products in a molten salt. In previous precipitation tests, the powder of lithium phosphate (Li3PO4) added into LiCl-KCl molten salt containing metal chlorides as a precipitation agent. The reaction of metal chlorides containing actinides and rare earths to recover with lithium phosphate in a molten salt known as solid-liquid reaction. The powder of lithium phosphate disperse in a molten salt by stirring thoroughly in order to enhance the precipitation reaction. As a result, metal phosphates as the reaction products precipitate on the bottom of the vessel and cutting at the lower part of the salt ingot considered as one of the recovery method of the precipitates. Recently, the vacuum distillation of upper part of the salt proposed as another recovering method. Cutting method of precipitate at the lower part of the salt ingot would be difficult to handle the increased size of the salt ingot produced from the practical scale equipment. In this presentation, a new method for collecting the precipitates of phosphorylation reaction into a small vessel is introduced with test results in a molten salt containing uranium and rare earths such as Nd, Ce, and La. As the first step of a series of test lithium phosphate ingot was prepared by melting the powder at a temperature 1,300°C, and the ingot put into LiCl-KCl molten salt at 500°C for more than three hours to examine the shape of ingot to be deformed or not. The phosphorylation experiments using lithium phosphate ingots carried out to collect the metal phosphate precipitates and the test result of this new method was feasible.

The Korea Atomic Energy Research Institute is developing a nuclide management process that separates high heat, high mobility, and long half-life nuclides that burden the disposal of spent fuel, and disposes of spent fuel by nuclide according to the characteristics of each nuclide. Various offgases (volatile and semi-volatile nuclides) generated in this process must be discharged to the atmosphere below the emission standard, so an off-gas trapping system is required. In this study, we introduce the analysis results of the parameters that affect the design of the off-gas trapping system. The analyzed contents are as follows. The physical quantities of the Cs, Tc/se, and I trapping filters according to the amount of spent nuclear fuel, the maximum exothermic temperature of the Cs trapping filter and the absorbed dose by distance by Cs radioactivity were analyzed according to the amount of spent nuclear fuel. In addition, a three-dimensional CFD (Computational Fluid Dynamics) analysis was performed according to operating parameters by simply modeling the off-gas trapping system, which is easy to modify mechanical design parameters. It is considered that the analysis results will greatly contribute to the development of the off-gas trapping system design requirements.