Decommissioning waste is generated with various types and large quantities within a short period. Concrete, a significant building material for nuclear facilities, is one of the largest decommissioning wastes, which is mixed with aggregate, sand, and cement with water by the relevant mixing ratio. Recently, the proposed treatment method for volume reduction of radioactive concrete waste was proven up to scale-up testing using unit equipment, which involved sequentially thermomechanical and chemical treatment. According to studies, the aggregate as non-radioactive material is separated from cement components with contaminated radionuclides as less than clearance criteria, so the volume of radioactive concrete waste is decreased effectively. However, some supplementation points were presented to commercialize the process. Hence, the process requires efficiency as possible to minimize the interface parts, either by integration or rearranging the equipment. In this study, feasibility testing was performed using integrated heating and grinding equipment, to supplement the possible issue of generated powder and dust during the process. Previously, heat treatment and grinding devices were configured separately for pilot-scale testing. But some problems such as leakage and pipe blockage occurred during the transportation of generated fine powder, which caused difficulties in maintaining the equipment. For that reason, we studied to reduce the interface between the equipment by integrating and rearranging the equipment. To evaluate the thermal grinding performance, the fraction of coarse and concrete fines based on 1mm particle size was measured, and the amount of residual cement in each part was analyzed by wet analysis using 4M hydrochloric acid. The result was compared with previous studies and the thermomechanical equipment could be selected to enhance the process. Therefore, it is expected that the equipment for commercialization could be optimized and composed the process compactly by this study.

Currently, the most promising fuel candidate for use in sodium fast reactors (SFRs) is metallic fuel, which is produced by a modified casting method in which the metallic fuel material is sequentially melted in an inert atmosphere to prevent volatilization, followed by melting in a graphite crucible, and then injection casting in a quartz (SiO2) mold to produce metallic fuel slugs. In previous studies, U-Zr metallic fuel slugs have been cast using Y2O3 reaction prevent coatings. However, U-Zr alloy-based metallic fuel slugs containing highly reactive rare earth (RE) elements are highly reactive with Y2O3-coated quartz (SiO2) molds and form a significant thickness of surface reaction layer on the surface of the metallic fuel slug. Cast parts that have reacted with nuclear fuel materials become radioactive waste. To decrease amount of radioactive waste, advanced reaction prevent material was developed. Each RE (Nd, Ce, Ln, Pr) element was placed on the reaction prevent material and thermal cycling experiments were carried out. In casting experiments with U-10wt% Zr, it was reported that Y2O3 layer has a high reaction prevent performance. Therefore, the reaction layer properties for RE elements with higher reactivity than uranium elements were evaluated. To investigate the reaction layer between RE and NdYO3, the reaction composition and phase properties as a function of RE content and location were investigated using SEM, EDS, and XRD. The results showed that NdYO3 ceramics had better antireaction performance than Y2O3.

Radioactive waste generated in large quantities from NPP decommissioning has various physicochemical and radiological characteristics, and therefore treatment technologies suitable for those characteristics should be developed. Radioactively contaminated concrete waste is one of major decommissioning wastes. The disposal cost of radioactive concrete waste is considerable portion for the total budget of NPP decommissioning. In this study, we developed an integrated technology with thermomechanical and chemical methods for volume reduction of concrete waste and stabilization of secondary waste. The unit devices for the treatment process were also studied at bench-scale tests. The volume of radioactive concrete waste was effectively reduced by separating clean aggregate from the concrete. The separated aggregate satisfied the clearance criteria in the test using radionuclides. The treatment of secondary waste from the chemical separation step was optimally designed, and the stabilization method was found for the waste form to meet the final disposal criteria in the repository site. The final volume reduction rates of 56.4~75.4% were possible according to the application scenario of our processes under simulated conditions. The commercial-scale system designs for the thermomechanical and chemical processes were completed. Also, it was found that the disposal cost for the contaminated concrete waste at domestic NPP could be reduced by more than 20 billion won per each unit. Therefore, it is expected that the application of this technology will improve the utilization of the radioactive waste disposal space and significantly reduce the waste disposal cost.

Decommissioning waste is generated at all stages during the decommissioning of nuclear facilities, and various types of radioactive waste are generated in large quantities within a short period. Concrete is a major building material for nuclear facilities. It is mixed with aggregate, sand, and cement with water by the relevant mixing ratio and dried for a certain period. Currently, the proposed treatment method for volume reduction of radioactive concrete waste was involved thermomechanical and chemical treatment sequentially. The aggregate as non-radioactive materials is separated from cement components as contaminated sources of radionuclides. However, to commercialize the process established in the laboratory, it is necessary to evaluate the scale-up potential by using the unit equipment. In this study, bench-scale testing was performed to evaluate the scale-up properties of the thermomechanical and chemical treatment process, which consisted of three stages (1: Thermomechanical treatment, 2: Chemical treatment, 3: Wastewater treatment). In the first stage, lab, bench, and pilot scale thermomechanical tests were performed to evaluate the treated coarse aggregate and fines. In the second stage, the fine particles generated by the thermomechanical treatment process, were chemically treated using dissolution equipment, after then the removal efficiency and residual of cement in the small aggregate was compared with laboratory results. The final stage, the secondary wastewater containing contaminant nuclides was treated, and the contaminant nuclides could be removed by chemical precipitation method in the scale-up reactors. Furthermore, an additional study was required on the solid-liquid separation, which connected each part of the equipment. It was conducted to optimize the separation method for the characteristics of the particles to be separated and the purpose of separation. Therefore, it is expected that the basic engineering data for commercialization was collected by this study.

Strong acidic wastewater containing a radionuclide is generated from the decontamination of radioactively contaminated wastes or equipment. This wastewater is generally treated though a precipitation process using an alkali (alkali earth) hydroxides. In this precipitation process, a significant amount of alkali (alkali earth) sulfates are generated, which is the reason for the increase in the radioactive waste generation. In this study, a method for separating only radionuclides and metal ions from the wastewater was evaluated. For this reason, precipitation behaviors of radionuclides and metal ions by hydrazine injections were investigated using equilibrium calculations. In addition, behaviors of hydrazine decomposition after removal of radionuclides and metal ions were analyzed for recycling the wastewater.

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.