To secure approval for a decommissioning plan in Korea, it is essential to evaluate contamination dispersion through groundwater during the decommissioning process. To achieve this, licensees must assess the groundwater characteristics of the facility’s site and subsequently develop a groundwater flow model. It is worth noting that Combustible Radioactive Waste Treatment Facility (CRWTF) is characterized by their simplicity and absence of liquid radioactive waste generation. Given these facility characteristics, the groundwater flow model for CRWTF utilizes data from neighboring facilities, with the feasibility of using reference data substantiated through comparative analysis involving groundwater characteristic testing and on-site modeling. To enable a comparison between the actual site’s groundwater characteristics and the referenced modeling, two types of hydraulic constant characterization tests were conducted. First, hydraulic conductivity was determined through long-term pumping and recovery tests. The ‘Theis’ and ‘Cooper-Jacob’ equations, along with the ‘Theis recovery’ equation, were applied to calculate hydraulic conductivity, and the final result adopted the average of the calculated values. Secondly, a groundwater flow test was conducted to confirm the alignment between the main flow direction of the referenced model and the groundwater flow in the CRWTF, utilizing the particle tracking technique. The evaluation of hydraulic conductivity from the hydraulic constant test revealed that the measured value at the actual site was approximately 1.84 times higher than the modeled value. This variance is considered valid, taking into consideration the modeling’s calibration range and the fact that measurements were taken during a period characterized by wet conditions. Furthermore, a close correspondence was observed between the groundwater flow direction in the reference model (ranging from 90° to 170°) and the facility’s actual flow direction (ranging from 78° to 95°). The results of reference data for the CRWTF, based on the nearby facility’s model, were validated through the hydraulic properties test. Consequently, the modeling data can be employed for the demolition plan of CRWTF. It is also anticipated that these comparative analysis methods will be instrumental in shaping the groundwater investigation plans for facilities with characteristics similar to CRWTF.

Various types of radioactive liquid and solid wastes are generated during the operation and decommissioning of nuclear power plants. To remove radionuclides Co-60, Cs-137 etc. from a liquid waste, the ion-exchange process based on organic resins has been commonly used for the operation of nuclear facilities. Due to the considerations for the final disposal of process endproduct, other treatment methods such as adsorption, precipitation using some inorganic materials have been suggested to prepare for large amounts of waste during decommissioning. This study evaluated sintering characteristics for radioactive precipitates generated during the liquid waste treatment process. The volume reduction efficiency and compressive strength of sintered pellets were the major parameters for the evaluation. Major components of a simulated precipitate were some coagulated (oxy) hydroxides containing light elements, such as Si, Al, Mg, Ca, and zeolite particles. Green pellets compressed to around 100 MPa were heated at a range of 750~850°C to synthesize sintered pellets. It was observed that the volume reduction percentages were higher than 50% in the appropriate sintering conditions. The volume reduction was caused by the reduction of void space between particles, which is an evidence of partial glassification and ceramization of the precipitates. This result can also be attributed to conversion reactions of zeolite particles into other minerals. The compressive strength ranged from 6 to 19 MPa. These results also showed a significant correlation with the volume reduction of sintered body. Although our lab-scale experiments showed many benefits of sintering for the precipitates, optimized conditions are needed for large-scale practical applications. Evaluation of sintering characteristics as a function of pellet size and further testing will be conducted in the future.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. The amount of these wastes must be defined in the Final Decommissioning Plan for approval of the licensing. Also, in the case of Metal radioactive waste, it is necessary to calculate the generation amount in order to treat radioactive waste at a Radioactive Waste Treatment Facility (RWTF). Since a large quantity of metal radioactive waste is generated during the decommissioning of a nuclear power plant, the application of a metal melter for reduction is considered. The metal waste is heated to a temperature above the melting point and separated into liquid and gas forms. Nuclides existing on the surface of metal waste vaporize in a melting furnace to become dust or collect in sludge. Nonvolatile nuclides such as Co, Fe and Mn remain in ingot, but other nuclides can be captured and reduced with dust and sludge. And the types of melting furnaces to be applied can be broadly classified into Atmospheric Induction Melter (AIM) and Vacuum Induction Melter (VIM). Therefore, this review intends to compare the two types of metal furnaces to be included in RWTF.

When aluminum is in an alkaline state, the aluminum oxide film surrounding aluminum is dissolved and moisture penetrates the exposed aluminum surface, causing corrosion of aluminum. At this time, hydrogen gas is generated and there is a risk of explosion due to the generated hydrogen gas. Aluminum radioactive waste is difficult to permanently dispose of because it does not meet the standards for the acquisition of low- and intermediate-level radioactive waste cave disposal facilities currently managed and operated by the Korea Nuclear Environment Corporation. However, because of this risk, it is necessary to study how to safely treat and dispose aluminum waste. In this study, overseas cases were investigated and analyzed to ensure the safety of aluminum waste disposal, and the current status of aluminum radioactive waste generated during decommissioning of the Korea Research Reactor 1&2 and a treatment plan to secure disposal suitability were presented. The process of removing a little remaining oxygen in molten steel during the reduction of iron oxide in the iron refining process is called deoxidation, and a representative material used for deoxidation is aluminum. In the case of metal melting decontamination, which is one of the decontamination processes of radioactive metal waste, a method of treating aluminum waste by using aluminum as a deoxidizer is proposed.

In light of recent significant seismic events in Korea and worldwide, there is an urgent need to reevaluate the adequacy of seismic assessments conducted during facility construction. This study reexamines the ongoing viability of the Safety Shutdown Earthquake (SSE) criteria assessment for the Combustible Radioactive Waste Treatment Facility (CRWTF) site at the Korea Atomic Energy Research Institute (KAERI), originally established in 1994. To validate the SSE assessment, we delineated 13 seismic structure zones within the Korean Peninsula and employed two distinct methodologies. Initially, we updated earthquake occurrence data from 1994 to the present year (2023) to assess changes in the site’s horizontal maximum earthquake acceleration (g). Subsequently, we conducted a comparative analysis using the same dataset, contrasting the outcomes derived from the existing distance attenuation equation with those from the most recent attenuation equations to evaluate the reliability of the applied attenuation model. The Safety Shutdown Earthquake (SSE) criterion of 0.2 g remains unexceeded, even when considering recent earthquake events since the original evaluation in 1994. Furthermore, when applying various assessment equations developed subsequently, the maximum value obtained from the previously utilized ‘Donvan and Bornstein’ attenuation equation is 0.1496 g, closely resembling the outcome derived from the recently employed ‘Lee’ reduction equation of 0.1451 g. The SSE criteria for CRWTF remain valid in the current context, even in light of recent seismic occurrences such as the 2016 Gyeongju earthquake. Additionally, the attenuation equation employed in the evaluation consistently yields conservative results when compared to methodologies used in recent assessments. Consequently, the existing SSE criteria remain valid at present. This study is expected to serve as a valuable reference for confirming the SSE criterion assessment of similarly constructed facilities within KAERI.

Radioactive iodine-129, a byproduct of nuclear fission in nuclear power plants, presents significant environmental and health risks due to its high solubility in water and volatility. Iodine-129, with its half-life of 1.57×1017 years, necessitates safe management and disposal. Therefore, safely capturing and managing I-129 during spent nuclear fuel reprocessing is of paramount importance. To address these challenges, various glass waste forms containing silver iodide have been developed, such as borosilicate, silver phosphate, silver vanadate, and silver tellurite glasses. These glasses effectively immobilize iodine, but the high cost of silver raises affordability concerns. This study introduces CuI·Cu2O·TeO2 glass waste forms for iodine immobilization, a novel approach. The cost-effectiveness of copper, in contrast to silver, makes it an attractive alternative. The CuI·Cu2O·TeO2 glass waste forms were synthesized with varying CuI content (x) in (1-x)(0.3Cu2O·0.7TeO2) glass matrices. Xray diffraction (XRD) confirmed amorphous structures, and X-ray fluorescence (XRF) quantified composition. X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy provided insights into structural properties. Durability assessments using a 7-day product consistency test (PCT-A) and inductively coupled plasma-mass spectrometry (ICP-MS) revealed compliance with U.S. glass regulations, making CuI·Cu2O·TeO2 glasses a promising choice for iodine immobilization in radioactive waste.

The decommissioning of Korea’s nuclear power facilities is expected to take place starting with the Kori Unit 1 followed by the Wolsong Unit 1. In Korea, since there is no experience of decommissioning, considerations of site selection for the waste treatment facilities and reasonable selection methods will be needed. Only when factors to be considered for construction are properly selected and their effects are properly analyzed, it will be possible to operate a treatment facility suitable for future decommissioning projects. Therefore, this study aims to derive factors to be considered for the site selection of treatment facilities and present a reasonable selection methodology through evaluation of these factors. In order to select a site for waste treatment facilities, three virtual locations were applied in this study: warehouse 1 to warehouse 3. Such a virtual warehouse could be regarded as a site for construction warehouses, material warehouses, annexed building sites, and parking lots in nuclear facilities. If the selection of preliminary sites was made in the draft, then it is necessary to select the influencing factors for these sites. The site of the treatment facility shall be suitable for the transfer of the waste from the place where the dismantling waste is generated to the treatment facility. In addition, in order for construction to take place, interference with existing facilities and safety should not be affected, and it should not be complicated or narrow during construction. Considering the foundation and accessibility, the construction of the facility should be economical, and the final dismantling of the facility should also be easy. In order to determine one final preferred plan with three hypothetical locations and five influencing factors, there will be complex aspects and it will be difficult to maintain consistency as the evaluation between each factor progresses. Therefore, we introduce the Analytic Hierarchical Process (AHP) methodology to perform pairwise comparison between factors to derive an optimal plan. One optimal plan was selected by evaluating the three virtual places and five factors of consideration presented in this study. Given the complexity and consistency of multiple influencing factors present and prioritizing them, AHP tools help users make decisions easier by providing simple and useful features. Above all, it will be most important to secure sufficient grounds for pairwise comparison between influencing factors and conduct an evaluation based on this.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. In addition, according to the level of radioactivity, it can be divided into high level, intermediate level, low level, and clearance level waste. In the case of solid radioactive waste, it is necessary to secure disposal suitability in order to deliver it to a disposal facility, so safe and efficient treatment of a large amount of radioactive waste generated during decommissioning is one of the most important issues. For the treatment of radioactive waste generated during decommissioning, technologies in various fields such as cutting, decontamination, melting, measurement, and packaging are required. Therefore, this study intends to present and application plan for decommissioning domestic nuclear power plants through overseas case studies for the treatment of radioactive waste expected to occur during nuclear power plant decommissioning.

Decommissioning of Nuclear Power Plant (NPP) projects in South Korea starts with permanent shutdown of Kori unit 1 and Wolsung unit 1. It is important to establish a treatment and disposal method for radioactive waste generated during the decommissioning of the nuclear power plants. Large quantities of the wastes during decommissioning of NPP are generated in a short period of time and the wastes have various types and characteristics. For efficient decommissioning of NPP process, the radioactive waste is classified by types and each treatment method and packaging concept is presented respectively in this paper. Radioactive waste generated during decommissioning of NPP is classified into reactor vessel, reactor internals, metals, Dry Active Waste (DAW), concreate, spent fuel storage rack, spent resin and spent filter, etc., and the packaging concept for each type should be established to meet the waste acceptance criteria. Major waste acceptance criteria requirements include nuclides concentration, filling rate, free water, surface radiation does rate and weight. Radioactive waste containers can be classified into packaging containers, transport containers, and disposal containers. The packaging container is used to contain, transport, and store radioactive waste within the radiation control area, and a control number has been assigned as a radioactive waste drum after the final treatment has been completed. The transport container is used for transporting radioactive waste filled-containers from a radiation control area through an uncontrolled area. In this paper, the concept of disposal of dismantled radioactive waste and packaging methods were reviewed in comprehensive consideration of domestic radioactive waste transport and storage regulations, permanent disposal environment, and development status of large containers.

The nuclear facilities at Korea Atomic Energy Research Institute (KAERI) have generated a variety of liquid radioactive waste and most of them have low-level radioactive or lower levels. Some of the liquid radioactive waste generated in KAERI is transported to Radioactive Waste Treatment Facility (RWTF) in 20 L container. Liquid radioactive waste transported in a 20 L container is stored in a Sewer Tank after passing through a solid-liquid separation filter. It is then transferred to a very low-level liquid radioactive waste Tank after removing impurities such as sludge through a pre-treatment device. The previous pre-treatment process involved an underwater pump and a cartridge filter device passively, but this presented challenges such as the inconvenience of having to install the underwater pump each time, radiation exposure for workers due to frequent replacement of the cartridge filter, and the generation of large amounts of radioactive waste from the filter. To address these challenges and improve efficiency and safety in radiation work, an automated liquid radioactive waste pre-treatment device was developed. The automated liquid radioactive waste pre-treatment device is a pressure filtration system that utilizes multiple overlapping filter plates and pump pressure to effectively remove impurities such as sludge from liquid radioactive waste. With just the push of a button, the device automatically supplies and processes the waste, reducing radiation hazards and ensuring worker safety. Its modular and mobile design allows for flexible utilization in various locations, enabling efficient pre-treatment of liquid radioactive waste. To evaluate the performance of the newly constructed automated liquid radioactive waste treatment device, samples were taken before and after treatment for 1 hour cycling and analyzed for turbidity. The results showed that the turbidity after treatment was more than about four times lower than before treatment, confirming the excellent performance of the device. Also, it is expected that the treatment efficiency will improve further as the treatment time and number of cycles increase.

As an example of research activities in decontamination for decommissioning, new data are presented on the options for corrosion layer dissolution during the decommissioning decontamination, or persulfate regeneration for decontamination solutions re-use. For the management of spent decontamination solutions, new method based on solvent extraction of radionuclides into ionic liquid followed by electrodeposition of the radionuclides has been developed. Fields of applications of composite inorganic-organic absorbers or solid extractants with polyacrylonitrile (PAN) binding matrix for the treatment of liquid radioactive waste are reviewed; a method for americium separation from the boric acid containing NPP evaporator concentrates based on the TODGA-PAN material is discussed in more detail. Performance of a model of radionuclide transport, developed and implemented within the GoldSim programming environment, for the safety studies of the LLW/ILW repository is demonstrated on the specific case of the Richard repository (Czech Republic). Continuation and even broadening of these activities are expected in connection with the approaching end of the lifespan of the first blocks of the Czech NPPs.