A large amount of radioactive wastes are generated during the decommissioning of the nuclear power plant. The need for radioactive waste management is emerging as the permanent suspension of domestic nuclear power plants Gori Unit 1 and Wolseong Unit 1 has been decided. According to the analysis of the number of land transportation of Gori Unit 1, when the dismantling period is expected to be five years, the number of land transportation is 459. Accordingly, measures are needed to improve the acceptability of residents on land transportation routes. Currently, there is little preparation for the acceptance of residents by transporting radioactive waste in Korea. This study analyzed the literature related to radioactive waste and nuclear power generation to derive factors that affect the acceptance of residents on the transport route of radioactive waste. Factors Affecting Resident Acceptability • Mistrust of the measured dose levels themselves • fear of radiation • Lack of training in radiation, nuclear power • Insufficient information for easy identification of dose and concentration • economic compensation In order to improve the acceptability of residents when transporting radioactive waste on land, it is necessary to scientifically prove the safety of the low and intermediate level radioactive waste transport path, and policy improvement considering the acceptability of residents is needed. In subsequent studies, it is necessary to specifically derive solutions to the above factors. This study is significant in that it derived factors that could affect low and intermediate level radioactive waste transportation, considering that no countermeasures have been prepared to improve the acceptance of residents by transporting low and intermediate level radioactive waste in Korea.

KHNP is carrying out international technical cooperation and joint research projects to decommission Wolsong unit 1 reactor. Construction data of the reactor structures, experience data on the pressure tube replacement projects, and the operation history were reviewed, and the amount of dismantled waste was calculated and waste was classified through activation analysis. By reviewing COG (CANDU owners Group) technical cooperation and experience in refurbishment projects, KHNP’s unique Wolsong unit 1 reactor decommissioning process was established, and basic design of a number of decommissioning equipment was carried out. Based on this, a study is being conducted to estimate the worker dose of dismantling workers. In order to evaluate the dose of external exposure of dismantling workers, detailed preparation and dismantling processes and radiation field evaluation of activated structures are required. The preparation process can be divided into dismantlement of existing facilities that interfere with the reactor dismantling work and construction of various facilities for the dismantlement process. Through process details, the work time, manpower, and location required for each process will be calculated. Radiation field evaluation takes into account changes in the shape of structures by process and calculates millions of areas by process, so integrated scripts are developed and utilized to integrate input text data. If the radiation field evaluation confirms that the radiation risk of workers is high, mutual feedback will be exchanged so that the process can be improved, such as the installation of temporary shields. The results of this study will be used as basic data for the final decommissioning plan for Wolsong unit 1. By reasonably estimating the dose of workers through computer analysis, safety will be the top priority when decommissioning.

According to IAEA PRIS, there is no record of dismantling commercial heavy water reactors among 57 heavy water reactors around the world. In Canada, which has the largest number of heavy water reactors, three of the 22 commercial heavy water reactors with more than 500 MW are permanently suspended, Gentilly unit 2 (2012), Pickering unit 2 (2007), and Pickering unit 3 (2008), all of which chose a delayed decommissioning strategy. On the other hand, Wolsong unit 1, which will be the world’s first heavy water reactor to be dismantled commercially, will be immediately carried out as a decommissioning strategy. KHNP has established various cooperation systems with advanced companies and international organizations related to overseas NPP decommission and is actively exchanging technologies. Among them, the most important focus is on research cooperation related to COG (CANDU owners Group). The first case is a joint study on Conceptual Calandria Segmentation. Four areas of process, waste management, ALARA, and cost for decommissioning reactors to be submitted to Canadian regulators for approval of Pickering and Gentilly-2’s preliminary decommissioning plan have been evaluated, and research on Wolsong unit 1 is currently underway. The second case is Decommissioning and long-term waste management R&D. Although the technical maturity is low, it studies the common interests of member companies in the decommissioning of heavy water reactor power generation companies and long-term waste management. Robotics for dismantling high-radiation structures, C- 14, H-3 measurement and removal methods, and concrete decontamination technology, which are characterized by heavy water, are being actively studied. KHNP is strengthening international cooperation with COG to prepare for the successful decommissioning of Wolsong unit 1. Based on previous studies by Pickering and Gentilly-2, an evaluation of the decommissioning of Wolsong unit 1 reactor is being conducted. In addition, it is preparing for decommissioning through experience analysis of the pressure tube replacement project.

Decommissioning plan of nuclear facilities require the radiological characterizations and the establishment of a decommissioning process that can ensure the safety and efficiency of the decommissioning workers. By utilizing the rapidly developed ICT technology, we have developed a technology that can acquire, analyze, and deliver information from the decommissioning work area to ensure the safety of decommissioning workers, optimize the decommissioning process, and actively respond to various decommissioning situations. The established a surveillance system that monitors nuclide inventory and radiation dose distribution at dismantling work area in real time and wireless transmits data for evaluation. Developed an evaluation program based on an evaluation model for optimizing the dismantling process by linking real-time measurement information. We developed a technology that can detect the location of dismantling workers in real time using stereovision cameras and artificial intelligence technology. The developed technology can be used for safety evaluation of dismantling workers and process optimization evaluation by linking the radionuclides inventory and dose distribution in dismantling work space of decommissioning nuclear power plant in the future.

Radioactive Oxide is formed on the surface of the coolant pipe of the nuclear power plant. In order to remove the oxide film that is formed on the surfaces of the coolant pipe, chemical and physical decontamination technologies are used. The disadvantage of traditional technologies is that they produce secondary radioactive wastes. Therefore, in this study, the short-pulsed laser eco-friendly technology was used in order to reduce the production of secondary radioactive wastes. It was also used to minimize the damage that was caused to the base material and to remove the contaminated oxide film. The study was carried out using a Stainless steel 304 specimen that was coated with nickel-ferrite particles. Additionally, a transport robot was 3D modeled and manufactured in order to efficiently remove the oxide film from the coolant pipe of the nuclear power plant. The transport robot has a fixed laser head to move inside the horizontal and vertical pipes. The rotating laser head removes the contaminated oxide film on the inner surface of the coolant pipe. In the future, as a condition of the 1064nm short-pulsed laser ablation technique determined by basic analysis, we plan to analyze whether the transport robot is applicable to the radiation contamination site of the nuclear power plant.

In this study, we evaluated the performance of phosphate-functionalized silica in adsorbing uranium and provided insights into optimizing the initial conditions of the uranium solution (concentration and pH), which are often overlooked in uranium adsorption studies. While most studies take into account the effect of pH on both the surface charge of the adsorbents and the dissolved speciation of uranium in solution, they often overlook the formation of solid phases such as β-UO2(OH)2 (cr) and UO3· 2H2O(cr), leading to an overestimation of the adsorption capacity. To address this issue, we considered the speciation of U(VI) calculated using thermodynamic data. Our findings suggest that it is reasonable to evaluate the adsorption performance at pH 4 and concentration below 1.35 mM. The formation of β-UO2(OH)2 (cr) starts at 23 μM (pH 5) and 1 μM (pH 6) and increases sharply with increasing concentration. To avoid interference from the formation of solid phases, experiments should be conducted at lower concentrations, which in turn require very small msorbent/Vsolution ratios. However, controlling small amounts of sorbent can be challenging, and increasing the volume of the solution can generate significant amounts of radioactive waste. We also used UV-vis spectra analysis to investigate the formation of solid phases. We found that a 100 mg L-1 uranium solution resulted in the formation of colloidal particles in the solid phase after 2.5 hours at pH 6, while at pH 4, no significant changes in absorbance were observed over 120 hours, indicating a stable ion phase. Based on these conditions, we obtained an excellent adsorption capacity of 110 mg g-1.

Aluminum’s exceptional properties, such as its high strength-to-weight ratio, excellent thermal conductivity, corrosion resistance, and low neutron absorption cross-section, make it an ideal material for diverse nuclear industry applications, including aluminum plating for the building envelope of nuclear power plants. However, plating aluminum presents challenges due to its high reactivity with oxygen and moisture, thus, complicating the process in the absence of controlled environments. Plating under an inert atmosphere is often used as an alternative. However, maintaining an inert atmosphere can be expensive and presents an economic challenge. To address these challenges, an innovative approach is introduced by using deep eutectic solvents (DES) as a substitute for traditional aqueous electrolytes due to the high solubility of metal salts, and wide electrochemical window. In addition, DESs offer the benefits of low toxicity, low flammability, and environmentally friendly, which makes DESs candidates for industrial-scale applications. In this study, we employed an AlCl3-Urea DES as the electrolyte and investigated its potential for producing aluminum coatings on copper substrates under controlled conditions, for example, current density, deposition duration, and temperature. A decane protective layer, non-polar molecular, has been used to shield the AlCl3-Urea electrolyte from the air during the electrodeposition process. The electrodeposition was successful after being left in the air for two weeks. This study presents a promising and innovative approach to optimizing aluminum electrodeposition using deep eutectic solvents, with potential applications in various areas of the nuclear industry, including fuel cladding, waste encapsulation, and radiation shielding.

Radioactive waste generated during nuclear power plant decommissioning is classified as radioactive waste before the concentration is identified, but more than 90% of the amount generated is at a level that can be by clearance. However, due to a problem in the analysis procedure, the analysis is not carried out at the place of on-site but is transported to an external institution to identify concentration, which implies a problem of human error because 100% manual. As a way to solve this problem, research is underway to develop a mobile radioactive waste nuclide analysis facility. The mobile radionuclide analysis facility consists of a preparation room, a sample storage room, a measurement room, a pretreatment room, and a waste storage room, and is connected to an external ventilation facility. In addition, since the automation module is built-in from the sample pre-threatening step to the separation step, safety can be improved and rapid analysis can be performed by being located in the decommissioning site. As an initial study for the introduction of a mobile nuclide analysis facility, Visiplan, a preliminary external exposure evaluation code, was used to derive the analysis workload by a single process and evaluate the exposure dose of workers. Based on this, as a follow-up study, the amount of analysis work according to the continuous process and the exposure dose of workers were evaluated. As a result of the evaluation, the Regulatory dose limit was satisfied, and in future studies, internal and external exposure doses were evaluated in consideration of the route of movement, and it is intended to be used as basic data in the field introduction process.

This study presents a methodology to determine the radionuclides of concern that are expected to be found during the final status survey of Kori Unit 1 decommissioning. The methodology involved reflecting the evaluation results of ORIGEN based on reference documents such as NUREG/CR-3474, NUREG/CR-4289, NUREG/CR-0130, WINCO-1191, and representative fuel loading. A list of potential radionuclides of concern was provided by reflecting the list of radionuclides of concern included in the Kori Unit 1 decommissioning plan. To select the radionuclides of concern, we analyzed the approach of US decommissioning plants based on the recommendations of NUREG-1757 Vol.2 Rev.1 and excluded certain radionuclides from the list. The final list of 23 radionuclides of concern was derived by excluding radionuclides that have a short half-life, low specific activity, analytically difficult to measure, inert gases, or naturally occurring radionuclides. This methodology can be applied to other nuclear power plants, such as the Wolsong Nuclear Power Plant, by reflecting the unique characteristics of the reactor.

In order to start decommissioning domestic nuclear facilities, the Final Decommissioning Plan (FDP) must be prepared and approved by the regulatory agency. The contents of domestic FDP consist of 12 chapters, and there is the decommissioning feasibility design that should be described in Chapter 5 as contents to be considered from the construction stage of nuclear facilities. The design of decommissioning feasibility for nuclear facilities seems to be largely divided into three items. In summary, there ae minimization of contaminations to facilities and the environment, easy of dismantling, and minimization of the radioactive waste generation. In addition, the design characteristics to which the ALARA principle is applied in terms of optimizing the exposure dose of workers and residents may also correspond to the decommissioning feasibility design. The design characteristics for decommissioning feasibility during the period leading up to the design, operation, and decommissioning of nuclear facilities can be listed as the main points as follows. Minimization of facility contamination will include contents related to the leakage of systems and components, minimization of effluents to the environment will involve gaseous and liquid effluents from systems and components to the environment, easy of dismantling will involves history and inspection records during operation, and minimization of radioactive waste generation can be the contents related to the radioactive waste management plans. The design characteristics of facilities and equipment to meet the ALARA principles can be listed as follows. It means taking into account the benefits and costs of the design improvement plan, and the elimination of unnecessary radiation exposure can be maintained at the exposure dose ALARA, which is in line with the decommissioning feasibility design. Among the requirements of licensing documents for decommissioning domestic nuclear facilities is the decommissioning feasibility design. This item relates to the design characteristics for decommissioning considered in the construction stage of the facility and should present the effectiveness of measures for them until operation and decommissioning. In this study, the regulatory requirements presented in the construction and operation stage and the contents presented in the U.S. case were reviewed, and it is hoped that it will be used as reference for the preparation of FDP.

Metals such as stainless steel and alloy 600 are used as structures and materials in nuclear power plants due to their excellent mechanical properties and heat resistance. And recently thermal and mechanical cutting technologies are being actively researched and developed for dismantling NPP. Among them, the mechanical cutting method has the advantage of less secondary waste generation such as fume and fine dust, but according to the wider the cutting range, the reaction force and the cutting device size are increased. In this paper, plasma assisted milling has been proposed to reduce the reaction force and device size, and the plasma efficiency was measured for SUS 316L. The plasma torch was operated at the level of 3 to 4 kW so that it was heated only without cutting. And the feedrate was set at 150 to 250 mm/min. The test confirmed that the plasma efficiency was 35% about SUS 316L, and it is expected that the numerical analysis using these test results can be used as basic data for plasma assisted milling.

In this study, in relation to the demolition of the building as a research reactor, in order to establish a basic design for preparation for relocation and installation of the TRIGA Mark-II, the present conditions such as actual measurements and structural safety were investigated, as well as technologies and cases related to the relocation and installation of cultural properties. Based on this, the basic design for the relocation and installation of cultural assets was established by reviewing the disassembly and transport design of the TRIGA Mark-II and the basic plan for the relocation site. Although the structural safety of the current self-weight of the structure is judged to be reasonable, when lifting the structure, it is necessary to consider a method of lifting the foundation by reinforcing the foundation so that the tensile force can be minimized in the structure. As for the technology to be applied before TRIGA Mark-II, the technology before non-transplacement was confirmed as the most reasonable method in terms of preserving the original form, securing safety, and securing economic feasibility. Among the non-replacement technologies, the methods that can be applied before reactor 1 can be largely classified into three types. The three methods to be reviewed can be largely classified into the traditional rail movement method, the movement method using transport equipment, and the crane movement method. Each required period was calculated from the basic design results, and the modular trailer method was judged to be the most efficient. From the basic design results, the required period for each stage according to the mobile construction method was calculated. Depending on the calculation result, the modular trailer method is judged to be the most efficient. However, the final construction method should be selected according to the detailed design results. Overall, the results obtained through this study suggest that it is possible to create a memorial hall without the previous installation of TRIGA Mark-II if the structure foundation is composed independently of the building foundation after conducting a detailed characteristic investigation on the foundation of the TRIGA Mark-II structure.

Radioactive waste generated in various forms needs to be technically stabilized for safe treatment, disposal, and long-term safety. Overseas, research on vitrification with excellent mechanical and physicochemical properties of high-level waste is being actively conducted. Vitrification is a process of converting radioactive waste into a stable and durable glass-like material, which is then safe storage and disposal. The stability of vitrified form is an important concern for environment safety, as any leakage or release of radionuclide could have serious consequences for health and the environment. Therefore, several studies are being conducted on the disposal stability of vitrification of radioactive waste. In order to evaluate the stability of the solidified form, mechanical properties such as density, microhardness, and compressive strength and chemical durability such as leaching properties should be performed. There are several types of leaching test methods to evaluate the chemical durability, which is important in the characterization of solidified forms. In this study, the leaching test method for chemical durability evaluation and the evaluation results of leaching characteristics according to pH were reviewed.

Seawater containing metals such as lithium and manganese is a “treasure trove” of infinite energy resources. Numerous domestic and foreign institutions are developing technologies to economically extract these resources from seawater. One method for extracting metal ions dissolved in seawater is the development of adsorbents with negative functional groups. Generally, adsorbents have adsorption performance that depends on factors such as seawater pH and temperature, but controlling the pH and temperature of seawater is practically impossible. On the other hand, thermal effluent discharged from power plants tends to be slightly higher in temperature than the surrounding environment. Therefore, this study investigates the potential for utilizing power plant effluent to extract dissolved resources in seawater. Results of investigations into several items related to the effluent from the Gori, Wolsong, Hanbit, and Hanul power plants are presented.

To develop technology for extracting energy resources from seawater, we first investigated the research experiences of domestic experts. The survey items included the types of adsorbents that can adsorb dissolved resources in seawater, the subjects of experiments, and the scope of research. We divided the types of adsorbents into organic and inorganic categories and compared their adsorption performance. We also examined how adsorption experiments were conducted using simulated solutions and confirmed whether there were any experiences of conducting experiments in actual seawater. A total of 14 domestic research papers on extracting dissolved resources from seawater were reviewed, excluding topics such as removing dissolved resources from seawater and seawater desalination. This review provides an understanding of domestic research trends and will be helpful in setting directions for future research and development.

Disposal of radioactive waste requires radiological characterization. Carbon-14 (C-14) is a volatile radionuclide with a long half-life, and it is one of the important radionuclides in a radioactive waste management. For the accurate liquid scintillation counter (LSC) analysis of a pure beta-emitting C-14, it should be separated from other beta emitters after extracted from the radioactive wastes since the LSC spectrum signals from C-14 overlaps with those from other beta-emitting nuclides in the extracted solutions. There have been three representative separation methods for the analysis of volatile C-14 such as acid digestion, wet oxidation, and pyrolysis. Each method has its own pros and cons. For example, the acid digestion method is easily accessible, but it involves the use of strong acids and generates large amount of secondary wastes. Moreover, it requires additional time-consuming purification steps and the skillful operators. In this study, more efficient and environment-friendly C-14 analysis method was suggested by adopting the photochemical reactions via in-situ decomposition using UV light source. As an initial step for the demonstration of the feasibility of the proposed method, instead of using radioactive C-14 standards, non-radioactive inorganic and organic standards were investigated to evaluate the recovery of carbon as a preliminary study. These standards were oxidized with chemical oxidants such as H2O2 or K2S2O8 under UV irradiations, and the generated CO2 was collected in Carbo-Sorb E solution. Recovery yield of carbon was measured based on the gravimetric method. As an advanced oxidation process, our photocatalytic oxidation will be promising as a time-saving method with less secondary wastes for the quantitative C-14 analysis in low-level radioactive wastes.

Kori Nuclear Power Plant Unit 1, which began operating in 1978, is Korea’s oldest commercial nuclear reactor. The reactor was permanently shut down in June 2017, and now the decommissioning process has begun. The decommissioning process will generate a significant amount of waste that requires appropriate management to minimize the impact on the environment and human health. And the waste routing, i.e. the activities and logistics for managing the material generated, is a key point in a decommissioning project. It determines the routes from the material inventory to the envisaged material end states. In this study, we review on several factors for the selection of the waste routes in a decommissioning project. In terms of sustainability, the ‘waste hierarchy’ should be applied to routing materials from nuclear facilities. According to the waste hierarchy, the preferred end state is reuse or recycling of the waste as material or, more preferably, the avoidance of waste generation. In addition, treatments (such as decontamination and thermal treatment) that can reduce the volumes requiring disposal as radioactive waste should be considered. Another important parameter is the need to secure availability and capacity of waste routes. Short-term bottlenecks or any delay in the removal of the waste from the site often has an impact on other site activities. If possible, at least two alternative waste routes should be identified for the main categories of waste and kept available throughout the decommissioning project. All routes should be direct to the material end state if possible, but it is more important that waste is removed from the site so that other site operations are not impeded.

As an initial part of Kori-1 & Wolsung-1 Unit decommissioning planning, a characterization plan is developed to define the nature, extent and location of contaminants, determine sampling locations and protocols, determine quality assurance objectives for characterization, and define documentation requirements. The actual characterization of a facility is an iterative process that involves initial sampling according to the characterization plan, field management (such as labeling, packaging, storing, and transport) of the samples, laboratory analysis, conformance to the data quality objectives (DQOs), and then identifying any additional sampling required, refining the DQOs, and modifying the characterization plan accordingly. The final product of the facility characterization is a document that describes the type, amount, and location of contaminants that will require consideration and removal during the decommissioning operations sufficient to prepare a decommissioning plan. In this study, implementing a characterization plan, developed in accordance with this standard, will result in obtaining or deriving the above information.

Kori unit 1, the first PWR (Pressurized Water Reactor) in Korea, was permanent shut down in 2017. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. Solid radioactive waste depending on the characteristics of the generation was classified into reactor vessel and reactor vessel internal, large components, small metals, spent nuclear fuel storage racks, insulation, wires, concrete debris, scattering concrete, asbestos, mixed waste, soil, spent resins and filters, and dry active waste. Radiological characterization of solid radioactive waste is performed to determine the characteristics of radioactive contamination, including the type and concentration of radionuclides. It is necessary to ensure the representativeness of the sample for the structures, systems and components to be evaluated and to apply appropriate evaluation methods and procedures according to the structure, material and type of contamination. Therefore, the radiological characterization is divided into concrete and structures, systems and components, and reactor vessel, reactor vessel internal and bioshield concrete. In this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. The results of this study can be used as a basis for the preparation of the FDP for the Kori unit 1.

Radioactive waste generated during decommissioning of nuclear power plants is classified according to the degree of radioactivity, of which concrete and soil are reclassified, some are discharged, and the rest is recycled. However, the management cost of large amounts of concrete and soil accounts for about 40% of the total waste management cost. In this study, a material that absorbs methyl iodine, a radioactive gas generated from nuclear power plants, was developed by materializing these concrete and soil, and performance evaluation was conducted. A ceramic filter was manufactured by forming and sintering mixed materials using waste concrete, waste soil, and by-products generated in steel mills, and TEDA was attached to the ceramic filter by 5wt% to 20wt% before adsorption performance test. During the deposition process, TEDA was vaporized at 95°C and attached to a ceramic filter, and the amount of TEDA deposition was analyzed using ICP-MS. The adsorption performance test device set experimental conditions based on ASTM-D3808. High purity nitrogen gas, nitrogen gas and methyl iodine mixed gas were used, the supply amount of methyl iodine was 1.75 ppm, the flow rate of gas was 12 m/min, and the supply of water was determined using the vapor pressure value of 30°C and the ideal gas equation to maintain 95%. Gas from the gas collector was sampled to analyze the removal efficiency of methyl iodine, and the amount of methyl iodine detected was measured using a methyl iodine detection tube.