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.

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.

The Korea government decided to shut down Kori-1 and Wolsung-1 nuclear power plants (NPPs) in 2017 and 2019, respectively, and their decommissioning plans are underway. Decommissioning of a NPP generates various types of radioactive wastes such as concrete, metal, liquid, plastic, paper, and clothe. Among the various radioactive wastes, we focused on radioactive-combustible waste due to its large amount (10,000–40,000 drums/NPP) and environmental issues. Incineration has been the traditional way to minimize volume of combustible waste, however, it is no longer available for this amount of waste. Accordingly, an alternative technique is required which can accomplish both high volume reduction and low emission of carbon dioxide. Recently, KAERI proposed a new decontamination process for volume reduction of radioactivecombustible waste generated during operation and decommissioning of NPPs. This thermochemical process operates via serial steps of carbonization-chlorination-solidification. The key function of the thermochemical decontamination process is to selectively recover and solidify radioactive metals so that radioactivity of the decontaminated carbon meets the release criteria. In this work, a preliminary version of mass flow diagram of the thermochemical decontamination process was established for representative wastes. Mass balance of each step was calculated based on physical and chemical properties of each constituent atoms. The mass flow diagram provides a platform to organize experimental results leading to key information of the process such as the final decontamination factor and radioactivity of each product.

The decommissioning of nuclear-related facilities at the end of their design life generates various types of radioactive waste. Therefore, the research on appropriate disposal methods according to the form of radioactive waste is needed. This study is about the solidification of uranium contaminated soils that may occur on the site of nuclear facilities. A large amount of radioactively contaminated soil waste was generated during the decommissioning of the uranium conversion plant in KAERI, and research on the proper disposal of this waste has been actively conducted. Numerous minerals in the soil can become glass-ceramic through the phase change of minerals during the sintering process. This method is effective in reducing the volume of waste and the glassceramic waste form has excellent mechanical strength and leaching resistance. In this study, the optimum temperature and time conditions were established for the production of glass-ceramic sintered body of soil. The compressive strength and leachability of the sintered body made by applying the optimal conditions to simulated waste was confirmed. The basic physicochemical properties of simulated soil waste were identified by measuring the pH, moisture content, density, and organic matter content. The elemental compositions in the soil was confirmed by XRF. Soils were classified by particle size, and each sample was compressed with a pressure of 150 MPa or more to prepare a green body. Based on the TG-DSC analysis, an appropriate heating temperature was set (>1,000°C), and the green body was maintained in a muffle furnace for 2~6 hours. The optimal sintering conditions were selected by measuring the compressive strength and volume reduction efficiency of the sintered body for each condition. The difference between the green body and sintered body was observed by XRD and SEM. In the experiments for evaluation of additives, the selected chemical substances were mixed with the soil sample in a rotator. Based on the results of TG-DSC, sintered body was made at 850°C, and the compressive strength and volume reduction were compared. Based on the results, the most effective additive was determined, and the appropriate ratio of the additive was found by adjusting the range of 1~5 wt%. This study was confirmed that the sintered soil waste showed sufficient stability to meet the disposal criteria and effective volume reduction for final disposal.

Especially for near-surface repository for disposal of the low- and intermediate-level radioactive waste, safety assessment in case of inadvertent human intrusion should be handled seriously. This is because this type of incident will possibly give rise to high acute, not chronic exposure dose even though its occurrence of likelihood could higher than rather deeper geological repository for disposal of high-level radioactive waste over long time span after closure of the repository. Recently well drilling scenario for the pumping groundwater from the aquifer near the repository, among other possible inadvertent human intrusion incidents, has been popularly evaluated for the worst case due to its relatively high possibility of occurrence in parallel with normal scenarios for the nuclide transport for post-closure safety assessment of the repository. Movement of nuclide plume both in the confined and unconfined aquifer under and over a radioactive waste repository is of importance especially around an extracting well. Through this study a simple comment regarding quantification between a pumping rate from the well drilled into the aquifer as well as quantification of the plume size flowing around the well is presented. Drawdown of the well which is the change of water level of the upper water surface of the aquifer due to well pumping makes a cone of depression. And capture zone in the aquifer which is formed around the well, by which the groundwater is removed out, is the groundwater volume or area in the aquifer that is considered to contribute the extraction of the well by pumping. Usually this capture zone does not encompass the entire aquifer thickness for the partially penetrating well, which means that not all the portion of flowing groundwater through the aquifer is drawn by the well. And this capture zone does not need to coincide with the volume of the cone. Furthermore, all the nuclide plume volume is not necessarily and completely mixed with the groundwater flowing the entire aquifer. Therefore, a strategical approach might be required to grasp the aquifer portion and the plume size influenced by pumping to evaluate rather accurate radiological consequences due to the well scenario avoiding overestimation and meaningless conservatism as well, which is especially very common in the mass balance modeling e.g., by GoldSim under assumption that all the groundwater volume from the aquifer near the well extracted by the well. Although the capture zone around the well should be determined both by use of global/local groundwater flow model in the aquifer but a simple analytical model could be sought. Capture zone analysis has been widely seen in the area of the design of groundwater remediation system. If for safety assessment of the subsurface repository the plume behavior in the aquifer under the repository should be well characterized and correctly modeled, then the current study is expected to be more or less helpful to develop a specific mass balance model for nuclide transport and groundwater flow for assessment of an abnormal well drilling scenario near the repository.

The Deep Borehole Disposal (DBD) method has various advantages, such as minimizing the use of site area and corrosion of the disposal container and improving long-term structural safety. However, it is necessary to review the problems that may occur in various technologies related to the emplacement and retrieval of the disposal container and the sealing of the borehole. Therefore, the purpose of this study is to evaluate the structural integrity of an emplacement and retrieval device (hereinafter, the disposal container connecting device) of a DBD container. The disposal connecting device was evaluated according to ANSI 14.6 and NUREG-0612 standards. The allowable stress should be less than the yield strength under the load condition of 3g. The length of the disposal container connecting device was about 2,900 mm, the diameter was 406 mm, and the weight was about 1.2 tons. In addition, 10 disposal containers weighing up to 2.2 tons were handled. The disposal container connecting device was made of stainless steel, and the maximum operating temperature was about 300°C. For structural evaluation, ABAQUS finite element analysis program was used. The analysis model was modeled only 1/2 part considering symmetry condition. The analysis model was modeled using 410,431 nodes and 344,119 solid elements. Three times load was applied to the weight of the disposal container. Axisymmetric conditions were applied to the symmetrical surface of the disposal container, and vertical restraints were applied to the upper lifting lugs. A surface-to-surface contact condition was applied to the part where the contact occurred. As a result of the analysis, the greatest stress was generated at the part supported by the clamp at the disposal container connector at 168.9 MPa. In the lugs and pins connecting the guide and the connecting device, a stress of 530.1 MPa was generated by shearing. In the bolts of the disposal container connecting device, a stress of 498MPa was generated and the safety margin was 1.73. A stress of 486.1 MPa was generated in the disposal container connecting device, and the safety margin was the smallest 1.16. As a result of the analysis, all components of the disposal container connecting device showed a safety margin of 1.16 or more at the maximum operating temperature and satisfied the allowable stress.

Copper is used for deep geological disposal canisters of spent nuclear fuels, because of excellent corrosion resistance in an oxygen-free environment. However, sulfide formation during the long-term exposure under deep geological disposal condition can be harmful for the integrity of copper canisters. Sulfur around the canisters can diffuse along grain boundaries of copper, causing grain boundary embrittlement by the formation of copper sulfides at the grain boundaries. The development of copper alloys preventing the formation of copper sulfides along grain boundaries is essential for the longterm safety of copper canisters. In this research, the mechanisms of copper sulfide formation at the grain boundary are identified, and possible alloying elements to prevent the copper sulfide formation are searched through the first principle calculations of solute atom-vacancy binding energy and the molecular dynamics calculation of grain boundary segregation energy. The comparison with the experimental literature results on the mitigation of copper embrittlement confirmed that the theoretically identified mechanisms of copper sulfide formation and the selected alloy elements are valid. Thereafter, binary copper alloys were prepared by using a vacuum arc melting furnace. Sulfur was added during casting of the copper alloys to induce the sulfide formation. The cast alloys were cold-rolled into a plate after homogenization heat treatment. The microstructure and mechanical property of each alloy were investigated after recrystallization in a vacuum tube heat treatment furnace. The copper alloys developed in this study are expected to contribute in increasing the long-term safety of deep geological disposal copper canisters by reducing the embrittlement caused by the sulfide formation.

Despite the increasing interest in Deep Borehole Disposal (DBD) for its capability of minimizing disposal area, detailed research about DBD operation system design should be conducted before the DBD can be implemented. Recently, DBD operation system applying wireline emplacement (WE) technique is under study due to its high flexibility and capability of minimizing surface equipment. In this study, a conceptual WE system, and operation procdure is introduced. The conceptual WE system consists of 3 main stations, which from the top are hoisting station (HS), canister connection station (CCS) and basement (BS). In HS, WE is controlled and monitored. The WE is controlled using wireline drum winch and sheaves, and load on wireline is measured using a load cell. HS also has a pressure control system (PCS), which monitors internal pressure of the system, and a lubricator, which act as housing for joint device, allowing the joint device to be easily inserted into the borehole. The joint device is used to connect the disposal canister to wireline for emplacement/retrieval. In CCS, a rail transporter brings a transport cask containing disposal canisters, then the transport cask is connected to the hoisting system and a PCS in the BS. The main component located at canister station are a sliding shielding door (SSD), and a slip. The SSD is used to prevent canister from falling into borehole during the connecting operation and prevent radiation from BS to affect the workers. The slip is located beneath the SSD and is used to hold the disposal canister before it is lowered into the borehole. In BS, PCS is installed to prevent overflow and blowout of borehole fluid. The PCS consists of wireline pressure valve, christmas tree and BOP, which all are a type of pressure valve to seal the borehole and release pressure inside the borehole. The WE procedure starts with transporting transport cask to CCS. The transport cask is connected to lubricator, and PCS. Joint device is lowered down to be connected with disposal canisters, then pulled up to check the load on the wireline. After the check-up, SSD is opened, and disposal canister is lowered into the borehole. When desired depth is reached, joint device is disconnected and retrieved for next emplacement. In this study, the conceptual deep borehole disposal system design implementing WE technique is introduced. Based on this study, further detailed design could be derived in future, and feasibility could be tested.

Currently, the most widely accepted disposal concept for long-term isolation of high level radioactive waste including spent nuclear fuels is to disposal in a deep geological repository designed and constructed with multiple barriers composed of engineered and natural barriers so that the waste can be completely isolated in a stable deep geological environment. In this concept, an important consideration is the heat generated from the waste due to the large amount of fission products present in the high level waste loaded in the disposal container. For safe and complete isolation of high level radioactive waste in the deep geology, the disposal concepts that meet the thermal requirements for the disposal system design have been developed by harmonizing the thermal characteristics of engineered and natural barriers in Korea. In this paper, the deposition hole configuration and the decay heat dissipation area (surface area) of disposal container were considered for the efficient thermal management in the deep geological disposal concept. Heat transfer through the waste form, its container and surrounding components and the rock will be mainly by conduction. Heat transfer by radiation and convection can be negligible after backfilling. When considering heat conduction, according to Fourier’s law, if the thermal conductivity of the repository components is the same, the greater the heat dissipation area and the adjacent temperature gradient, the greater the conduction effect. Therefore, rather than the conventional concept of loading 4 PWR spent fuel assemblies per disposal container and placing one disposal container in a deposition hole, it is better to load one assembly per disposal container and place 4 disposal containers in a deposition hole. In this case, it was found that the disposal area could be reduced through efficient thermal management. Considering this thermal management method as an alternative to the concept of deep geological disposal, additional research is needed.

It is expected that around 576,000 bundles of CANDU spent nuclear fuels (SNF) will be generated from the four CANDU reactors located at the Wolsong site, according to the 2nd National Plan for the management of High-Level radioactive Waste (HLW). The CANDU SNFs are currently stored at the dry storage facilities at the Wolsong site. The authors proposed KRS+ geological disposal system consisting of two different concepts, Swedish KBS-3V type and Canadian NWMO type, for the final management of CANDU SNF. Both the concepts were designed based on the geological data obtained from the KURT (KAERI Underground Research Tunnel). The NWMO type is an in-room horizontal placement method. In this study, we try to determine the reference concept among the two proposed concepts at 500 meters below the ground surface. Assuming 10,000 tU of CANDU SNF and the KURT site, we design two engineered barrier systems, that is disposal canisters and buffers. The copper disposal canister is designed with a copper thickness of 10 mm based on a cold spray coating technique for both the disposal concepts. The domestic Ca-bentonite is used for the compact bentonite buffer with dry density of 1.6 g/cm3. Two concepts are compared in terms of safety, economics of the engineered barriers, and environment-friendliness. Because the same amounts of CANDU SNF are disposed of at the same depth, the differences in the disposal area are neglected. For the comparison in terms of safety, the corrosion lifetimes of the disposal canisters of two disposal systems are quantitatively calculated, and the capacities for retarding radionuclide releases of the compacted bentonite buffers are assessed. A computer tool developed by the authors is used in order to assess the lifetime of a disposal canister. In this study, the case that corrosion of a copper canister by sulfide from groundwater through intact buffer is analyzed. The sulfide concentration in groundwater is assumed to be 3 ppm. The most important safety function of buffer is to retard the radionuclide release. Twelve long-lived radionuclides are selected to compare the capacities for retarding the radionuclide transport through the buffer using an analytical solution. The retention time by an engineered barrier consisting of a disposal canister and a buffer is compared with twenty times the half-life of each radionuclide for both the disposal systems. The selected reference concept will be compared with the alternative geological concepts through a further study.