Cement is widely used as representative industrial material. In Korea, about 50 million tons of cement are consumed every year. In the manufacture of cement, raw materials containing NORM such as fly ash and bauxite are used. Therefore, the workers can be subjected to radiation exposure. The major exposure pathway in NORM industries is internal exposure due to inhalation of aerosol. Internal radiation dose due to aerosol inhalation varies depending on physicochemical properties of the aerosol. Therefore, the objective of this study was to investigate aerosol properties influencing inhalation dose in cement industries. In this study, aerosol properties were measured for two cement manufacturers. A particulate size distribution and concentration at various processing areas in cement manufacturing industries in Korea were analyzed using a cascade impactor. The mass density of raw materials and byproducts were measured using pycnometer. Shape of particulates was analyzed using SEM. The radioactivity concentration of Ra-226, Ra-228 for U/Th decay series was measured using HPGe. Particulate concentration by size was distributed log-normally with maximum at particle size about 7.2 μm in manufacturer A and 5.2 μm in manufacturer B. The mass density of fly ash and cement were 2.3±0.06, 3.2±0.02 g/cm3 respectively in manufacturer A. In manufacturer B, the mass density of bauxite and cement were 3.4±0.02, 2.9±0.01 g/cm3 respectively. The shape of particulates appeared as spherical shape in manufacturer A and B regardless of sampling area. Thus, a shape factor of unity could be assumed. The radioactivity concentrations of Ra-226, Ra-228 were 82±9, 82±8 Bq/kg for fly ash, and 25±4, 23±3 Bq/kg for cement in manufacturer A. In manufacturer B, the radioactivity concentrations of Ra-226, Ra-228 were 344±34, 391±32 Bq/kg for bauxite, and 122±13, 145±12 Bq/kg for cement. The radioactivity concentrations of Ra-226, Ra-228 in cement were less than raw materials such as fly ash and bauxite. It is because the dilution of the radioactivity concentration occurred during mixing with other raw materials in cement production process. This study results will be used as database for accurate dose assessment due to airborne particulate inhalation by workers in cement industries.

In general, after the decommissioning of nuclear facilities, buildings on the site can be demolished or reused. The NSSC (Nuclear Safety and Security Commission) Notice No. 2021-11 suggests that when reusing the building on the decommissioning site, a safety assessment should be performed to confirm the effect of residual radioactivity. However, in Korea, there are currently no decommissioning experiences of nuclear power plants, and the experiences of building reuse safety assessment are also insufficient. Therefore, in this study, we analyzed the foreign cases of building reuse safety assessment after decommissioning of nuclear facilities. In this study, we investigated the Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. For each case, the source term, exposure scenario, exposure pathway, input parameter, and building DCGLs were analyzed. In the case of source term, each facility selected 9~26 radionuclides according to the characteristics of facilities. In the case of exposure scenario, building occupancy scenario which individuals occupy in reusing buildings was selected for all cases. Additionally, Rancho Seco also selected building renovation scenario for maintenance of building. All facilities selected 5 exposure pathways, 1) external exposure directly from a source, 2) external exposure by air submersion, 3) external exposure by deposited on the floor and wall, 4) internal exposure by inhalation, and 5) internal exposure by inadvertent ingestion. For the assessment, we used RESRAD-BUILD code for deriving building DCGLs. Input parameters are classified into building parameter, receptor parameter, and source parameter. Building parameter includes compartment height and area, receptor parameter includes indoor occupancy fraction, ingestion rate, and inhalation rate, and source parameter includes source thickness and density. The input parameters were differently selected according to the characteristics of each nuclear facility. Finally, they derived building DCGLs based on the selected source term, exposure scenario, exposure pathway, and input parameters. As a result, it was found that the maximum DCGL was 1.40×108 dpm/100 cm2, 1.30×107 dpm/100 cm2, and 1.41×109 dpm/100 cm2 for Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility, respectively. In this study, we investigated the case of building reuse safety assessment after decommissioning of the Yankee Rowe nuclear power Plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. Source terms, exposure scenarios, exposure pathways, input parameters, and building DCGLs were analyzed, and they were found to be different depending on the characteristics of the building. This study is expected to be used in the future building reuse safety assessment after decommissioning of domestic nuclear power plants. This work was

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.

With the aging of nuclear power plants (NPPs) in 37 countries around the world, 207 out of 437 NPPs have been permanently shutdown as of August 2022 according to the IAEA. In Korea, the decommissioning of NPPs is emerging as a challenge due to the permanent shutdown of Kori Unit 1 and Wolsong Unit 1. However, there are no cases of decommissioning activities for Heavy Water Reactor (HWR) such as Wolsong Unit 1 although most of the decommissioning technologies for Light Water Reactor (LWR) such as Kori Unit 1 have been developed and there are cases of overseas decommissioning activities. This study shows the development of a decommissioning waste amount/cost/process linkage program for decommissioning Pressurized Heavy Water Reactor (PHWR), i.e. CANDU NPPs. The proposed program is an integrated management program that can derive optimal processes from an economic and safety perspective when decommissioning PHWR based on 3D modeling of the structures and digital mock-up system that links the characteristic data of PHWR, equipment and construction methods. This program can be used to simulate the nuclear decommissioning activities in a virtual space in three dimensions, and to evaluate the decommissioning operation characteristics, waste amount, cost, and exposure dose to worker. In order to verify the results, our methods for calculating optimal decommissioning quantity, which are closely related to radiological impact on workers and cost reduction during decommissioning, were compared with the methods of the foreign specialized institution (NAGRA). The optimal decommissioning quantity can be calculated by classifying the radioactivity level through MCNP modeling of waste, investigating domestic disposal containers, and selecting cutting sizes, so that costs can be reduced according to the final disposal waste reduction. As the target waste to be decommissioning for comparative study with NAGRA, the calandria in PHWR was modeled using MCNP. For packaging waste container, NAGRA selected three (P2A, P3, MOSAIK), and we selected two (P2A, P3) and compared them. It is intended to develop an integrated management program to derive the optimal process for decommissioning PHWR by linking the optimal decommissioning quantity calculation methodology with the detailed studies on exposure dose to worker, decommissioning order, difficulty of work, and cost evaluation. As a result, it is considered that it can be used not only for PHWR but also for other types of NPPs decommissioning in the future to derive optimal results such as worker safety and cost reduction.

The fuel fabrication facility has been built and is being operated by KAERI since licensing research reactor fuel fabrication in 2004. After almost 20 years of operation, outdated equipment for fabrication or inspection has been replaced by automated, digitalized ones to assure a higher quality of nuclear fuels. However, the generation of a large amount of radioactive waste is another concern for the replacement in terms of its volume and various types of it that should be categorized before disposal. The regulatory body, NSSC (Nuclear Safety and Security Commission) released a notice related to the classification of radioactive wastes, and most accessory equipment can be classified into the clearance levels, called self-disposal waste. In this study, the practice of self-disposal of metal radioactive waste is carried out to reduce its volume and downgrade its radioactivity. For metal radioactive waste, which is expected to occupy the most amount, analysis status and legal limitations were performed as follows: First, the disposal plan was established after an investigation of the use history for equipment. Second, those were classified by types of materials, and their surface radio-contamination was measured for checking self-disposable or not. After collecting data, the plan for the self-disposal was written and submitted to the Korea Institute of Nuclear Safety (KINS) for approval.

The reliable information on the hydraulic characteristics of rock mass is one of the key site factors for design and construction of deep subsurface structures such as geological radioactive nuclear waste disposal repository, underground energy storage facility, underground research laboratory, etc. In order to avoid relying on foreign field test technology in future projects, we have independently designed and made integrated type main frame, 120 bar waterproof downhole sonde, and 1,200 m wireline cable winch through a series of R&D activities. They are core apparatuses of the Deep borehole Hydraulic Test System (DHTS). Integration of individual test equipment into a single main frame allows safe and efficient work in the harsh field condition. The DHTS was developed aiming primarily for constant pressure (head) injection test and pulse test in deep impermeable rock mass. The maximum testing depth of the DHTS is about 1,050 m from the surface. Using this system, it is possible to make precise stepwise control of downhole net injection pressure in less than 2.0 kgf/cm2 with dual hydraulic volume controller and also to inject and measure the very low flow rate below 0.01 l/min with micro flow rate injection/control module. Over the past two years, we have successfully completed more than 50 in situ hydraulic tests at 5 deep boreholes located in the Mesozoic granite and sedimentary rock regions in Korea. Among them, the deepest testing depth was more than 920 m. In this paper, the major characteristics of the DHTS are introduced and also some results obtained from the high precision field tests in the deep and low permeable rock mass environment are briefly discussed.