In the decommissioning site of Korean Research Reactor 1&2 (KRR-1&2), according to Low and Intermediate-level Radioactive Waste Disposal Acceptance Criteria of the Korea Radioactive Waste Agency (WAC-SIL-2022-1), characteristics of radioactive waste was conducted on approximately 550 drums of concrete and soil waste for a year starting from 2021. Among them, 50 drums of concrete waste transported and disposed to Gyeongju LILW disposal facility at the end of 2022. For the remaining approximately 500 drums of concrete and soil waste stored on-site, they were reclassified into two categories: permanent disposal grade and clearance grade. This classification was based on calculating the sum of fractions (SOF) per drum for each radionuclides. The plan is to dispose of around 200 drums in the permanent disposal grade and about 300 drums in the clearance grade by the end of 2023. Since concrete and soil decommissioning wastes are generated in large quantities over a short period with similar origins, they were grouped within five drums as suggested by the acceptance criteria. Mixed samples were collected from each group and used for radionuclide analysis. When utilizing mixed samples, three distinct samples are collected and analyzed for each group. The maximum value among these three radionuclide analysis results is then uniformly applied as the radionuclide concentration value for all drums within that group. Radioactive nuclides contained in similar types of radioactive waste with similar origins can be expected to have some statistical distribution. However, There has been no verification as to whether the maximum value among the three mixed samples exists within the statistical distribution or if it deviates from this distribution to represent a different value. In this study, we confirmed characteristics of radionuclide concentration distribution by examining and comparing radionuclide concentration distributions for radioactive wastes drum grouped for nuclear characteristic among 50 concrete wastes drum disposed in year 2022 and 500 concretes & soils drum scheduled for disposal (clearance or permanent disposal) in year 2023. In particular, when comparing tritium to other nuclides, it was observed that the standard deviation for the distribution of maximum values was approximately 318 times larger.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR 1&2), the first research reactors in South Korea, began in 1997 and the decommissioning status is currently proceeding with phase 3. It is expected that more than 5,000 tons of dismantled wastes will be generated as the contaminated building is demolished. Since these dismantled wastes must be disposed of in an efficient method considering economic feasibility, it is desirable to clearance extremely low-level wastes whose contamination is so minimal that the radiological risk is negligible. In Korea, in order to approve the clearance of radioactive waste, it must be proven that the nuclide concentration standards are met or that the dose to individuals and collectives is below the allowable dose value. At the KRR 1&2 decommissioning site, dismantled wastes have been steadily being disposed of through clearance procedure since 2021. Clearance was approved by the Korean Institute of Nuclear Safety (KINS) for one case of concrete waste in 2021 and two cases of metal waste in 2022. In 2023, the clearance of metal waste and asbestos waste has been approved so far, and in particular, this is the first case in Korea for asbestos waste. In this study, we compared the dose assessment methods and results of clearance wastes at the KRR 1&2 decommissioning site from 2021 to present. Dose assessment was conducted by applying the landfill scenario for concrete and asbestos and the recycling scenario for metal waste. The calculation codes used were RESRAD-onsite 7.2 and RESRAD-recycle 3.10. The dose conversion factors (DCF) for each age group (infant, 1y, 5y, 10y, 15y, adult) of the target nuclide used the values presented in ICRP-72, and in particular, geo-hydrological data of the actual landfill site was used as an input factor when evaluating landfill scenarios. As a result of the dose assessment, when landfilling concrete wastes in 2020, the personal dose and collective dose were evaluated the most at 2.80E+00 μSv/y and 4.83E-02 man·Sv/y, respectively.

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.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR-1&2), the first research reactors in South Korea, began in 1997. Approximately 5,000 tons of waste will be generated when the contaminated buildings are demolished. Various types of radioactive waste are generated in large quantities during the operation and decommissioning of nuclear facilities, and in order to dispose of them in a disposal facility, it is necessary to physico-chemically characterize the radioactive waste. The need to transparently and clearly conduct and manage radioactive waste characterization methods and results in accordance with relevant laws, regulations, acceptance standards is emerging. For radioactive waste characterization information, all information must be provided to the disposal facility by measuring and testing the physical, chemical, and radiological characteristics and inputting related documents. At this time, field workers have the inconvenience of performing computerized work after manually inputting radioactive waste characterization information, and there is always a possibility that human errors may occur during manual input. Furthermore, when disposing of radioactive waste, the production of the documents necessary for disposal is also done manually, resulting in the aforementioned human error and very low production efficiency of numerous documents. In addition, as quality control is applied to the entire process from generation to treatment and disposal of radioactive waste, it is necessary to physically protect data and investigate data quality in order to manage the history information of radioactive waste produced in computerized work. In this study, we develop a system that can directly compute the radioactive waste characterization information at the field site where the test and measurement are performed, protect the stored radioactive waste characterization data, and provide a system that can secure reliability.

During the operation or decommission of nuclear facilities, a large amount of dry active waste and cable waste with various shape and material is generated. Most of these wastes have almost no radioactive contamination and can be disposed of by incineration, landfill, recycling, etc. under clearance regulation. For clearance of radioactive waste, it is necessary to verify the characteristics of radiological contamination and prove that it meets the criteria for clearance regulation. According to the domestic clearance regulation, if it is difficult to measure radioactivity of wastes due to their surface condition using direct or indirect measurement methods, representative samples should be collected and analyzed for radioactivity. When sampling, it is desirable to collect samples of about 1 kg that can represent waste contamination per 200 kg or per 1 m2, and the homogeneity of the samples also should be demonstrated. However, in the case of dry active wastes, it is very difficult to prove the homogeneity of the samples because of surface shapes and conditions of the wastes. In particular, considering cable waste generated during the decommission, it is hardly capable to prove the representativeness of the sample, even though the inner shell of the covering material and the copper wire are almost uncontaminated. In this study, we show the development of a treatment system that makes it easy to prove the representativeness of samples when disposing of dry active waste or cable waste generated in nuclear facilities. The treatment device is designed in such a way that it has different storage unit and cutting unit suitable for the material characteristics of each waste type (soft, hard and cable), and therefore optimizes the efficiency of the shredding or cutting process. In addition, it is expected that the work efficiency in the radioactive treatment site with a narrow area can also be improved by providing a moving part on the device.