Recently, BNS (Best System) developed a system for evaluation and classification of soil and concrete wastes generated from nuclear power plant decommissioning. It is composed of various modules for container loading, weight measurement, contamination evaluation, waste classification, stacking, storage and control. The contamination evaluation module of the system has two sub modules. One is for quick measurement with NaI (Tl) detector and the other is for accurate measurement with HPGe detector. The container used at the system for wastes handling has capacity of 100 kg and made of stainless steel. According to the measurement result of Co-60 and Cs-137, the waste is classified as waste for disposal or waste for clearance. Performance of the system was demonstrated using RM (Reference Material) radiation source. This year, necessity of system improvement was suggested due to revised operation requirements. So, the system should show throughput of more than 1 ton/hr and Minimum Detectable Activity (MDA) of less than 0.01 Bq/g (1/10 of criteria for regulatory clearance) for Co-60 and Cs-137. And soil waste become main target of the system. For this, the container used for soil waste handling should have capacity of 200 kg. As a result, material for the container need to be changed from stainless steel to plastic or FRP (Fiber Reinforced Plastics). And large area detector should be introduced to the system to enhance processing speed of the system. Additionally, container storage rack and conveyor system should be modified to handle 200 kg capacity container. Finally, moving path of the container will be redesigned for enhanced throughput of the system. In this paper, concept development of the system was suggested and based on that, system development will be followed.

Recently, the nuclear decommissioning and environmental restoration industries has significantly attracted as a new industry field due to the decision to decommission the KORI#1 and WOLSONG #1 nuclear power plant. In order to dispose of the decommissioning radioactive wastes generated during nuclear decommissioning, proper analysis is required, and disposal decisions are determined based on the analysis results. When dismantling a nuclear power plant, a few thousand of tons decommissioning waste are produced, so these require analysis for proper disposal. Therefore, a radionuclide facility for decommissioning waste analysis is essential for the disposal of the large quantities of decommissioning waste generated during nuclear power plant decommissioning. Korea Research Institute of Decommissioning (KRID) was established radionuclide analysis facilities to address above issues and support nuclear power plant decommissioning projects. The plan is to perform classification by type and radionuclide for all waste produced during nuclear power plant decommissioning and to support the disposal of radioactive wastes. In addition, we plan to establish validation methods for samples where verification methods are not established, in order to conduct efficient analysis and management. In this presentation, we will introduce the radionuclide facility currently under construction at KRID and present the space design, equipment layout, and utilization plans.

Support for nuclear power plant (NPP) dismantling & decommissioing (D&D) industry is necessary through development of the infrastructure and the D&D technology. Because KORI#1 and Wolsong#1 is planned to decommission until around 2030. Korea research institute of decommissioing (KRID) was established through the preliminary feasibility study. KRID has plan to support nuclear companies to join D&D industry. Normal facilities (Lv.1) of KRID infracstucture are currently being constructed and radiation management facilities (Lv.2) construction is expected to begin in October. Further, KRID is planning the construction of equipment to develop the procedure for radionuclide analysis through R&D project. A total period of the R&D project is 45 months, and the total R&D funding for this period is 19.4 billion won. The ultimate goal of the R&D project is to build the infractstucture base to analyze decommissioning radioactive wastes. Furthermore, the R&D project is important to reliably perform the NPP D&D.

During decommissioning and site remediation of nuclear power plant, large amount of wastes (including radioactive waste) with various type will be generated within very short time. Among those wastes, soil and concrete wastes is known to account for more than 70% of total waste generated. So, efficient management of these wastes is very essential for effective NPP decommissioning. Recently, BNS (Best System) developed a system for evaluation and classification of soil and concrete wastes from the generation. The system is composed of various modules for container loading, weight measurement, contamination evaluation, waste classification, stacking, storage and control. By adopting modular type, the system is good for dealing with variable situation where system capacity needs to be expanded or contracted depending on the decommissioning schedule, good for minimizing secondary waste generated during maintenance of failed part and also good for disassemble, transfer and assemble. The contamination evaluation module of the system has two sub module. One is for quick measurement with NaI(Tl) detector and the other is for accurate measurement with HPGe detector. For waste transfer, the system adopts LTS (Linear Transfer System) conveyor system showing low vibration and noise during operation. This will be helpful for minimizing scattering of dust from the waste container. And for real time positioning of waste container, wireless tag was adopted. The tag also used for information management of waste history from the generation. Once a container with about 100 kg of soil or concrete is loaded, it is moved to the weight measurement module and then it transfers to quick measurement module. When measured value for radioactivity concentration of Co- 60 and Cs-137 is more than 1.0 Bq/g, then the container is classified as waste for disposal and directly transferred to stacking and storage rack. Otherwise, the container is transferred to accurate measurement module. At the accurate module, the container is classified as waste for disposal or waste for regulatory clearance depending on the measurement result of 0.1 Bq/g. As the storage rack has a sections for disposal and regulatory clearance respectively, the classified containers will be positioned at one of the sections depending on the results from the contamination evaluation module. The system can control the movement of lots of container at the same time. So, the system will be helpful for the effective nuclear power plant decommissioning in view of time and budget.

Transport packages have been developed to transport the decommissioning waste from the nuclear power plant. The packages are classified with Type IP-2 package. The IAEA requirements for Type IP-2 packages include that a free drop test should be performed for normal conditions of transport. In this study, drop tests of the packages were performed to prove the structural integrity and to verify the reliability of the analysis results by comparing the test and analysis results. Half-scale models were used for the drop tests and drop position was considered as 0.3 m oblique drop on packages weighing more than 15 tons. The strain and impact acceleration data were obtained to verify the reliability of the analysis results. Before and after the drop tests, radiation shielding tests were performed to confirm that the dose rate increase was within 20% at the external surface of the package. Also, measurement of bolt torque, and visual inspection were performed to confirm the loss or dispersion of the radioactive contents. After each drop test, slight deformations occurred in some packages. However, there was no loss of pretension in the lid bolts and the shielding thickness was not reduced for metal shields. In the package with concrete shield, the surface dose rate did not increase and there was no cracks or damage to the concrete. Therefore, the transport packages met the legal requirements (no more than a 20% increase of radiation level and no loss or dispersion of radioactive contents). Safety verifications were performed using the measured strain and acceleration data from the test, and the appropriate conservatism for the analysis results and the validity of the analysis model were confirmed. Therefore, it was found that the structural integrity of the packages was maintained under the drop test conditions. The results of this study were used as design data of the transport packages, and the packages will be used in the NPP decommissioning project in the future.

Decommissioning of a nuclear power plant (NPP) generate large amounts of various types of wastes. In accordance with the Nuclear Safety and Security Commission Notice of Korea (No. 2020- 6), they are classified as High Level Waste (HLW), Intermediate Level Waste (ILW), Low Level Waste (LLW), Very Low Level Waste (VLLW) and Exempt Waste (EW) according to specific activities. More than 90% of the wastes are at exempt level, mostly metal and concrete wastes with low radioactivity, of which the concentrations of nuclides is less than the allowable concentration of self-disposal. The self-disposal or recycling of these wastes is widely used worldwide. More than 10,000 drums, based on 200 L drum, are expected to be produced in the decommissioning process of a unit of nuclear power plant. Due to the limited storage capacity of the intermediate & low level waste disposal facility in Gyeongju, recycling and self-disposal of EW are actively recommended in Korea. A variety of scenarios were proposed for recycling and self-disposal of decommissioning metal/ concrete wastes, and a computational program called REDISA was developed to perform the dose evaluation for each recycling and self-disposal scenario. The REDISA computer program can calculate external and internal exposure doses by simulating the exposure pathways from waste generation, thru transport, processing, manufacture, to the final destination of recycling or self-disposal. In this study, the self-disposal scenario was only considered for the dose evaluation. Many studies have been conducted to evaluate the exposure doses of the radioactive waste disposal sites. However, there have been few researches on dose evaluation for self-disposal landfills. In particular, the dose evaluation is important not only during the operation period, but also for a long period after the facility is closed. To this end, we developed a conceptual model for dose evaluation for post-closure scenarios of the self-disposal landfill of decommissioning metal/concrete wastes with reference to the methodology of IAEA-TECDOC-1380. The model incorporates three exposure pathways, including external exposure from contaminated soil, internal exposure by inhalation, and internal exposure by ingestion of water and food grown in contaminated soil. The duration of the dose evaluation is set to 100,000 years after the closure of landfill facility. Co-60 was selected as dominant nuclide, and dose evaluation was performed based on unit specific activity of 1 Bq/g. Exposure doses shall be verified for their application in accordance with the annual dose limit of 10 Sv/yr for self-disposal. As a result, the post-closure scenario of selfdisposal landfills have shown negligible effects on public health, which means that the exposures doses from transportation and operational processes should be considered more carefully for selfdisposal of decommissioning metal/concrete wastes.

Waste containers for packaging, transportation and disposal of NPP (Nuclear Power Plant) decommissioning wastes are being developed. In this study, drop tests were conducted to prove the safety of containers for packaging of the wastes and to verify the reliability of the analysis results by comparing the test and analysis results. The drop height of the waste containers was considered to be 30 mm, which is the maximum lifting speed of a 50 tons crane in the waste treatment facility converted to the drop height. Drop orientation of the containers was considered for bottom-end on drop. The impact acceleration and strain data were obtained to verify the reliability of the analysis results. Before and after the drop tests, measurement of the dose rate and the radiographic testing for concrete wall, and measurement of the wall thickness of steel plate were conducted to evaluate the radiation shielding integrity. Also, measurement of bolt torque, and visual inspection were conducted to evaluate the loss or dispersion of radioactive contents. After the drop tests, the radiation dose rate on the container surface did not increase by more than 20%, and there was no crack in the concrete. In addition, the thickness of the steel plate did not change within the measurement error. Therefore, the radiation shielding integrity of the container was maintained. After the drop tests, the lid bolts were not damaged and there was no loss of pretension in the lid bolts. In addition, there was no loss or dispersion of the contents as a result of visual inspection. In order to prove the reliability of the drop analysis results, safety verifications were performed using the drop test results, and the appropriate conservatism for the analysis results and the validity of the analysis model were confirmed. Therefore, the structural integrity of the waste containers was maintained under the drop test conditions.

Organic waste generated by small and medium-sized (S&M-sized) metal decontamination in NPP decommissioning. To lower the concentration of these organic substances for a level acceptable at the disposal site, the project of “Development of Treatment Process of Organic Decontamination Liquid Wastes from Decommissioning of Nuclear Power Plants” is being carried out. The conditioning and treatment process of organic liquid waste was designed. Also, the literature was investigated to make simulated organic liquid waste, and the composition of these waste was analyzed and compared. As the decontamination agent, organic acids such as EDTA, oxalic acid, citric acid are used. The sum of the concentrations of these organic materials was set to a maximum value of 1,000 ppm. The major metal ions of the decontamination liquid waste estimated are 59Fe, 51Cr, 54Mn, 63Ni, and the concentrations are respectively 527, 163, 161, 159 ppm. Additional major metal ions are 60Co, 58Co, 137Cs. 58Co is replaced by 60Co because it has the same chemical properties as 60Co. Unlike the HLW, the contamination level of S&M-sized metal in primary system was quite low, so 60Co is set to 2,000 Bq/g. Considering the contribution of fission and gamma ray dose constant, 137Cs was estimated to 360 Bq/g. Also, suspended solids of decontamination liquid waste were set at 500 ppm. Under these assumptions, the simulated organic liquid waste was made, and then organic substances and metal ions were analyzed with TOC analyzer and ICP-OES. The TOC analysis value was expected to 392 ppm in consideration of the equivalent organic quantity. the test result was 302 ppm. Some of organics appears to have been decomposed by acid. The values of metal ions (Fe3+, Cr3+, Mn2+, Ni2+) analyzed by ICP-OES are 139, 4, 152, 158 ppm, respectively. A large amount of Cr3+ and Fe3+ were expected to exist as ions, but they existed in the form of suspended solid. Mn2+ and Ni2+ came out similar to the expected values. The designed conditioning and treatment process is largely divided into pretreatment, conditioning, and decomposition processes. After collecting in the primary liquid waste storage tank, large particulate impurities and suspensions are removed through a pretreatment process. In the conditioning process, treated liquid waste passes through UF/RO membrane system, and pure water is discharged to the environment after monitoring. Concentrated water is decomposed in the electrochemical catalyst decomposition process, then this water secondarily passes through the RO membrane system and then discharged to the environment after monitoring. Through an additional experiment, the conditioning and treatment process will be verified.

In this study, the current situation of recycling domestic and foreign metal clearance waste was reviewed to suggest the optimal recycling scenario for metal clearance waste that occurs the most when decommission nuclear power plants. Factors that can directly or indirectly affect the recycling of metal clearance waste were analyzed and evaluation criteria that can be used to evaluate optimal recycling measures were prepared. Using this, a scenario for recycling the optimal metal clearance waste suitable for the domestic environment was proposed. As a result of comparing/reviewing the importance of the first level of the evaluation criteria, public acceptance, national policy, and regulatory requirements were evaluated as the most important ones, and recycling acceptance and regulatory requirements were evaluated as the most important the second level of evaluation criteria. As a result of reviewing the clearance waste recycling scenario, it was evaluated that unrestricted recycling scenario was preferred. This may be because the survey subjects are composed of experts in the nuclear power field, so they know recycling of clearance waste in general industries does not significantly affect radiation safety. However even if it is clearance waste, the public may feel reluctant to recycle just because it was discharged from nuclear power plants, so policy and institutional improvements are needed to reassure the public along with the scientific safety of clearance waste. In addition, in order to improve public acceptance, it seems necessary to prepare specific measures to ensure the participation of public in the entire decommissioning process, share related information, and disclose all routes from generation to disposal of decommissioning waste. Considering that research on domestic clearance waste recycling options has not been activated, this study is significant in that it derives a scenario for recycling metal clearance waste that can be implemented. Also, it is expected that the evaluation criteria derived from this study will be used significantly when establishing a radioactive waste management strategy.