• The Journal of Engineering Geology
• /
• v.33 no.2
• /
• pp.335-353
• /
• 2023

Site selection processes for high-level radioactive waste disposal facilities in different countries differ in terms of local geology and degree of public engagement. There seem to be three alternative processes for site selection: (1) selection with community consent after government choice; (2) selection with continuous community engagement after exclusion of unsuitable areas based on existing survey data; or (3) site selection where communities have expressed a willingness to participate. The Yucca Mountain site in Nevada, USA, was selected as the final disposal site by process (1) through six stages, but its development was suspended owing to opposition from the local governor and environmental groups. In Sweden, Switzerland, and Germany, process (2) is used and sites are selected through three stages. Sweden and Switzerland have completed site selection, and Germany is currently engaged in the process. The UK adopted process (3) with six stages, although the process has been suspended owing to poor community participation. In Korea, temporary storage facilities for spent nuclear fuel will reach saturation from 2030, so site selection must be promoted through various laws and systems, with continuous communication with local communities based on transparent and scientifically undertaken procedures.

• The Journal of Engineering Geology
• /
• v.33 no.4
• /
• pp.725-736
• /
• 2023

The compliance of deep geological disposal facilities for high-level radioactive waste with safety objectives requires consideration of uncertainties owing to temporal changes in the disposal system. A comprehensive review and analysis of the characteristics of this evolution should be undertaken to identify the effects on multiple barriers and the biosphere. We analyzed the evolution of the buffer, backfill, plug, and closure regions during the early phase of the post-closure period as part of a long-term performance assessment for an operating license application for a deep geological repository in Finland. Degradation mechanisms generally expected in engineered barriers were considered, and long-term evolution features were examined for use in performance assessments. The importance of evolution features was classified into six categories based on the design of the Finnish case. Results are expected to be useful as a technical basis for performance and safety assessment in developing the Korean deep geological disposal system for high-level radioactive waste. However, for a more detailed review and evaluation of each feature, it is necessary to obtain data for the final disposal site and facility-specific design, and to assess its impact in advance.

• Journal of Radiation Industry
• /
• v.17 no.1
• /
• pp.33-44
• /
• 2023

The deep geological repository of high-level radioactive waste shall be designed to meet the safety objective set in the form of radiation dose or corresponding risk to protect human and the environment from radiation exposure. Engineering feasibility and conformity with the safety objective of the facility conceptual design can be demonstrated by comparing the assessment result using the computational model for scenario(s) describing the radionuclide release and transport from repository to biosphere system. In this study, as the preliminary study for developing the high-level radioactive waste disposal facility in Korea, we reviewed and analyzed the entire list of FEPs and how to handle each FEP from a general point of view, which are selected for the geosphere region in the radiological safety assessment performed for the license application of the KBS-3 type deep geological repository in Finland and Sweden. In Finland, five FEPs (i.e., stress redistribution, creep, stress redistribution, erosion and sedimentation in fractures, methane hydrate formation, and salt exclusion) were excluded or ignored in the radionuclide release and transport assessment. And, in Sweden, six FEPs (i.e., creep, surface weathering and erosion, erosion/sedimentation in fractures, methane hydrate formation, radiation effects (rock and grout), and earth current) were not considered for all time frames and earthquake out of a total of 25 FEPs for the geosphere. Based on these results, an FEP list (draft) for the geosphere was derived, and the relative importance of each item was evaluated for conducting the radiological safety assessment of the domestic deep geological disposal facility. Since most of information on the disposal facility in Korea has not been determined as of now, it is judged that all FEP items presented in Table 3 should be considered for the radiological safety assessment, and the relative importance derived from this study can be used in determining whether to apply each item in the future.

• Nuclear Engineering and Technology
• /
• v.56 no.4
• /
• pp.1385-1397
• /
• 2024

Bentonite is a recommended material for the multiple barriers in the final disposal of low-level radioactive waste (LLW) to prevent groundwater intrusion and nuclear species migration. However, after drying-wetting cycling during the repository construction stage and ion exchange with the concrete barrier in the long-term repository, the bentonite mechanical behaviors, including swelling capacity and hydraulic conductivity, would be further influenced by the groundwater intrusion, resulting in radioactive leakage. To comprehensively examine the factors on the mechanical characteristics of bentonite, this study presented scenarios involving MX-80 and KV-1 bentonites subjected to drying-wetting cycling and accelerated ion migration. The experiments subsequently measured free swelling, swelling pressure, and hydraulic conductivity of bentonites with intrusions of seawater, high pH, and low pH solutions. The results indicated that the solutions caused a reduction in swelling volume and pressure, and an increase in hydraulic conductivity. Specifically, the swelling capability of bentonite with drying-wetting cycling in the seawater decreased significantly by 60%, while hydraulic conductivity increased by more than three times. Therefore, the study suggested minimizing drying-wetting cycling and preventing seawater intrusion, ensuring a long service life of the multiple barriers in the LLW repository.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2003.11a
• /
• pp.168-171
• /
• 2003

The IMEF(Irradiated Materials Examination Facility), located in KAERI site, is a hot cell facility to test and evaluate the irradiation defects or embrittlement through post-irradiation examination(PIEs) of irradiated nuclear fuels and structural materials which are come from HANARO research reactor and commercial nuclear power plants. Therefore, to carry out its own function, the high level solid radioactive wastes, produced through PIEs, are periodically carried out and managed from hot cell to monolith. So far, approximately 30 drums which contains 50 liters are transported to monolith, and it is shown that the quantity is slowly increasing, In this paper, the procedures and work contents of the high level solid radwaste carrying out and management for IMEF are described in detail.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2004.02a
• /
• pp.130-139
• /
• 2004

The cladding materials remaining after reprocessing process of the nuclear fuel, generally called as hulls, are classified as a high-level radioactive waste. They are usually packaged in the container for disposal after being compacted, melted, or solidified into the matrix. The efforts to fabricate a better ingot for a more favorable disposal to the environment have failed due to the technical difficulties encountered in the chemical decontamination method. In the early 1990s, the accumulation of radio-chemical data on hulls and the advent of new technology such as a laser or plasma have made the pre-treatment of the hulls more efficient. This paper summarizes the information regarding the radio-chemical analysis of the hull through a literature survey and determines the characteristics of the hull and depth profile of the radio-nuclides within the hull thickness. The feasibility study was carried out to evaluate the reduction of the radioactivity by peeling off the surface of the hull with the application of laser technology.

• Nuclear Engineering and Technology
• /
• v.53 no.10
• /
• pp.3359-3366
• /
• 2021

An engineered barrier system (EBS) for the deep geological disposal of high-level radioactive waste (HLW) is composed of a disposal canister, buffer material, gap-filling material, and backfill material. As the buffer fills the empty space between the disposal canisters and the near-field rock mass, heat energy from the canisters is released to the surrounding buffer material. It is vital that this heat energy is rapidly dissipated to the near-field rock mass, and thus the thermal conductivity of the buffer is a key parameter to consider when evaluating the safety of the overall disposal system. Therefore, to take into consideration the sizeable amount of heat being released from such canisters, this study investigated the thermal conductivity of Korean compacted bentonites and its variation within a temperature range of 25 ℃ to 80-90 ℃. As a result, thermal conductivity increased by 5-20% as the temperature increased. Furthermore, temperature had a greater effect under higher degrees of saturation and a lower impact under higher dry densities. This study also conducted a regression analysis with 147 sets of data to estimate the thermal conductivity of the compacted bentonite considering the initial dry density, water content, and variations in temperature. Furthermore, the Kriging method was adopted to establish an uncertainty metamodel of thermal conductivity to verify the regression model. The R² value of the regression model was 0.925, and the regression model and metamodel showed similar results.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.745-746
• /
• 2004

• Tunnel and Underground Space
• /
• v.17 no.1 s.66
• /
• pp.32-42
• /
• 2007

Severe damage can occur around deposition holes due to complex interaction of thermo-hydro-mechanical (THM) loading during the long term operation of high level radioactive waste repository. Many candidate sites for repository are located in crystalline rock mass, therefore mechanism of damage follows the form of brittle fracture and failure. This paper briefly introduces major outcomes from 15 years international collaborative project, DECOVALEX, and presents major study results for current ongoing benchmark test study from DECOVALEX-THMC, to evaluate the effect of THM loading to rock mass in excavation damaged zone (EDZ) near deposition holes. Through benchmark test model by simplifying THM loading to boundary loading obtained numerical results are compared, and discrete fracture interaction after up to 1 million years operation is discussed.

• Journal of Radiation Industry
• /
• v.17 no.1
• /
• pp.45-59
• /
• 2023

The transport calculation for a wide variety of radionuclides contained in high-level radioactive waste, especially spent nuclear fuel, is computationally difficult, and input data collection for this also take a considerable amount of time. Accordingly, considering limited resources, it is possible to reduce the calculation time while minimizing impact on accuracy by including only radionuclides important to calculation result through applying some criteria among potential radiation source terms that may release into environment. In this paper, therefore, we reviewed and analyzed the screening process performed to select radionuclides to be considered in the safety assessment for the KBS-3 type repository in Sweden and Finland. In both countries, it was confirmed that a list of radionuclides was selected by comprehensively considering screening criteria such as radioactivity inventory, half-life, radiotoxicity, risk quotient, and transport properties, and etc. A comparison of radionuclides included in the radiological safety assessment in both countries suggests that most of nuclides are considered in common, and a few nuclides considered only in one country are due to differences in decay chain treatment or spent fuel types. As of now, since most of information on the disposal facility in Korea has not been determined, it is necessary to comprehensively model release and transport of all radionuclides considered in Sweden and Finland when performing the radiological safety assessment. Based on these results, we derived the screening concept of selecting a list of radionuclides to be considered in the radiological safety assessment for the domestic KBS-3 type geological disposal facility, and this result is expected to be used as technical basis for confirming conformity with the safety objective. In a more detailed evaluation reflecting domestic characteristics in the future, it would be desirable to consider only radionuclides selected in accordance with the screening procedure. However, further research should be conducted to determine the quantitative limit for each criteria.