• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.63-74
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• 2020

The operational safety assessment is an important part of a safety case for the deep geological repository of spent fuels. It consists of different stages such as the identification of initiating events, event tree analysis, fault tree analysis, and evaluation of exposure doses to the public and radiation workers. This study develops a probabilistic safety assessment method for the operational safety assessment and establishes an assessment framework. For the event and fault tree analyses, we propose the advanced information management system for probabilistic safety assessment (AIMS-PSA Manager). In addition, we propose the Radiological Safety Analysis Computer (RSAC) program to evaluate exposure doses to the public and radiation workers. Furthermore, we check the applicability of the assessment framework with respect to drop accidents of a spent fuel assembly arising out of crane failure, at the surface facility of the KRS⁺ (KAERI Reference disposal System for SNFs). The methods and tools established through this study can be used for the development of a safety case for the KRS⁺ system as well as for the design modification and the operational safety assessment of the KRS⁺ system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.1-12
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• 2022

Bentonite is the most probable candidate to be used as a buffer in a deep geological repository with high swelling properties, hydraulic conductivity, thermal conductivity, and radionuclide sorption ability. Among them, the radionuclide sorption ability prevents or delays the transport of radionuclides into the nearby environment when an accident occurs and the radionuclide leaks from the canister, so it needs to be strengthened in terms of long-term disposal safety. Here, we proposed a surface modification method in which some inorganic additives were added to form NaP zeolite on the surface of the bentonite yielded at Yeonil, South Korea. We confirmed that the NaP zeolite was well-formed on the bentonite surface, which also increased the sorption efficiency of Cs and Sr from groundwater conditions. Both NaP and NaX zeolite can be produced and we have demonstrated that the generation mechanism of NaX and NaP is due to the number of homogeneous/heterogeneous nucleation sites and the number of nutrients supplied from an aluminosilicate gel during the surface modification process. This study showed the potential of surface modification on bentonite to enhance the safety of deep geological radioactive waste repository by improving the radionuclide sorption ability of bentonite.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2819-2827
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• 2022

The disposal area of a deep geological repository (DGR) for the disposal of spent nuclear fuels (SNFs) is estimated considering the spacing between deposition holes and between disposal tunnels, as determined by a thermal analysis using the decay heat of a reference SNF. Given the relatively large amount of decay heat of the reference SNF, the disposal area of the DGR is found to be overestimated. Therefore, we develop a computer program using MATLAB, termed ACom (Assembly Combination), to combine SNFs when stored in canisters such that the decay heat per canister is evenly distributed. The stability of ACom was checked and the overall distribution of the decay heat per canister was analyzed. Finally, ACom was applied to disposal scenarios suggested in the conceptual design of a DGR for SNFs, and it was confirmed that the decay heat per canister could be evenly distributed and that the maximum decay heat of the canister could be much lower than that of a canister estimated using a reference SNF. ACom can be used to improve the disposal efficiency by reducing the disposal area of a DGR for SNFs by ensuringg a relatively even distribution of decay heat per canister.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4083-4095
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• 2023

Currently, various research institutes in Korea are conducting research to develop a safety case for deep geological repository for high-level radioactive waste (HLW). In the past, low and intermediate-level waste (LILW) was approved by a regulatory body by performing a post-closure safety assessment, but HLW has different disposal characteristics and safety objectives are different. Therefore, in the case of HLW, safety assessment should be performed based on these changed conditions, and specific procedures are also under development. In this paper, the regulatory status of prior research institutes, feature, event and process (FEP) and scenario development cases were investigated for well-organized FEP and scenario development methodologies. In addition, through the results of these surveys, the requirements and procedures necessary for the FEP and scenario development stage during the safety assessment of repository for HLW were presented. These review results are expected to be used to identify the overall status of previous studies in conducting post-closure risk assessment for HLW repository, starting with identifying regulatory requirements, the most basic element.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.393-400
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• 2006

In design of a deep geological repository for the high level wastes, it is very important that the temperature of the bentonite block should not be over $100^{\circ}C$ to maintain the integrity of the bentonite buffer block from the decay heat. In this study, for the layout of the repository to meet the requirement, the analysis of the disposal tunnel and disposal pit spacing was carried out. To do this, based on the reference repository concept, several cases of cooling times and disposal tunnel and disposal pit spacing were compared. The thermal stabilities of the disposal systems were analyzed in terms of the cooling time and spacing. The results showed that it was more desirable to determine the layout of the repository in terms of disposal pit spacing than the disposal tunnel spacing. The results of these analyses can be used in the deep geological repository design. The detailed analyses with the exact site characteristics data will reduce the uncertainty of the results.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2005.10a
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• pp.470-473
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• 2005

Spent nuclear fuels are regarded as a high level radioactive waste and they will be disposed in a deep geological repository. To maintain the safety of the repository for hundreds of thousands of years, the spent fuels are encapsulated in a disposal canister and the canister containing spent fuels should have the structural integrity and the corrosion resistance below the several hundreds meters from the ground surface. In this study, the concept of the spent fuel encapsulation process and the process equipment fur deep geological disposal were established. To do this, the design requirements, such as the functions and the spent fuel accumulations, were reviewed. Also, the design principles and the bases were established. Based on the requirements and the bases, the encapsulation process and the equipment from spent fuel receiving process to transferring canister into the underground repository including hot cell processes was established. The established concept of the spent fuel encapsulation process and the process equipment will be improved continuously with the future studies. And this concept can be effectively used in implementing the reference repository system of our own case.

• Nuclear Engineering and Technology
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• v.31 no.6
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• pp.23-38
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• 1999

This study addresses the reference spent fuel and its characteristics for developing a geological repository concept. As a disposal capacity of the reference repository system to be developed, spent fuel inventories were projected based on the basis of the Nuclear Energy Plan of the Long-term National Power Program. The reference spent fuel encompassing a variability in characteristics of all existing and future spent fuels of interest was defined. Key parameters in the reference fuel screening processes were the nuclear and mechanical design parameters and the burnup histories for existing spent fuels as of 1996 and for future spent fuels with the more extended burnup the initial enrichment and its expected turnup. The selected reference fuel was characterized in terms of initial enrichment, bumup, dimension, gross weight and age. Also the isotopic composition and the radiological properties are quantitatively identified. This information provided in this study could be used as input for repository system development and performance assessment and applied in fuel material balance evaluation for the various types of back-end fuel cycle studies.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.4
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• pp.349-358
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• 2005

In this study, analysis of the disposal tunnel spacing and disposal pit pitch was carried out, as a factor of the design to estimate the scale and layout of the repository. To do this, based on the reference repository concept and the engineered barrier concept, several cross sections of the disposal tunnel and disposal pit were established. After then, the mechanical and thermal stabilities of the established tunnels were analyzed. Also, an optimized disposal tunnel spacing and the disposal pit pitch reducing the excavation volume was proposed. The results of these analyses can be used in the deep geological repository design. The detailed analyses by the exact site characteristics data to reduce the uncertainty of the site and the modification for the optimization are required.

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

• Tunnel and Underground Space
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• v.21 no.6
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• pp.518-525
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• 2011

In this study, the natural ventilation pressure resulting from the large altitude difference which is a characteristic of high radioactive waste repository and the caloric value of the heat emitted by wastes was calculated and based on the results, natural ventilation quantities were calculated. A high radioactive waste repository can be considered as being operated through closed cycle thermodynamic processes similar to those of thermal engines. The heat produced by the heating of high radioactive wastes in the underground repository is added to the surrounding air, and the air goes up through the upcast vertical shaft due to the added heat while working on its surroundings. Part of the heat added by the work done by the air can be temporarily changed into mechanical energy to promote the air flow. Therefore, if a sustained and powerful heat source exists in the repository, the heat source will naturally enable continued cyclic flows of air. Based on this assumption, the quantity of natural ventilation made during the disposal of high radioactive wastes in a deep geological layer was mathematically calculated and based on the results, natural ventilation pressure of $74{\sim}183$Pa made by the stack effect was identified along with the resultant natural ventilation quantity of $92.5{\sim}147.7m^3/s$. The result of an analysis by CFD was $82{\sim}143m^3/s$ which was very similar to the results obtained by the mathematical method.