• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.212-218
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• 2004

저준위 방사성폐기물 안전성 평가를 위해 대상 부지를 선정하고 경계 조건을 도입하였다. 연안에 처분장이 위치할 경우를 가상하여 처분장 심도 및 단열까지의 거리등에 대한 지하수 유동 민감도를 분석하였다. 또한 처분장 진입 터널이 지하수 유동에 미치는 영향을 평가하였다. 이를 통하여 각 암반별 이동 거리 및 시간을 CONNECTFLOW를 이용해 산정하고, 그 결과들이 방사선적 안전성 종합 평가 코드인 MASCOT의 입력 자료로 활용되도록 하였다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.287-297
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• 2014

For several decades, many countries operating nuclear power plants have been studying the various disposal alternatives to dispose of the spent nuclear fuel or high-level radioactive waste safely. In this paper, as a direct disposal of spent nuclear fuels for deep geological disposal concept, the rod consolidation from spent fuel assembly for the disposal efficiency was considered and analyzed. To do this, a concept of spent fuel rod consolidation was described and the related concepts of disposal canister and disposal system were reviewed. With these concepts, several thermal analyses were carried out to determine whether the most important requirement of the temperature limit for a buffer material was satisfiedin designing an engineered barrier of a deep geological disposal system. Based on the results of thermal analyses, the deposition hole distance, disposal tunnel spacing and heat release area of a disposal canister were reviewed. And the unit disposal areas for each case were calculated and the disposal efficiencies were evaluated. This evaluation showed that the rod consolidation of spent nuclear fuel had no advantages in terms of disposal efficiency. In addition, the cooling time of spent nuclear fuels from nuclear power plant were reviewed. It showed that the disposal efficiency for the consolidated spent fuel rods could be improved in the case that cooling time was 70 years or more. But, the integrity of fuels and other conditions due to the longer term storage before disposal should be analyzed.

• Proceedings of the Korean Society of Soil and Groundwater Environment Conference
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• 2001.09a
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• pp.120-123
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• 2001

국내 원전에서 배출되는 사용후핵연료의 안정한 장기 보관과 평화적 재활용을 위한 많은 연구중에서 금속저장체 연구는 사용후핵연료의 장기 보관에 따른 부식 문제 해결과 방사성 붕괴열 제거 관점에서 관심을 끌고 있다. 본 논문에서는 고준위 방사성폐기물로 분류되는 금속저장체를 국내 심지층 암반에 영구 처분했을 때 발생하는 처분장 요구 면적과 영구 처분에 따른 방사선적 안전성에 관한 고찰을 통해 금속 저장체를 이용한 사용후핵연료 관리의 장단점을 분석하였다. 예비 조사 결과 주어진 데이터 조건에서는 금속 저장체는 고준위 방사성폐기물 처분장의 면적을 5 배 정도 줄일 수 있을 것이며 방사선적 안전성 또한 우수한 것으로 판명되었으나 향후 경제성과 핵비확산성을 고려한 종합적인 타당성 연구가 수행되는 것이 바람직하다고 판명되었다.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.29-35
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• 1990

Uranium solubilities have been calculated for a range of conditions expected in a nuclear waste disposal repository. Variables taken into consideration include the pH and Eh range expected for deep groundeaters, the effect of the composition of groundwater. The model used in these calculations is based on the assumption of chemical equilibrium. Calculations show that the major variables influencing uranium solubility under the repository conditions are pH and Eh. The results of this study can be applied to an assessment of the nuclear waste disposal.

• Journal of Korean Tunnelling and Underground Space Association
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• v.25 no.6
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• pp.509-523
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• 2023

The high-level nuclear waste (HLW) repository is exposed to complex environmental conditions consisting of high temperature, high humidity, and radiation, resulting in structural deterioration. Therefore, structural health monitoring is essential, and piezo sensors are used to detect cracks and estimate strength. However, since the monitoring sensors installed in the disposal tunnel and disposal container cannot be replaced or removed, the quantitative life of the monitoring sensor and its suitability must be assessed. In this study, the life of a piezo sensor for monitoring was assessed using an accelerated life test (ALT). The failure mode and mechanism of the piezo sensor under high temperature conditions were determined, and temperature stress's influence on the piezo sensor's life was analyzed. ALT was conducted on temperature stress and the relationship between temperature stress and piezo sensor life was suggested. The life of the piezo sensor was assessed using the Weibull probability distribution and the Arrhenius acceleration model. The suggested relationship can be used in multiple stress ALT designs for more precise life assessment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.1
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• pp.73-86
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• 2015

The use of complimentary indicators, other than radiation dose and risk, to assess the safety of radioactive waste disposal has been discussed in a number of publications for providing the reasonable assurance of disposal safety and convincing the public audience. In this study, the radionuclide flux was selected as performance indicator to appraise the performance of engineered barriers and natural barrier in the Wolsong low- and intermediate-level waste disposal facility. Radionuclide flux showing the retention capability by each compartment of the disposal system is independent of assumptions in biosphere model and exposure pathways. The scenario considered as the normal scenario of disposal facility has been divided into intact or degraded silo concrete conditions. In the intact silo concrete, the radionuclide flux has been assessed with respect to the radionuclide retardation performance of each engineered barrier. In the degraded silo concrete, the radionuclide flux has been explored based on the performance degradation of engineered barriers and the relative significance of natural barrier quantitatively. The results can be used to optimally design the near-surface disposal facility being planned as the second project phase. In the future, additional complimentary indicators will be employed for strengthening the safety case for improving the public acceptance of low- and intermediate-level waste disposal facility.

• Journal of Korean Tunnelling and Underground Space Association
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• v.26 no.3
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• pp.191-208
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• 2024

In this study, a hydro-mechanical-damage coupled analysis model was developed to evaluate the structural safety of radioactive waste disposal structures. The Mazars damage model, widely used to model the fracture behavior of brittle materials such as rocks or concrete, was coupled with conventional hydro-mechanical analysis and the developed model was verified via theoretical solutions from literature. To derive the numerical input values for damage-coupled analysis, uniaxial compressive strength and Brazilian tensile strength tests were performed on concrete samples made using the mix ratio of the disposal concrete silo cured under dry and saturated conditions. The input factors derived from the laboratory-scale experiments were applied to a two-dimensional finite element model of the concrete silos at the Wolseong Nuclear Environmental Management Center in Gyeongju and numerical analysis was conducted to analyze the effects of damage consideration, analysis technique, and waste loading conditions. The hydro-mechanical-damage coupled model developed in this study will be applied to the long-term behavior and stability analysis of deep geological repositories for high-level radioactive waste disposal.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.491-505
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• 2018

As an alternative to deep geological disposal technology, which is considered as a reference concept, the domestic applicability of deep borehole disposal technology for high level radioactive waste, including spent fuel, has been preliminarily evaluated. Usually, the environment of deep borehole disposal, at a depth of 3 to 5 km, has more stable geological and geo-hydrological conditions. For this purpose, the characteristics of rock distribution in the domestic area were analyzed and drilling and investigation technologies for deep boreholes with large diameter were evaluated. Based on the results of these analyses, design criteria and requirements for the deep borehole disposal system were reviewed, and preliminary reference concept for a deep borehole disposal system, including disposal container and sealing system meeting the criteria and requirements, was developed. Subsequently, various performance assessments, including thermal stability analysis of the system and simulation of the disposal process, were performed in a 3D graphic disposal environment. With these analysis results, the preliminary evaluation of the domestic applicability of the deep borehole disposal system was performed from various points of view. In summary, due to disposal depth and simplicity, the deep borehole disposal system should bring many safety and economic benefits. However, to reduce uncertainty and to obtain the assent of the regulatory authority, an in-situ demonstration of this technology should be carried out. The current results can be used as input to establish a national high-level radioactive waste management policy. In addition, they may be provided as basic information necessary for stakeholders interested in deep borehole disposal technology.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.29-36
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• 2019

High-level radioactive waste (HLW) such as spent fuel is inevitably produced when nuclear power plants are operated. A geological repository has been considered as one of the most adequate options for the disposal of HLW, and it will be constructed in host rock at a depth of 500~1,000 meters below ground level with the concept of an engineered barrier system (EBS) and a natural barrier system. The compacted bentonite buffer is one of the most important components of the EBS. As the compacted bentonite buffer is located between disposal canisters with spent fuel and the host rock, it can restrain the release of radionuclides and protect canisters from the inflow of groundwater. Because of inflow of groundwater into the compacted bentonite buffer, it is essential to investigate soil-water characteristic curves (SWCC) of the compacted bentonite buffer in order to evaluate the entire safety performance of the EBS. Therefore, this paper conducted laboratory experiments to analyze the SWCC for a Korean Ca-type compacted bentonite buffer considering dry density, confined or unconfined condition, and drying or wetting path. There was no significant difference of SWCC considering dry density under unconfined condition. Furthermore, it was found that there was higher water suction in unconfined condition that in confined condition, and higher water suction during drying path than during wetting path.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.63-75
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• 2012

This paper gives some basic requirements and preferences of various geological environmental conditions for the final deep geological repository of spent nuclear fuel (SNF). This study also indicates how the requirements and preferences are to be considered prior to the selection of sites for a site investigation as well as the final disposal in Korea. The results of the study are based on the knowledge and experience from the IAEA and NEA/OECD as well as the advanced countries in SNF disposal project. This study discusses and suggests preliminary guideline of the disposal requirements including geological, mechanical, thermal, hydrogeological, chemical and transport properties of host rock with long term geological stabilities which influence the functions of a multi-barrier disposal system. To apply and determine whether requirements and preferences for a given parameter are satisfied at different stages during a site selection and suitability assessment of a final disposal site, the quantitative criteria in each area should be formulated with credibility through relevant research and development efforts for the deep geological environment during the site screening and selection processes as well as specific studies such as productions of safety cases and validation studies using a generic underground research laboratory (URL) in Korea.