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

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.690-695
• /
• 1998

원자력시설에서 배출되는 고준위 방사선 페기물이나 TRU 둥의 심지층처분의 보완책으로서 핵변환 (Transmutation) 처리방안이 연구되고 있다 이 핵변환시스템의 냉각재로서 액체금속류가 고려되고 있다. 본 연구에서는 핵변환로에 적합한 냉각물질을 도출하기 위해서 보다 합리적인 선정방법으로서 의사결정방법을 이용하여 중점비교 대상인 나트륨(Na), 나트륨-칼륨 합금(Na-K alloy), 납(Pb), 납-비스므스 합금(Pb-Bi alloy)에 대한 정량적 평가를 시도하였다. 아울러 이 냉각재 후보물질에 대한 냉각재로서의 적합성 여부를 비교 검토하였다. 본 방법을 이용한 결과 핵 변화로의 냉각재로서는 납-비스므스 합금이 가장 적합한 것으로 평가되었다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.9 no.3
• /
• pp.169-179
• /
• 2011

There are two types of nuclear reactors in Korea and they are PWR type and CANDU type. The safe management of the spent fuels from these reactors is very important factor to maintain the sustainable energy supply with nuclear power plant. In Korea, a reference disposal system for the spent fuels has been developed through a study on the direct disposal of the PWR and CANDU spent fuel. Recently, the research on the demonstration and the efficiency analyses of the disposal system has been performed to make the disposal system safer and more economic. PWR spent fuels which include a lot of reusable material can be considered being recycled and a study on the disposal of HLW from this recycling process is being performed. CANDU spent fuels are considered being disposed of directly in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System (KRS) which was to dispose of both PWR type and CANDU type, the more effective CANDU spent fuel disposal systems were developed. To do this, the disposal canister for CANDU spent fuels was modified to hold the storage basket for 60 bundles which is used in nuclear power plant. With these modified disposal canister concepts, the disposal concepts to meet the thermal requirement that the temperature of the buffer materials should not be over $100^{\circ}C$ were developed. These disposal concepts were reviewed and analyzed in terms of disposal effective factors which were thermal effectiveness, U-density, disposal area, excavation volume, material volume etc. and the most effective concept was proposed. The results of this study will be used in the development of various wastes disposal system together with the HLW wastes from the PWR spent fuel recycling process.

• Tunnel and Underground Space
• /
• v.34 no.2
• /
• pp.104-126
• /
• 2024

Buffer and backfill play an essential role in isolating high-level radioactive waste and retard the migration of leaked radionuclides in deep geological disposal system. A bentonite mixture, which exhibits a swelling property, is considered for buffer and backfill materials, and excessive groundwater inflow from surrounding rock mass may affect stability and efficiency of their role as an engineered barrier. Therefore, stringent quality control as well as in-situ installation management and inflow water constrol for buffer and backfill are required to ensure the safety of deep disposal facilities. In this study, we analyzed the design requirements of buffer and backfill by examining various laboratory tests and a field study of the Steel Tunnel Test at the Äspö Hard Rock Laboratory in Sweden. We introduced how to control the quality of buffer and backfill construction in-field, and also presented how to handle excessive groundwater inflow into disposal caverns, validating the groundwater retention capacity of bentonite pellets and the effectiveness of geotexile use.

• Journal of the Korean Geotechnical Society
• /
• v.37 no.4
• /
• pp.19-26
• /
• 2021

The high-level radioactive waste (HLW) produced from nuclear power plants is disposed in a rock-mass at a depth of hundreds meters below the ground level. Since HLW is very dangerous to human being, it must be disposed of safely by the engineered barrier system (EBS). The EBS consists of a disposal canister, backfill material, buffer material, and so on. When the components of EBS are installed, gaps inevitably exist not only between the rock-mass and buffer material but also between the canister and buffer material. The gap can reduce water-retarding capacity and heat release efficiency of the buffer material, so it is necessary to investigate properties of gap-filling materials and to analyze gap spacing effect. Furthermore, there has been few researches considering domestic disposal system compared to overseas researches. In this reason, this research derived the peak temperature of the bentonite buffer material considering domestic disposal system based on the numerical analysis. The gap between the canister and buffer material had a minor effect on the peak temperature of the bentonite buffer material, but there was 40% difference of the peak temperature of the bentonite buffer material because of the gap existence between the buffer material and rock mass.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.6 no.2
• /
• pp.119-128
• /
• 2008

Thermal assessment of a new CANDU spent fuel disposal system, which improves the retrievability of the spent fuel and enhances the densification factor compared with the Korean Reference disposal System, is carried out in this study. The canisters for CANDU spent fuels are stored for long term and cooled by natural convection in the proposed disposal system for the retrievability. The steady state thermal analyses for proposed CANDU disposal system are carried out with the ANSYS 10.0 CFX code. The thermal analyses are performed through two steps. At the first step, the sensitivity of the disposal tunnel spacing is analysed. The differences of maximum temperatures by several tunnel spacings are calculated at three points in the disposal tunnel. The result shows that the differences of the temperature at the three points are almost negligible because 99% of the decay heat is removed by natural convection. At the second procedure, 60m tunnel spacing with a ventilation system instead of natural convection is considered. The result is applied to the calculation of the canister surface temperature in disposal tunnel as boundary conditions. Consequently, the average and the maximum surface temperature of disposal canisters are $79.9^{\circ}C$ and $119^{\circ}C$, respectively. The inner maximum temperature of a basket in the disposal canister is calculated as $140.9^{\circ}C$. The maximum temperature of the basket meets the thermal requirement for the CANDU spent fuel cladding.

• Economic and Environmental Geology
• /
• v.56 no.5
• /
• pp.603-618
• /
• 2023

This study focused on evaluating the suitability of the WRK (waste repository Korea) bentonite as a buffer material in the SNF (spent nuclear fuel) repository. The U (uranium) adsorption/desorption characteristics and the adsorption mechanisms of the WRK bentonite were presented through various analyses, adsorption/desorption experiments, and kinetic adsorption modeling at various pH conditions. Mineralogical and structural analyses supported that the major mineral of the WRK bentonite is the Ca-montmorillonite having the great possibility for the U adsorption. From results of the U adsorption/desorption experiments (intial U concentration: 1 mg/L) for the WRK bentonite, despite the low ratio of the WRK bentonite/U (2 g/L), high U adsorption efficiency (>74%) and low U desorption rate (<14%) were acquired at pH 5, 6, 10, and 11 in solution, supporting that the WRK bentonite can be used as the buffer material preventing the U migration in the SNF repository. Relatively low U adsorption efficiency (<45%) for the WRK bentonite was acquired at pH 3 and 7 because the U exists as various species in solution depending on pH and thus its U adsorption mechanisms are different due to the U speciation. Based on experimental results and previous studies, the main U adsorption mechanisms of the WRK bentonite were understood in viewpoint of the chemical adsorption. At the acid conditions (＜pH 3), the U is apt to adsorb as forms of UO₂²⁺, mainly due to the ionic bond with Si-O or Al-O(OH) present on the WRK bentonite rather than the ion exchange with Ca²⁺ among layers of the WRK bentonite, showing the relatively low U adsorption efficiency. At the alkaline conditions (>pH 7), the U could be adsorbed in the form of anionic U-hydroxy complexes (UO₂(OH)₃^-, UO₂(OH)₄^2-, (UO₂)₃(OH)₇^-, etc.), mainly by bonding with oxygen (O^-) from Si-O or Al-O(OH) on the WRK bentonite or by co-precipitation in the form of hydroxide, showing the high U adsorption. At pH 7, the relatively low U adsorption efficiency (42%) was acquired in this study and it was due to the existence of the U-carbonates in solution, having relatively high solubility than other U species. The U adsorption efficiency of the WRK bentonite can be increased by maintaining a neutral or highly alkaline condition because of the formation of U-hydroxyl complexes rather than the uranyl ion (UO₂²⁺) in solution,and by restraining the formation of U-carbonate complexes in solution.

• Economic and Environmental Geology
• /
• v.55 no.1
• /
• pp.1-17
• /
• 2022

The mechanical and thermal properties of the rock masses can affect the performance associated with both the isolating and retarding capacities of radioactive materials within the deep geological disposal system for High-Level Radioactive Waste (HLW). In this study, the essential parameters for the site descriptive model (SDM) related to the rock mechanics and thermal properties of the HLW disposal facilities site were reviewed, and the technical background was explored through the cases of the preceding site descriptive models developed by SKB (Swedish Nuclear and Fuel Management Company), Sweden and Posiva, Finland. SKB and Posiva studied parameters essential for the investigation and evaluation of mechanical and thermal properties, and derived a rock mechanics site descriptive model for safety evaluation and construction of the HLW disposal facilities. The rock mechanics SDM includes the results obtained from investigation and evaluation of the strength and deformability of intact rocks, fractures, and fractured rock masses, as well as the geometry of large-scaled deformation zones, the small-scaled fracture network system, thermal properties of rocks, and the in situ stress distribution of the disposal site. In addition, the site descriptive model should provide the sensitivity analysis results for the input parameters, and present the results obtained from evaluation of uncertainty.

• Tunnel and Underground Space
• /
• v.29 no.6
• /
• pp.439-455
• /
• 2019

A buffer is the major component of a high level radioactive waste repository. Due to their thermal conductivity and low permeability, bentonites have been considered as a key component of a buffer system in most countries. The deep geological condition generates ground water inflow and results in swelling pressure in the buffer and backfill. Investigation of swelling pressure of bentonite buffer is an important task for the safe disposal system. The swelling pressure that can be critical is affected by mechanical and hydro properties of the system. Therefore, in this study, a sensitivity analysis was conducted to examine the effect of hydro-mechanical (HM) behaviors in the MX-80 bentonite. Based on the results of the swelling pressure generation with HM model parameters, a coupled HM analysis of an unsaturated buffer and backfill in a deep geological repository was also carried out to investigate the major factor of the swelling pressure generation.

• Journal of Korean Tunnelling and Underground Space Association
• /
• v.24 no.6
• /
• pp.451-464
• /
• 2022

The high-level nuclear waste repository is a deep geological disposal system exposed to complex environmental conditions such as high temperature, radiation, and ground-water due to handling spent nuclear fuel. Continuous exposure can lead to cracking and deterioration of the structure over time. On the other hand, the high-level nuclear waste repository requires an ultra-long life expectancy. Thus long-term structural health monitoring is essential. Various sensors such as an accelerometer, earth pressure gauge, and displacement meter can be used to monitor the health of a structure, and a piezoelectric sensor is generally used. Therefore, it is necessary to develop a highly durable sensor based on the durability assessment of the piezoelectric sensor. This study designed an accelerated life test for durability assessment and life prediction of the piezoelectric sensor. Based on the literature review, the number of accelerated stress levels for a single stress factor, and the number of samples for each level were selected. The failure mode and mechanism of the piezoelectric sensor that can occur in the environmental conditions of the high-level waste repository were analyzed. In addition, two methods were proposed to investigate the maximum harsh condition for the temperature stress factor. The reliable operating limit of the piezoelectric sensor was derived, and a reasonable accelerated stress level was set for the accelerated life test. The suggested methods contain economical and practical ideas and can be widely used in designing accelerated life tests of piezoelectric sensors.