• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.2
• /
• pp.261-270
• /
• 2018

In this study, to replace the 'J-slot joint', a joint device between a disposal canister and an emplacement jig in Deep Borehole Disposal process, a novel joint device was designed and tested. The novel joint device was composed of a wedge on top of a disposal canister and a hook box at the end of a winch system. The designed joint device had merits in that it can recombine an emplaced canister freely without the replacement of the joint component. Moreover, it can be applied to various emplacement jigs such as drill pipes, wire-lines, and coiled tubing. To demonstrate the designed joint device, the joint device (${\Phi}110mm$, H 148 mm), a twin canister string (${\Phi}140mm$, H 1,105 mm), and a water tube (${\Phi}150mm$, H 1,500 mm) as a borehole model were manufactured at 1/3 scale. As deployment muds, Na-type bentonite (MX-80) and Ca-type (GJ II) bentonite muds were prepared at solid contents of 7wt% and 28wt%, respectively. The manufactured joint device showed good performance in pure water and viscous muds, with an operation speed of $10m{\cdot}min^{-1}$. It was concluded that the newly developed joint device can be used for the emplacement and retrieval of a deep disposal canister, below 3~5 km, in the future.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.4
• /
• pp.405-418
• /
• 2019

Based on spent fuels characteristics from domestic nuclear power plants and a disposal scenario from the current basic plan for high-level radioactive waste management, an improved disposal system has been proposed that enhances disposal efficiency and economic effectiveness compared to the existing disposal system. For this purpose, two disposal canisters concepts were derived from the length of the spent fuel generated from the nuclear power plants. In the disposal scenario, the acceptable amount of decay heat for each disposal container was determined, taking into account the discharge and disposal times of spent fuels in accordance with the current basic plan. Based on the determined decay heat of the two types of disposal canisters and the associated disposal system, thermal stability analyses were performed to confirm their suitability to the proposed disposal system design requirement and disposal efficiency assessment. The results of this study confirm 20% reduction in the disposal area and 20% increase in disposal density for the proposed disposal system compared to the existing system. These results can be used to establish a spent fuel management policy and to design a viable commercial disposal system.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.22 no.2
• /
• pp.211-217
• /
• 2024

The growing significance of sustainable energy technologies underscores the need for safe and efficient management of spent nuclear fuels (SNFs), particularly via deep geological disposal (DGD). DGD involves the long-term isolation of SNFs from the biosphere to ensure public safety and environmental protection, necessitating materials with high corrosion resistance for DGD canisters. This study investigated the feasibility of a Cu-Ni film, fabricated via additive manufacturing (AM), as a corrosion-resistant layer for DGD canister applications. A wire-fed AM technique was used to deposit a millimeter-scale Cu-Ni film onto a carbon steel (CS) substrate. Electrochemical analyses were conducted using aerated groundwater from the KAERI underground research tunnel (KURT) as an electrolyte with an NaCl additive to characterize the oxic corrosion behavior of the Cu-Ni film. The results demonstrated that the AM-fabricated Cu-Ni film exhibited enhanced corrosion resistance (manifested as lower corrosion current density and formation of a dense passive layer) in an NaCl-supplemented groundwater solution. Extensive investigations are necessary to elucidate microstructural performance, mechanical properties, and corrosion resistance in the presence of various corroding agents to simplify the implementation of this technology for DGD canisters.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.15 no.3
• /
• pp.199-206
• /
• 2017

A geological repository for the disposal of high-level radioactive waste is generally constructed in host rock at depths of 500~1,000 meters below the ground surface. A geological repository system consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. The buffer is indispensable to assure the disposal safety of high-level radioactive waste, and it can restrain the release of radionuclides and protect the canister from the inflow of groundwater. Since high temperature in a disposal canister is released to the surrounding buffer material, the thermal properties of the buffer material are very important in determining the entire disposal safety. Even though there have been many studies on thermal conductivity, there have been only few studies that have investigates the specific heat capacity of the bentonite buffer. Therefore, this paper presents a specific heat capacity prediction model for compacted Gyeongju bentonite buffer material, which is a Ca-bentonite produced in Korea. Specific heat capacity of the compacted bentonite buffer was measured using a dual probe method according to various degrees of saturation and dry density. A regression model to predict the specific heat capacity of the compacted bentonite buffer was suggested and fitted using 33 sets of data obtained by the dual probe method.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.3
• /
• pp.339-346
• /
• 2018

A geological repository system consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. The buffer is indispensable to assure the disposal safety of high-level radioactive waste. Since the heat generated from spent nuclear fuel in a disposal canister is released to the surrounding buffer materials, the thermal properties of the buffer material are very important in determining the entire disposal safety. Especially, since thermal expansion can cause thermal stress to the intact rock mass in the near-field, it is very important to evaluate thermal expansion characteristics of bentonite buffer materials. Therefore, this paper presents a thermal expansion coefficient prediction model of the Gyeongju bentonite buffer materials which is a Ca-bentonite produced in South Korea. The linear thermal expansion coefficient was measured considering heating rate, dry density and temperature variation using dilatometer equipment. Thermal expansion coefficient values of the Gyeongju bentonite buffer materials were $4.0{\sim}6.0{\times}10^{-6}/^{\circ}C$. Based on the experimental results, a non-linear regression model to predict the thermal expansion coefficient was suggested and fitted according to the dry density.

• Tunnel and Underground Space
• /
• v.32 no.2
• /
• pp.107-119
• /
• 2022

Excavation-Disturbed Zone (EDZ) is an important design factor in constructing final disposal facilities for spent nuclear fuel, since EDZ affects mechanical stability including a spacing between disposal holes, and the hydraulic properties within EDZ plays a significant role in estimating in-flow rate of groundwater as well as a subsequent corrosion rate of a canister. Thus, it is highly required to characterize in-situ EDZ with precision and control the EDZ occurrence while excavating disposal facilities and constructing relevant underground research facilities. In this report, we not only reviewed EDZ-related researches carried out in the ONKALO facility of Finland but also examined appropriate methods for field inspection and quality control of EDZ occurrence. From the review, GPR can be the most efficient method for in-situ characterization of EDZ since it does not demand drilling a borehole that may disturb a surrounding environment of caverns. And the EDZ occurrence was dominant at a cavern floor and it ranged from 0 to 70 cm. These can provide useful information in developing necessary EDZ-related regulations for domestic disposal facilities.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.22 no.1
• /
• pp.9-16
• /
• 2024

The bentonite buffer material is a crucial component in an engineered barrier system used for the disposal of high-level radioactive waste. Because a large amount of heat from the disposal canister is released into the bentonite buffer material, the thermal conductivity of the bentonite buffer is a crucial parameter that determines the design temperature. At the Korea Atomic Energy Research Institute (KAERI), a new standard bentonite (Bentonil-WRK) has been used since 2022 because Gyeongju (KJ) bentonite is no longer produced. However, the currently available data are insufficient, making it essential to investigate both the basic and complex properties of Bentonil-WRK. Thus, this study evaluated its geotechnical and thermal properties and developed a thermal conductivity empirical model that considers its dry density, water content, and temperature variations from room temperature to 90℃. The coefficient of determination (R²) for the model was found to be 0.986. The thermal conductivity values of Bentonil-WRK were 1-10% lower than those of KJ bentonite and 10-40% higher than those of MX-80 bentonites, which were attributable to mineral-composition differences. The thermal conductivity of Bentonil-WRK ranged between 0.504 and 1.149 W·(m^-1·K^-1), while the specific heat capacity varied from 0.826 to 1.138 (kJ·(kg^-1·K^-1)).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.3
• /
• pp.313-319
• /
• 2019

The A-KRS (Advanced Korean Reference Disposal System) is the disposal concept for pyroprocessed waste, which has been developed by the Korea Atomic Energy Research Institute. In this disposal concept, the amount of high-level radioactive waste is minimized using pyrochemical process, called pyroprocessing. The produced pyroprocessed waste is then solidified in the form of monazite ceramic. The final product of ceramic wastes will be disposed of in a deep geological repository. By the way, the decay heat is generated due to the radioactive decay of fission products and raises the temperature of buffer materials in the near field of radioactive waste repository. However, the buffer temperature must be kept below $100^{\circ}C$ according to the safety regulation. Usually, the temperature can be controlled by variation of the canister interdistance. However, KAERI has modelled thermal analysis under the boundary condition, where the waste canisters are in direct contact with each other. Therefore, a reliable temperature analysis in the disposal system may fail because of unknown thermal resistence values caused by the spatial gap between waste canisters. In the present work, we have performed thermal analyses considering the gap between heating elements and canisters at the beginning of canister loading into the radioactive waste repository. All thermal analyses were performed using the COMSOL software package.

• Tunnel and Underground Space
• /
• v.13 no.2
• /
• pp.153-168
• /
• 2003

The objective of this present study is to understand long term(500 years) thermohydromechanical interaction behavior in the vicinity of a repository cavern on the joint location and repository depth variations. The model includes a saturated discontinuous granitic rock mass, PWR spent nuclear fuel in a disposal canister surrounded with compacted bentonite inside a deposition hole, and mixed bentonite backfilled in the rest of the space within a repository cavern. It is assumed that two joint sets exist within the model. Joint set 1 includes joints of 56$^{\circ}$ dip angle, spaced at 20 m, and joint set 2 is in the perpendicular direction to joint set 1 and includes joints of 34$^{\circ}$ dip angle, spaced at 20 m. In order to understand the behavior change on the joint location variations, 5 different models of 500m in depth are analyzed, and additional 3 different models of 1000 m in depth are analyzed to understand the effect of depth variation.