• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.2
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• pp.135-142
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• 2010

To obtain the input data for the design and long-term performance assessment of a high-level waste repository, the thermal conductivities of several granite rocks which were taken from the rock cores from the declined borehole were measured. The thermal conductivities of granite were measured under the different conditions of water content to investigate the effects of the water content on the thermal conductivity. A simple empirical correlation was proposed to predict the thermal conductivity of granite as a function of effective porosity and water content which can be measured with relative ease while neglecting the possible effects of mineralogy, structure and anisotropy. The correlation could predict the thermal conductivity of granite with the effective porosity below 2.7% from the KURT site with an estimated error below 10%.

• The Journal of Engineering Geology
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• v.9 no.2
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• pp.89-100
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• 1999

This study was carried out to assess the physicochemical properties of domestic bentonites and zeolites from Tertiary Formation as the candidate material for a engineered barrier of a radioactive waste repository. Natural bentonite and zeolite samples were collected from nine bentonite mines and six zeolite mines in Yeonil-Gampo area. The commercial products of bentonite and zeolite were obtained from local companies. The collected samples were investigated to study the following physicochemical properties: X-ray diffraction patterns, swelling, cation exchange capacity(CEC), specific surface area, montmorillonite content, pH, organic carbon content, thermal property, microstruciure and chemical composition. Based on the physicochemical properties of bentonite and zeolite, the bentonites from U-41 and G-46 mines and the zeolites from Daedo and Y-1 mines are regarded as the most desirable candidate materials.

• Nuclear Engineering and Technology
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• v.40 no.7
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• pp.593-606
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• 2008

In this study, data on radionuclide migration and retardation processes in the engineered and natural barriers of High-Level Radioactive Waste (HLW) repository have been reviewed and compiled for use in the performance assessment of a HLW disposal system in Korea. The status of the database on radionuclide migration and retardation that is being developed in Korea is investigated and summarized in this study. The solubilities of major actinides such as D, Th, Am, Np, and Pu both in Korean bentonite porewater and in deep Korean groundwater are calculated by using the geochemical code PHREEQC (Ver. 2.0) based on the KAERI-TDB(Korea Atomic Energy Research Institute-Thermochemical Database), which is under development. Databases for the diffusion coefficients ($D^b_e$ values) and distribution coefficients ($K^b_d$ values) of some radionuclides in the compacted Korean Ca-bentonite are developed based upon domestic experimental results. Databases for the rock matrix diffusion coefficients ($D^r_e$ values) and distribution coefficients ($K^r_d$ values) of some radionuclides for Korean granite rock and deep groundwater are also developed based upon domestic experimental results. Finally, data related to colloids such as the characteristics of natural groundwater colloids and the pseudo-colloid formation constants ($K_{pc}$ values) are provided for the consideration of colloid effects in the performance assessment.

• Tunnel and Underground Space
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• v.32 no.6
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• pp.397-410
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• 2022

The safety of the disposal repository for high level radioactive waste should be guaranteed for a quite long period so that the precise evaluations are required. The site characteristics of the discontinuities are essential part of the safe repository design including engineered barrier and natural barrier systems. The discontinuities act like weak planes and at the same time, they act as flow paths so that their features should be investigated thoroughly. RQD (Rock Quality Designation) is one of the most widely applied characterizing methods due to its simplicity, however, modified designations have been proposed because RQD has some drawbacks, such as its directivity and dependence on the threshold length. This study aims to evaluate the applicability of the modified designations by applying them to the rock mass around KURT and to produce fundamental database that will be utilized in future studies.

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

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.65-72
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• 2008

The purpose of the HLW deep geological disposal is to isolate and to delay the radioactive material release to human beings and the environment for a long time so that the toxicity does not affect to the environment. The main requirements for the HLW repository design is to keep the buffer temperature below $100\;^{\circ}C$ in order to maintain its integrity. So the cooling time of spent fuels discharged from the nuclear power plant is the key consideration factors for efficiency and economic feasibility of the repository. The disposal tunnel/disposal hole spacing, the disposal area and thermal capacity required for the deep geological repository layout which satisfies the temperature requirement of the disposal system is analyzed to set the optimized spent fuels cooling time. To do this, based on the reference disposal concept, thermal stability analyses of the disposal system have been performed and the derived results have been compared by setting the spent fuels cooling time and the disposal tunnel/disposal hole spacing in various ways. From these results, desirable spent fuels cooling time in view of disposal area is derived. The results shows that the time reaching the maximum temperature within the design limit of the temperature in the disposal site is likely shortened as the cooling time of spent fuels becomes short. Also it seems that the temperature-rising and-dropping patterns in the disposal site are of smoothly varying form as the cooling time of spent fuels becomes long. In addition, it is revealed that a desirable cooling time of spent fuels is approximately 40-50 years when spent fuels are supposedly disposed in the deep geological disposal site with its structural scale under consideration in this study.

• Tunnel and Underground Space
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• v.31 no.4
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• pp.289-308
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• 2021

This study conducts coupled thermo-hydro-mechanical numerical modeling to investigate the maximum temperature and conditions for securing mechanical stability of the high-level radioactive waste repository when temperature criteria of bentonite buffer are 100℃ and 125℃, respectively. In case of temperature criterion of buffer as 100℃, the maximum temperatures at the interface between canister and buffer are calculated to be 99.4℃ and 99.8℃, respectively for a case with disposal tunnel spacing of 40 m and deposition hole spacing of 5.5 m and for the other case with disposal tunnel spacing of 30 m and deposition hole spacing of 6.5 m. In case of temperature criterion of buffer as 125℃, spacings of disposal tunnel and deposition hole could be decreased to 30 m and 4.5 m, respectively, which reduces the disposal area up to 55% compared to the disposal area of KRS⁺. According to analysis of mechanical stability for various disposal spacings, RMR of rock mass for KRS⁺ should be larger than 72.4 which belongs to good rock in RMR classification to prevent failure of rock mass. As disposal spacing is decreased, required RMR of rock mass is increased. In order to prevent failure of rock mass for a case with disposal tunnel spacing of 30 m and deposition hole spacing of 4.5 m, RMR larger than 87.3 is needed. However, mechanical stability of the repository is secured for all cases with RMR over 75 considering the enhancement of rock strength due to confining stress induced by swelling of the bentonite buffer and backfill.

• Geophysics and Geophysical Exploration
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• v.7 no.1
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• pp.70-79
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• 2004

We have examined the applicability of the triaxial drill-bit VSP method (TAD-VSP) to the geological survey of possible sites for a high-level radioactive waste disposal repository. The seismic energy generated by a drill bit is measured by a downhole multi-component detector, and the resulting signals are processed to image the geological structure deep underground. In order to apply the TAD-VSP method to civil-engineering-scale drilling, we have developed a small but highly sensitive and precise three-component downhole seismic measurement system, and recorded drill-bit signals at a granite quarry. We have successfully imaged discontinuities in the granite, possibly related to fractures, as highly reflective zones. The discontinuities imaged by the TAD-VSP method correlate well with the results of other borehole observations. In conclusion, the TAD-VSP method is usable in geological investigations for civil engineering because the equipment is compact and it is simple to acquire the drill-bit signal.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2604-2612
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• 2023

Spent nuclear fuel and long-lived radioactive waste must be carefully handled before disposing them off to a geological repository. After the pre-storage period in water pools, spent nuclear fuel is stored in casks, which are widely used for interim storage. Interim storage in casks is very important part in the whole cycle of nuclear energy generation. This paper presents the results of the numerical study that was performed to evaluate the thermal behavior of a metal-concrete CONSTOR RBMK-1500 cask loaded with spent nuclear fuel and placed in an open type interim storage facility which is under fire conditions (steady-state, fire, post-fire). The modelling was performed using the ANSYS Fluent code. Also, a local sensitivity analysis of thermal parameters on temperature variation was performed. The analysis demonstrated that the maximum increase in the fuel load temperatures is about 10 ℃ and 8 ℃ for 30 min 800 ℃ and 60 min 600 ℃ fires respectively. Therefore, during the fire and the post-fire periods, the fuel load temperatures did not exceed the 300 ℃ limiting temperature set for an RBMK SNF cladding for long-term storage. This ensures that fire accident does not cause overheating of fuel rods in a cask.