• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.41-52
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• 2013

Adequate design of engineered barriers, including canister, buffer and backfill, is important for the safe disposal of high-level radioactive waste. Three-dimensional computer simulations were carried out under different condition to examine the thermal and mechanical behavior of engineered barriers and rock mass. The research looked at five areas of importance, the effect of the swelling pressure, water content of buffer, density of compacted bentonite, emplacement type and the selection of failure criteria. The results highlighted the need to consider tensile stress in the outer shell of a canister due to thermal expansion of the canister and the swelling pressure from the buffer for a more reliable design of an underground repository system. In addition, an adequate failure criterion should be used for the buffer and backfill.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.2
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• pp.105-112
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• 2004

A waste container, one of the key components of a multi-barrier system in a potential high level radioactive waste (HLW) repository in Korea ensures the mechanical stability against the lithostatic pressure of a deep geologic medium and the swelling pressure of the bentonite buffer. Also, it delays potential release of radionuclides for a certain period of time, before it is corroded by intruding impurities. Even though the material of a waste container is carefully chosen and its manufacturing processes are under quality assurance processes, there is a possibility of initial defects in a waste container during manufacturing. Also, during the deposition of a waste container in a repository, there is a chance of an incident affecting the integrity of a waste container. In this study, the appropriate Features, Events, and Processes(FEP's) to describe these incidents and the associated scenario on radionuclide release from a container to the biosphere are developed. Then the total system performance assessment on the Initial waste Container Failure (ICF) scenario was carried out by the MASCOT-K, one of the probabilistic safety assessment tools KAERI has developed. Results show that for the data set used in this paper, the annual individual dose for the ICF scenario meets the Korean regulation on the post closure radiological safety of a repository.

• 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.

• Journal of the Korean Geotechnical Society
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• v.37 no.10
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• pp.5-11
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• 2021

The compacted bentonite buffer is one of the most important components in an engineered barrier system (EBS) to dispose of high-level radioactive waste (HLW) produced by nuclear power generation. The compacted bentonite buffer has a crucial role in protecting the disposal canister against the external impact and penetration of groundwater, so it has to satisfy the thermal-hydraulic-mechanical requirements. Even though there have been various researches on the investigation of thermal-hydraulic properties, few studies have been conducted to evaluate mechanical properties for the compacted bentonite buffer. For this reason, this paper conducted a series of unconfined compression tests and obtained mechanical properties such as unconfined compressive strength, elastic modulus, and void ratio of Korean compacted bentonite specimens with different water content and dry density values. The unconfined compressive strength and elastic modulus increased, and the Poisson's ratio decreased a little with increasing dry density. It showed that unconfined compressive strength and elastic modulus were proportional to dry density. However, there was not a remarkable correlation between mechanical properties and water content.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.43-49
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• 2020

An engineered barrier system (EBS) for the geological disposal of high-level radioactive waste (HLW) consists of a disposal canister packed with spent fuel, buffer material, backfill material, and gap-filling material. The buffer material fills the space between the canister and the near-field rock, thus serving to restrain the release of radionuclides and protect the canister from groundwater penetration. Furthermore, as significant amounts of heat energy are released from the canister to the surrounding rock, the thermal conductivity of the buffer plays an important role in maintaining the safety of the entire disposal system. Therefore, given the high levels of heat released from disposal canisters, this study measured the thermal conductivities of compacted bentonite buffers from Gyeongju under temperature variations ranging 25 to 80~90℃. There was a 5~20% increase in thermal conductivity as the temperature increased, and the temperature effect increased as the degree of saturation increased.

• 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.

• 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.

• Journal of the Korean Geotechnical Society
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• v.34 no.1
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• pp.17-24
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• 2018

A buffer in a geological disposal system minimizes groundwater inflow from the surrounding rock and protects the disposed high-level waste (HLW) against any mechanical impact. As decay heat of a spent fuel causes temperature variation in the buffer that affects the mechanical performance of the system, an accurate estimation of the temperature variation is substantial. The temperature variation is affected by thermal and material properties of the system such as thermal conductivity, density and specific heat capacity of the buffer, and thus these factors should be properly included in the design of the system. In particular, as the thermal properties are variable depending on the density and water content of the buffer, consideration of the effects should be included in the analysis. Hence, in this study, a numerical model based on finite element method (FEM) which is able to consider the change of density and water content of the buffer was established. In addition, using the numerical model, a parametric study was conducted to investigate the effect of each thermal property on the temperature variation of the buffer.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.219-228
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• 2012

A long-term R&D program for HLW disposal technology development was launched in 1997 in Korea and Korea Reference disposal System(KRS) for spent fuels had been developed. After then, a recycling process for PWR spent fuels to get the reusable material such as uranium or TRU and to reduce the volume of radioactive waste, called Pyro-process, is being developed. This Pyro-process produces several kinds of wastes including metal waste and ceramic waste. In this study, the characteristics of the waste from Pyro-process and the concepts of a disposal container for the wastes were described. Based on these concepts, thermal analyses were carried out to determine a layout of the disposal area of the ceramic wastes which was classified as a high level waste and to develop the disposal system called A-KRS. The location of the final repository for A-KRS is not determined yet, thus to review the potential repository domains, the possible layout in the geological characteristics of KURT facility site was proposed. These results will be used in developing a repository system design and in performing the safety assessment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.1
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• pp.123-136
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• 2018

The objective of this study is to propose a new methodology to fabricate a reliable engineering-scale buffer block, which shows homogeneous and uniform distribution in buffer block density, for in-situ experiments. In this study, for the first time in Korea, floating die press and CIP (Cold Isostatic Press) are applied for the manufacture of an engineering-scale bentonite buffer. The optimized condition and field applicability are also evaluated with respect to the method of manufacturing the buffer blocks. It is found that the standard deviation of the densities obtained decreases noticeably and that the average dry density increases slightly. In addition, buffer size is reduced by about 5% at the same time. Through the test production, it is indicated that the stress release phenomenon decreases after the application of the CIP method, which leads to a reduction in crack generation on the surface of the buffer blocks over time. Therefore, it is confirmed that the production of homogeneous buffer blocks on industrial scale is possible using the method suggested in this study, and that the produced blocks also meet the design conditions for dry density of buffer blocks in the AKRS (Advanced Korea Reference Disposal System of HLW).