• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.201-205
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• 2010

Most of iodine was captured by the block when NaI solution flowed through a bentonite block sorbed silver to retard the migration of iodine released from high-level radioactive wastes. In order to understand in detail the mechanism for the retardation of the iodine by the silver ion, X-ray Absorption Near Edge Structure (XANES) and Extended X-ray Absorption Fine Structure (EXAFS) spectra of the silver sorbed bentonite before and after the contact with iodide were compared with those of AgO, $Ag_2O$ and AgI as references. This examination suggests that the silver ion sorbed on the bentonite is desorbed, and then it retards the migration of iodine by forming the cluster of AgI precipitate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.137-164
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• 2018

The guarantee of the performance of the engineered barriers in a geological repository is very important for the long-term safety of disposal as well as the efficient design of the repository. Therefore, the performance of the engineered barriers under repository condition should be demonstrated by in-situ experiments conducted in an underground research laboratory. This article provides a review of the major in-situ experiments that have been carried out over the past several decades at underground research laboratories around the world to validate the performance of engineered barriers of a repository, as well as their results. In-situ experiments to study the coupled thermal-hydraulic-mechanical behavior of the engineered barrier system used to simulate the post-closure performance of the repository are analyzed as a priority. In addition, in-situ experiments to investigate the performance of the buffer material under a real repository environment have been reviewed. State-of-the art in-situ validations of the buffer-concrete interaction, and the installation of the buffer, backfill and plug, as well as characterization of the near-field rock and the corrosion of the canister materials are, also performed.

• Proceedings of the Korean Information Science Society Conference
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• 2004.04b
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• pp.541-543
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• 2004

능동적 학습(active learning)은 제한된 시간과 인력으로 가능한 정확도가 높은 분류기(classifier)를 생성하기 위하여, 훈련집합에 추가할 예제 즉 문의예제(query example)의 선정과 확장된 훈련집합으로 다시 학습하는 과정을 반복하여 수행한다. 능동적 학습의 핵심은 사용자에게 카테고리(category) 부여를 요청할 문의예제를 선정하는 과정에 있다. 효과적인 문의예제를 선정하기 위하여 다양한 방안들이 제안되었으나, 이들은 매 문의단계마다 하나의 문의예제를 선정하는 경우에 가장 적합하도록 고안되었다. 능동적 학습이 복수의 예제를 사용자에게 문의할 수 있다면, 사용자는 문의예제들을 서로 비교해 가면서 작업할 수 있으므로 카테고리 부여작업을 보다 빠르고 정확하게 수행할 수 있을 것이다. 또한 충분한 인력을 보유한 상황에서는, 카테고리 부여작업을 병렬로 처리할 수 있어 전반적인 학습시간의 단축에 큰 도움이 될 것이다. 하지만, 각 예제의 문의예제로써의 적합 정도를 추정하면 유사한 예제들은 서로 비슷한 수준으로 평가되므로, 기존의 방안들을 복수의 문의예제 선정작업에 그대로 적용할 경우, 유사한 예제들이 문의예제로 동시에 선정되어 능동적 학습의 효율이 저하되는 현상이 나타날 수 있다. 본 논문에서는 특정 예제를 문의예제로 선정하면 이와 일정 수준이상 유사한 예제들은 해당 예제와 함께 문의예제로 선정하지 않음으로써, 이러한 문제점을 극복할 수 있는 방안을 제안한다. 제안한 방안을 문서분류 문제에 적용해 본 결과 기존 문의예제 선정방안으로 복수 문의예제를 선정할 때 발생할 수 있는 문제점을 상당히 완화시킬 있을 뿐 아니라, 복수의 문의예제를 선정하더라도 각 문의 단계마다 하나의 예제를 선정하는 경우에 비해 큰 성능의 저하가 없음을 실험적으로 확인하였다./$m\ell$로 나타났다.TEX>${HCO_3}^-$ 이온의 탈착은 서서히 진행되었다. R＆D investment increases are directly not liked to R＆D productivities because of delays and side effects during transition periods between different stages of technology development. Thus, It is necessary to develope strategies in order to enhance efficiency of technological development process by perceiving the switching pattern. 기여할 수 있을 것으로 기대된다. 것이다.'ity, and warm water discharges from a power plant, etc.h to the way to dispose heavy water adsorbent. Through this we could reduce solid waste products and the expense of permanent disposal of radioactive waste products and also we could contribute nuclear power plant run safely. According to the result we could keep the best condition of radiation safety super vision and we could help people believe in safety with Radioactivity wastes control for harmony with Environ

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.479-490
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• 2018

The granite melting concept, which was suggested by Gibb's group for the closing of a deep borehole, was experimentally checked for KURT granite. The granite melting experiments were performed in two pressure conditions of atmospheric melting with certain inorganic additives and high pressure melting formed by water vaporization. The results of atmospheric tests showed that KURT granite started to melt at a lower temperature of $1,000^{\circ}C$ with NaOH addition and that needle shaped crystals were formed around partially melted crystals. In high pressure tests, vapor pressure was increased by adding water with maximum pressure of about 400 bars. KURT granite was partially melted at $1,000^{\circ}C$ when vapor pressure was low. However, it was not melted at vapor pressures higher than 200 bars. Therefore, it was determined that high pressure with a small amount of water vapor more effectively decreased the melting point of granite. Meanwhile, high temperature and high pressure vapor caused severe corrosion of the reactor wall.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.305-315
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• 2020

For geological disposal of radioactive wastes, a method was proposed to evaluate the radionuclide transport in the biosphere by calculating the elapsed time of nuclide migration. The radionuclides were supposed to be introduced from a natural barrier and reached a large surface water body following a groundwater flow in a shallow subsurface. The biosphere was defined as a shallow subsurface environment that included aquifers on a host rock. Using the proposed method, a calculation algorithm was established, and a computer code that implemented the algorithm was developed. The developed code was verified by comparing the simulation results of the simple cases with the results of the analytical solution and a public program, which has been widely used to evaluate the radiation dose using the radionuclide transport near the surface. A case study was constructed using the previous research for radionuclide transport from the hypothetical geological disposal repository. In the case study, the code calculated the mass discharge rate of radionuclide to a stream in the biosphere. Because the previous research only demonstrated the transport of radionuclides from the hypothetical repository to the host rock, the developed code in the present study could help identify the total transport of radionuclide along the complete pathway.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.295-303
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• 2012

Based on the data observed and analyzed on a groundwater flow system in the KURT (KAERI Underground Research Tunnel) site, the transport of radionuclides, which were assumed to be released at the supposed position, was calculated on the time-domain. A groundwater pathway from the release position to the surface was identified by simulating the groundwater flow model with the hydrogeological characteristics measured from the field tests in the KURT site. The elapsed time when the radionuclides moved through the pathway is evaluated using TDRW (Time Domain Random Walk) method for simulating the transport on the time-domain. Some retention mechanisms, such as radioactive decay, equilibrium sorption, and matrix diffusion, as well as the advection-dispersion were selected as the factors to influence on the elapsed time. From the simulation results, the effects of the sorption and matrix diffusion, determined by the properties of the radionuclides and underground media, on the transport of the radionuclides were analyzed and a decay chain of the radionuclides was also examined. The radionuclide ratio of the mass discharge into the surface environment to the mass released from the supposed repository did not exceed $10^{-3}$, and it decreased when the matrix diffusion were considered. The method used in this study could be used in preparing the data on radionuclide transport for a safety assessment of a geological disposal facility because the method could evaluate the travel time of the radionuclides considering the transport retention mechanism.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.107-119
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• 2014

The worldwide Status-of-Art survey for the in-situ experiments of the engineered barrier system for HLW underground disposal was performed as a preliminary action for the design of the in-situ demonstration at KURT. Some nations, which have executed or is ongoing the in-situ experiments at their underground research facilities, were summarized in this review. The demonstration projects reviewed were TBT/Sweden-France, LOT/Sweden, HE-E/Switzerland, PRACLAY/Belgium, FEBEX/Spain, HORONOBE/Japan, and BCE/Canada. The investigated items for the projects were mainly their purposes, constitutional structures, test conditions, monitoring parameters and the measuring tools, and test results. In this review, the hardware design and the assembling of the test system were more concentrated rather than their experimental results, because the purpose of this review is to achieve the necessary information for the practical design of the in-situ experiment to be installed at KURT. A mid scale in-situ demonstration of EBS at KURT, that is called IN-DBES, will be launched right after the completion the expanding project of KURT in 2015. It is hoped that the structural design, installing methods, hardware equipments required in the establishment of IN-DEBS will be referred on this review.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.1
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• pp.75-94
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• 2019

As the present single-layer repository concept requires too large an area for the site of the repository, a multi-layer repository concept has been suggested to improve the disposal density. The effects of the excavation damaged zone around the multi-layer repository constructed in the deep host rock on the temperature distribution in the repository were analyzed. For the thermal analysis of the multi-layer repository, the hydrothermal model was used to consider the resaturation process occurring in the buffer, backfill and rock. The existence of an excavation damaged zone has a significant effect on the temperature distribution in the repository, and the maximum peak temperatures of double-layer and triple-layer repositories can rise to $7^{\circ}C$ and $12^{\circ}C$, respectively depending on the size of the excavation damaged zone and the degree of decrease of the thermal conductivity. The dominant factor affecting the peak temperature in the multi-layer repository is the decrease of thermal conductivity in the excavation damaged zone, and the excavation damaged zone formed around the deposition hole has more significant effects on the peak temperature than does the excavation damaged zone formed around the disposal tunnel.

• Nuclear Engineering and Technology
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• v.16 no.2
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• pp.64-69
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• 1984

With regard to the radioactive waste disposal, adsorption properties and migration rates have been evaluated for Cs-137 and Sr-90 with the domestic clay sampled from Cnyang, Sanchong and Mooan. Sorption coefficients (Ksorp) were determined by batch experiments. The measured values of Ksorp were ranged from 8000 to 17,000 ml/gr for Cs-137 of 0.1$\mu$Ci/ml, and from 10,000 to 15,000m1/gr for Sr-90 of 0.l$\mu$Ci/ml. Remarkably, Mooan clay showed lower values of Ksorp than those of the others. This could be explained by the poor soprtion capacity of the quartz found only in the Mooan clay. For the quantitative analysis, sorption isotherm equations of Freundlich type were made with the obtained values of Ksorp. $C_{R}$=18.0 $C_{A}$$^{0.74}$ : Cs-137, $C_{R}$=0.84 $C_{A}$$^{0.45}$ : Sr-90. By introducing the BOX model combined with the above relationships, simulation of underground nuclide movement was carried out. The results showed that the domestic clays could be the effective backfill material for repositories.itories.ies.

• 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.