• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.3
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• pp.229-238
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• 2010

In order to acquire a realistic forecast for the lifetime and post-closure period of the LILW (Low- and Intermediate-Level Radioactive Waste) repository and to establish the overall management plan associated gas issues. it is essential to carry out the long-term experimental research in a similar condition to actual disposal environment. Regarding this, as a part of the following-up actions on a construction and operation license for the first stage of the LILW repository at Gyeongju city, a large-scale in-situ experiment is being planned. For securing basic data on the experiment, the experimental researches related to gas generation previously performed in foreign countries are reviewed in detail. Consequently, it is judged that data on the gas generation experiment in Finland could be practically applied as the benchmark for our large-scale in-situ experiment because the same disposal concept as the Korean repository is adopted and the experiment is performed in a scale large enough to allow the use of regular waste packages.

• Progress in Medical Physics
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• v.31 no.4
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• pp.194-204
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• 2020

This study investigated and analyzed the Korean and international status of radioactive waste management for medical linear accelerators (linacs) and proceed prior research to suggest radiation safety regulations and guidelines for the safe use of radiation. We analyzed the number of linacs installed in the radiation oncology departments of 103 institutions. In addition, we analyzed the procedures and standards for disposal in Korea and foreign countries. For foreign countries, we analyzed the status based on reports from the United States, Japan, Europe, and Canada. A total of 182 linacs are installed in Korea and 95% of them use more than 10 MV of energy. In Korea, standards for managing radioactive waste from a linac, disposal procedures, and clearance criteria have yet to be established. Therefore, radioactive waste is disposed of in different ways depending on the hospitals where they originate. Japan, the US, and Canada have recommended clearance levels and procedures for linacs. Other countries have provided management guidelines for research or large-scale accelerators, but not for medical purposes. In this study, we investigated the management of radioactive waste from medical linacs in Korea and abroad. Several foreign countries have suggested a clearance level and criteria for disposing of waste storage drums. For the safe management of medical linacs, it is necessary to establish safety management regulations. In Korea, standards for disposal, such as radiation or dose limits, are required for medical linacs. A system for clearance when disposing at a medical institution should be created.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.64-72
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• 2004

The dryer (ten per unit) are operating to remove tritium in PHWR(Pressurized Heavy Water Reactor). There are coming out heavy water adsorbent from operating the dryer (95 drums for ten year per unit) The amount of radioactivity of heavy water adsorbent almost exceed ninety times more than disposal limit-in-itself showed by The Ministry of Science and Technology. It has to be disposed whole radioactive waste products, however there are problems of increase at the expense of their permanent disposal. In this research, We have studied how to remove kinds of nuclear materials and amount of tritium with in heavy water adsorbent. As the result we could develop disposal equipment and apply it. D20 adsorbent have to contain below Gamma nuclide O.3Bq/g and tritium 100Bq/g "The Regulation for disposal of the radioactivity wastes" showed by The Ministry of Science and Technology. There fore. So as to remove amount of tritium and kinds of nuclear materials (DTO) we needed a equipment. Also we have studied how to remove effectively radioactivity with in Adsorbent. As cleaning heavy water adsorbent and drying on each condition (temperature for drying and hours for cleaning). Because there is something to return heavy water adsorbent by removing impurities within adsorbent when it is dried o high temperature. After operating, we have been applying this research 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 Environment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.4
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• pp.387-398
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• 2008

To obtain base-line data in the low- and intermediate-level radioactive waste disposal facility, radiological environment investigation is required at least two years prior to the facility operation near the disposal facility. The investigation has been performed since January 2007 for the Wolsong LILW disposal center. In this paper, investigation procedure and detailed information obtained for six months in 2008 are described. Based on the current results, future investigation planning is also discussed for the radiological environment management of the disposal facility.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.219-230
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• 2017

This study was conducted to predict and evaluate the uncertainty of safety after closure of the second phase surface disposal facility of the Gyeongju intermediate and low level repository in Korea. In this study, four scenarios are developed considering both intact and degraded states of multi-layered covers and disposal containers; also, the fluid flow by a rainfall into the disposal facility is simulated. The rainfall conditions are implemented based on the monthly average data of the past 30 years (1985~2014); the simulation period is 300 years, the management period regulated by institutional provisions. As a result of the evaluation of the basic scenario, in which the integrity of both of the containers and the covers is maintained, it was confirmed that penetration of rainfall does not completely saturate the inside of the disposal facility. It is revealed that the multiple cover layers and concrete containers effectively play the role of barrier against the permeation of rainfall.

• Journal of the Korean Geotechnical Society
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• v.33 no.7
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• pp.55-64
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• 2017

A geological repository has been considered one of the most adequate options for the disposal of high-level radioactive waste. A geological repository will be constructed in a host rock at a depth of 500~1,000 meters below the ground surface. The geological repository system consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. The buffer is very important to assure the disposal safety of high-level radioactive waste. It can restrain the release of radionuclide and protect the canister from the inflow of groundwater. High temperature in a disposal canister is released into the surrounding buffer material, and thus the thermal transfer behavior of the buffer material is very important to analyze the entire disposal safety. Therefore, this paper presents a thermal conductivity prediction model for the Kyungju compacted bentonite buffer material which is the only bentonite produced in Korea. Thermal conductivity of Kyungju bentonite was measured using a hot wire method according to various water contents and dry densities. With 39 data obtained by the hot wire method, a regression model to predict the thermal conductivity of Kyungju bentonite was suggested.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3359-3366
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• 2021

An engineered barrier system (EBS) for the deep geological disposal of high-level radioactive waste (HLW) is composed of a disposal canister, buffer material, gap-filling material, and backfill material. As the buffer fills the empty space between the disposal canisters and the near-field rock mass, heat energy from the canisters is released to the surrounding buffer material. It is vital that this heat energy is rapidly dissipated to the near-field rock mass, and thus the thermal conductivity of the buffer is a key parameter to consider when evaluating the safety of the overall disposal system. Therefore, to take into consideration the sizeable amount of heat being released from such canisters, this study investigated the thermal conductivity of Korean compacted bentonites and its variation within a temperature range of 25 ℃ to 80-90 ℃. As a result, thermal conductivity increased by 5-20% as the temperature increased. Furthermore, temperature had a greater effect under higher degrees of saturation and a lower impact under higher dry densities. This study also conducted a regression analysis with 147 sets of data to estimate the thermal conductivity of the compacted bentonite considering the initial dry density, water content, and variations in temperature. Furthermore, the Kriging method was adopted to establish an uncertainty metamodel of thermal conductivity to verify the regression model. The R² value of the regression model was 0.925, and the regression model and metamodel showed similar results.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.261-265
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• 2003

This paper proposes acceptance requirements and the corresponding implementation strategy for the near-surface repository of low- and intermediate-level radioactive wastes. The proposed requirements include details on waste classification and radionuclide concentration limitation and inventory analysis. The strategy considers a relevant linking between predisposal waste management, disposal safety assessment and the overall disposal system.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2021.05a
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• pp.191-192
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• 2021

For the stable management of radioactive waste, it is necessary to secure a solidification treatment technology capable of immobilizing hazardous radioactive elements in a solid matrix. In this study, the feasibility of using recycled cement recovered from waste concrete as a solidifying material for radioactive waste was analyzed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.439-455
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• 2020

Numerical model was developed that simulates radionuclide (3H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.