• Journal of the Korean Society of Radiology
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• v.13 no.2
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• pp.297-303
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• 2019

The nuclear medicine department of a domestic medical institution uses $^{99m}TcI$, a radionuclide, from $^{99}Mo/^{99m}TcI$ Generator, to inject radioactive drugs into patients. Among the expired generators, imported from foreign countries, the medical institution implements its own disposal. Each medical institution shall satisfy the permitted in-house disposal concentration of radioactive wastes. The guidelines for self-disposal presented in Korea suggested that self-disposal can be performed 80 days after the generator is used. The purpose of these guidelines is to analyze them by comparing them with the data measured directly with the generator and to study if they are feasible. As a result, the generator with a capacity of 1,000 mCi has the longest half-life, and when tested with a high-radiation Mo(molybdenum) column, the number of days that are below the permitted concentration of body disposal with radioactive waste was 72 days and 71 days that were derived from direct column measurement. The results of the direct study confirmed that the guidelines for in-house disposal in Korea were reasonable, as there were 8 to 9 days of storage compared to the number of in-house disposal days provided in the guidelines.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.201-211
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• 2005

In this study, conventional head-end processes of spent TRISO fuel have been reviewed to develope more effective treatment methods. The main concerns in the TRISO treatment are to effectively separate the carbon and SiC contained in the TRISO particles. The crush-burn scheme which was considered in the early stages of the development has been replaced by the crush-leach process because of $^{14}C$ problems as a second waste being generated during the process. However, there are still many obstacles to overcome in the reported processes. Hence, innovative thermomechanical concepts and a molten salt electrochemical approach to breach the coating layers of the TRISO particle with a minimized amount of second waste are proposed in this paper and their principles are described in detail.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.268-277
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• 2004

process (ACP), electrochemical properties of high heat-generating alkali and alkali earth oxides in molten salt were measured and the behavior of those elements were analyzed. The reduction potentials of Cs, Sr, and Ba in a molten LiCl-$Li_2O$ system were more cathodic than that of Li and closely located one another. Thus, it is expected that the alkali and alkali earth would not hinder the reaction mechanism which is via lithium reduction. Alkali and alkali earth metals are likely to recycle into molten salt when the process is operated beyond metal reduction potentials and the effect of electric current on the mass transport is also determined by measuring the metal concentrations in the molten salt phase at different current conditions.

• Membrane Journal
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• v.7 no.4
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• pp.167-174
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• 1997

The concentration of silica is required to meet a certain level because silica affects fuel and materials integrity by forming a zeolite layer on fuel cladding surfaces. When the established Feed and Bleed method is employed, nuclear waste increase and the corresponding amount of boric acid is constantly consumed. This study concentrates on minimizing the amount of nuclear waste and consumption of boric acid. Using five different membranes, operating conditions such as temperatur, feed water flow rate, boric acid recovery and silica removal rate were examined. A silica-selective removal system was designed based on the above optimization procedures. Three-stage system was designed with two characteristically different membranes so that it could correspond with the different situation easily. Compared to the pevious results of the Feed and Bleed method, the current method showed that the amount of nuclear waste was reduced to 7%, and the consumption of boric acid to 15.7%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.285-299
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• 2006

At the end of 2005, 443 nuclear reactors were operating in 32 countries worldwide. They had provided about 3,000 TWh, which was just over 16 percent of global electricity supply. With the generating capacity of 368 GWe in 2004, the spent fuel generation rate worldwide, now becomes at about 11,000 tHM/y. Projections indicate that cumulative amounts to be generated by the year 2020, the time when most of the existing NPP will be closed to the end of their licensed lifetime, may be close to 445,000 tHM. In this regard, spent fuel management is a common issue in all countries with nuclear reactors. Whatever their national policy and/or strategy is selected for the backend of the nuclear fuel cycle, the management of spent fuel will contribute an impending and imminent issues to be resolved in the foreseeable future. The 2nd Review Meeting of the Contracting Parties to the Joint Convention was held in Vienna from 15 to 24 May 2006. The meeting gave an opportunity to exchange information on the national policy and strategy of spent fuel management of the Contracting Parties, to discuss their situations, prospects and the major factors influencing the national policies in this field and to identify the most important directions that national efforts and international co-operation in this area should be taken. In this paper, an overview of national and global trends of spent fuel management is discussed. In addition, some directions are identified and recent activities of each Member States in the subject area are summarized.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.1
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• pp.1-11
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• 2004

The hot cell facility for research activities related to the lithium reduction of spent fuel, which is designed to permit safe handling of source materials with radioactivity levels up to 1,385 TBq, is planned to be built. To meet this goal, the facility is designed to keep gamma and neutron radiation lower than the recommended dose-rate in normally occupied areas. The calculations peformed with QAD-CGGP and MCNP-4C are used to evaluate the proposed engineering design concepts that would provide acceptable dose-rates during a normal operation in hot cell facility. The maximum effective gamma dose-rates on the surfaces of the facility at operation area and at service area calculated by QAD-CGGP are estimated to be $2.10{\times}10^{-3}, 2.97{\times}10^{-3} and 1.01{\times}10{-1}$ mSv/h, respectively. And those calculated by MCNP-4C are $1.60{\times}10^{-3}, 2.99{\times}10^{-3} and 7.88{\times}10^{-2}$ mSv/h, respectively, The dose-rates contributed by neutrons are one order of magnitude less than that of gamma sources. Therefore, it is confirmed that the radiological design for hot cell facility satisfies the Korean criterion of 0.01 mSv/h for the operation area and 0.15 mSv/h for the service (maintenance) area.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.231-238
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• 2005

On the basis of flow resistance theory the conceptual model and related mathematical descriptions is proposed for resistance modeling of groundwater flow in CPM(continuum Porous medium), DFN(discrete fracture network) and fractured-porous medium. The proposed model is developed on the basis of finite volume method assuming steady-state, constant density groundwater flow. The basic approach of the method is to evaluate inter-block flow resistance values for a staggered grid arrangement, i.e. fluxes are stored at cell walls and scalars at cell centers. The balance of forces, i.e. the Darcy law, is utilized for each control volume centered around the point where the velocity component is stored. The transmissivity (or permeability) at the interface is assumed to be the harmonic average of neighboring blocks. Flow resistance theory was utilized to relate the fluxes between the grid blocks with residual pressures. The flow within porous medium is described by three dimensional equations and that within an individual fracture is described by a two dimensional equivalent of the flow equations for a porous medium. Newly proposed models would contribute to develop flow simulation techniques with various matrix characteristics.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.2
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• pp.147-154
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• 2008

It is expected that the temporary storage facilities at the nuclear power plants will be full of the spent fuels within 10 years. Provided that a centralized interim storage facility is constructed along the coast of the Korean peninsula to solve this problem, a substantial amount of spent fuels should be transported by sea or by land every year. In this paper we developed a computer program for the analysis of transportation logistics of the spent fuels from 4 different nuclear power plant sites to the hypothetical centralized interim storage facility and the final repository. Mass balance equations were used to analyze the logistics between the nuclear power plants and the interim storage facility. To this end a computer program, CASK, was developed by using the VISUAL BASIC language. The annual transportation rates of spent fuels from the four nuclear power plant sites were determined by using the CASK program. The parameter study with the program illustrated the easiness of logistics analysis. The program could be used for the cost analysis of the spent fuel transportation as well.

• Economic and Environmental Geology
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• v.55 no.3
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• pp.219-229
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• 2022

Bentonite is being considered as a candidate for buffer material in geological disposal systems for high-level radioactive wastes. In this study, the effect of cement-bentonite interactions on bentonite alteration was investigated by reviewing the literature on studies of cement-bentonite interactions. The major bentonite alteration by hyperalkaline fluids produced by the interaction of cementitious materials with groundwater includes cation exchange, montmorillonite dissolution, secondary mineral precipitation, and illitization. When the hyperalkaline leachate from the reaction of the cementitious material with the groundwater comes into contact with bentonite, montmorillonite, the main component of bentonite, is dissolved and a small amount of secondary minerals such as zeolite, calcium silicate hydrate, and calcite is produced. When montmorillonite is continuously dissolved, the physicochemical properties of bentonite may change, which may ultimately causes changes in bentonite performance as a buffer material such as adsorption capacity, swelling capacity, and hydraulic conductivity. In addition, the bentonite alteration is affected by various factors such as temperature, reaction period, pressure, composition of pore water, bentonite constituent minerals, chemical composition of montmorillonite, and types of interlayer cations. This study can be used as basic information for the long-term stability verification study of the buffer material in the geological disposal system for high-level radioactive wastes.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.137-144
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• 2005

Silica in primary system combines with an alkali grammatical particle metal and forms the zeolite layer which is hindering the heat transfer on the surface of the cladding. Zeolite layer becomes the cause of the damage in this way. The problems of the NPP's primary system have been issued steadily by EPRI. Through a series of experiments of the laboratory scale, we confirmed the applicability of NF membrane for silica removal, as silica rejection rate of NF membrane is about $60\;{\sim}\;70\%$ and boron rejection rate is about $10\;{\sim}\;20\%$. We accomplished a site experiment about four NF membranes manufactured by FilmTec and Osmonics Inc. In experiment using 400L of SFP water, when operation pressure is $10kg_{f}/cm^2$, we confirmed that the silica rejection rate of NF90-2540 manufactured by FilmTec Inc. is about $98\%$, boron rejection rate is about $43\%$. The silica rejection rate of NF270-2540 is about $38\%$, boron rejection rate is about $3.5\%$. Afterward, through additional experiments, such as long term characteristic experiments, we are going to design a optimum NF membrane system for silica removal.