• Nuclear Engineering and Technology
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• v.6 no.2
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• pp.97-111
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• 1974

The adsorption mechanism of Cs-137 in low level radioactive solution by vermiculite treated with Na ion is studied in order to investigate its effective utilization for the radioactive effluent treatment. The beneficial role of Na-vermiculite is that Na ion can induce the wider c-axis spacing in which Cs ion can be sorbed in vermiculite. Cation exchange capacity and distribution coefficient of cesium seems to be influenced by the variation of c-axis spacing of vermiculite. Comparative identification and detection with the characteristic analyses of X-ray diffraction and electron diffraction patterns, diffrential thermal analysis and electron microscopy of Na-, K- and Cs-vermiculite are studied for the phemomena of Cs adsorption by vermiculite. This importance of the utilization in terms of adsorption and fixation of cesium involving vermiculite is discussed. It is found that the Na-vermiculite is valuable outside charging material for high level radioactive liquid waste storage tank of underground to protect the pollution of the underground water.

• 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.10 no.2
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• pp.117-123
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• 2012

In order to dispose of the LILW(low and intermediate level radioactive waste) stored at KAERI, the radwaste drum assay system will be introduced to evaluate the radioisotopes inventory of stored drums. At present, the construction project of the dedicated assay facility to operate it and carry out routine maintenance of that equipment has been conducting at the radwaste treatment facility. Since that facility will be constructed in front of a 1st radwaste storage facility as well as the radwaste drums to be assayed and the transmission source in the radwaste drum assay system are in that facility, they could act as the radioactive sources and then, would affect the dose rate at the inside and the outside of the facility. Therefore, the radiation shielding should be evaluated through the concrete wall near to the radioactive sources whether the wall thickness is sufficient against the regulations. In this study, the radiation safety for the concrete wall around the radiation controlled area in the radwaste drum assay facility was evaluated by the MCNP code. From the evaluation results, the thickness of those concrete walls which are under consideration of about 30 cm was enough to shield the radiation from the radioactive sources.

• Journal of the Korea Concrete Institute
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• v.25 no.3
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• pp.321-329
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• 2013

Recently, the medium-low level radioactive waste from nuclear power plant must be transported from temporary storage to the final repository. Medium-low level radioactive waste, which is composed mainly of the liquid ion exchange resin, has been consolidated with cementitious material in the plastic or iron container. Since cementitious material is brittle, it would generate cracks by impact load during transportation, signifying leakage of radioactive ray. In order to design the safety transporting equipment, there is a need to check the compressive strength of the current waste. However, because it is impossible to measure strength by direct method due to leakage of radioactive ray, we will estimate the strength indirectly by the dynamic modulus of elasticity. Therefore, it must be identified the relationship between of strength and dynamic modulus of elasticity. According to the waste acceptance criteria, the compressive strength of cement based solid is defined as more than 3.44 MPa (500 psi). Compressive strength of the present solid is likely to be significantly higher than this baseline because of continuous hydration of cement during long period. On this background, we have tried to produce the specimens of the 28 day's compressive strength of 3 to 30 MPa having the same material composition as the solid product for the medium-low level radioactive waste, and analyze the relationship between the strength and the dynamic modulus of elasticity. By controling the addition rates of AE agent, we made the mixture containing the ion exchange resin and showing the target compressive strength (3~30 MPa). The dynamic modulus of elasticity of this mixtures is 4.1~10.2 GPa, about 20 GPa lower in the equivalent compressive strength level than that of ordinary concrete, and increasing the discrepancy according to increase strength. The compressive strength and the dynamic modulus of elasticity show the liner relationship.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.453-462
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• 1995

In connection with the safe storage of high level nuclear waste, effect of $H_2O$$_2$ on the corrosion behavior of 304L stainless steel was examined. Open circuit potentials and polarization curves were measured with and without $H_2O$$_2$. The experimental results show that $H_2O$$_2$ increased corrosion potential and decreased pitting potential. The passive range, therefore, decreased as $H_2O$$_2$ concentration increased, indicating that pitting resistance was decreased by the existence of $H_2O$$_2$ in the electrolyte. These effect of $H_2O$$_2$ on corrosion of 304L stainless steel are considered to be similar to those of ${\gamma}$-irradiation. To compare the effects of $H_2O$$_2$ with those of $O_2$, cathodic and anodic polarization curves ore made in three types of electrolyte such as aerated, deaerated, and stirred electrolyte. The experimental results show that the effects of $H_2O$$_2$ on the corrosion behavior were tory similar to those of $O_2$ such as increase of corrosion potential, decrease of pitting resistance, and increase of repassivation potential. In acid and alkaline media, the corrosion potential shifts by $H_2O$$_2$ were restricted by the large current density of proton reduction and by the le Chatelier's principle respectively.y.

• The Korean Journal of Nuclear Medicine Technology
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• v.16 no.2
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• pp.126-130
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• 2012

Purpose : Methode an effective block was investigated to deal with volatile radioactive gas, short lived radioactive waste generated as a result of the routinely produced radiopharmaceuticals FDG (2-deoxy-2-[$^{18}F$]fluoro-D-glucose) and compound with $^{11}C$. Materials and Methods : All components of the radiation stack monitoring and data management system for continuous radioactive gas detection in the air extract system purchase from fixed noble gas monitor of Berthold company. TEDLAR gas sampling bags purchase from the Dongbanghitech company. TEDLAR gas sampling bags (volume: 10 L) connected via paraflex or PTFE tubing and Teflon 3 way stopcock. When installing TEDLAR gas sampling bags in Hot cell on the inside and not radioactive gas concentrations were compared. According to whether the Hot cell inside a activated carbon filter installed, compare the difference in concentration of the radioactive gas $^{18}F$. Comparison of radiation emission concentration difference of module a FASTlab and TRACElab. Results : Activated carbon filter are installed in the Hot cell, a measure of the concentration of radioactive gas was 8 $Bq/m^3$. Without activated carbone filter in the hot cell was 300 $Bq/m^3$. Tedlar bag prior to installation of the radioactive gases a measure of the concentration was 3,500 $Bq/m^3$, $^{11}C$ synthesis of the measured concentration was 27,000 $Bq/m^3$. After installed a Tedlar bag and a measure concentration of the radioactive gases was 300 $Bq/m^3$ and $^{11}C$ synthesis was 1,000$Bq/m^3$. Conclusion : $^{11}C$ radioactive gas that was ejected out of the Hot cell, with the use of a Tedlar gas sampling bag stored inside. A compound of 11C is not absorbed onto activated carbon filter. But can block the release out by storing in a Tedlar gas sampling bag. We was able to reduce the radiation exposure of the worker by efficient radiation protection.

• Journal of the Korean Geotechnical Society
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• v.21 no.4
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• pp.41-49
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• 2005

Recently, ground has been often exposed to high temperature environments such as chemical ground improvement, thermal energy storage system, and underground nuclear waste disposal system. Since the behavior of clay is sensitive to temperature change, the studies on the engineering properties of clay subjected to high temperature history may be important. This paper presents the mechanical behavior of clay with high temperature condition. $\bar{CU}$ tests using a high temperature and pressure triaxial compression test apparatus were carried out in order to investigate characteristics of deformation, shear strength, compression and consolidation of clay. During tests, the temperature was varied from $20^{\circ}C,\;50^{\circ}C,\;75^{\circ}C,\;80^{\circ}C\;to\;100^{\circ}C$.

• The Journal of Engineering Geology
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• v.8 no.3
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• pp.225-234
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• 1998

The behavior of the underground opening depends mainly on the magnitude of the initial stress existing before excavation and on the stress redistribution due to the excavation. In the case of elasto-plastic materials such as rock mass, as the structural behavior of surrounded opening due to excavation depends on the stress path, methods and sequence of excavation have influences on the results of numerical analysis. Therefore, in order to design underground openings with large cross-section such as underground nuclear power plants, radioactive waste disposal cavems, oil storage caverns, and so on more reasonably it is desirable to consider the effect of the excavation sequence in the analysis. In this paper, the underground structure is analyzed using the finite element method and the distinct element methods with a view to review the the effect of the excavation sequence. Based on the results of the analysis the followings are discussed : influence of excavation shape and sequence, effect of structural reinforcements, influence of multi caverns.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.3
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• pp.169-179
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• 2011

There are two types of nuclear reactors in Korea and they are PWR type and CANDU type. The safe management of the spent fuels from these reactors is very important factor to maintain the sustainable energy supply with nuclear power plant. In Korea, a reference disposal system for the spent fuels has been developed through a study on the direct disposal of the PWR and CANDU spent fuel. Recently, the research on the demonstration and the efficiency analyses of the disposal system has been performed to make the disposal system safer and more economic. PWR spent fuels which include a lot of reusable material can be considered being recycled and a study on the disposal of HLW from this recycling process is being performed. CANDU spent fuels are considered being disposed of directly in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System (KRS) which was to dispose of both PWR type and CANDU type, the more effective CANDU spent fuel disposal systems were developed. To do this, the disposal canister for CANDU spent fuels was modified to hold the storage basket for 60 bundles which is used in nuclear power plant. With these modified disposal canister concepts, the disposal concepts to meet the thermal requirement that the temperature of the buffer materials should not be over $100^{\circ}C$ were developed. These disposal concepts were reviewed and analyzed in terms of disposal effective factors which were thermal effectiveness, U-density, disposal area, excavation volume, material volume etc. and the most effective concept was proposed. The results of this study will be used in the development of various wastes disposal system together with the HLW wastes from the PWR spent fuel recycling process.

• Resources Recycling
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• v.31 no.5
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• pp.20-25
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• 2022

In this study, the applicability of microalgae was evaluated for eco-friendly decontamination of cesium-137 (Cs-137) and strontium-90 (Sr-90), which are radioactive nuclides contained in radioactive waste. The monolithic radioactive solution used in the experiment was manufactured at a concentration of 1.5 Bq/mL Cs-137 and 1.0 Bq/mL Sr-90 by diluting a standard radioactive solution and distilled water. This experiment used two types of microalgae, Chlorella Vulgaris was used for Sr-90 decontamination and Hematococcus pluvialis for Cs-137 decontamination. The experimental method is to put the microalgae cultured for 2 weeks into a bottle with a semi-permeable membrane, and then put the bottle in which the microalgae was put into the manufactured radioactive solution, so that the microalgae and the radioactive solution react through the semi-permeable membrane for 48 hours. For the radioactivity concentration analysis of each sample, a gamma-ray nuclide analyzer was used for Cs-137, a γ-ray isotope, and a Liquid Scintillation Count(LSC) was used f or Sr-90, a β-ray isotope. As a result of the experiment, it was confirmed that about 88.0 % of Cs-137 and about 89.7 % of Sr-90 could be decontaminated, and about 98.6 % of Sr-90 was finally able to be decontaminated by the two-stage decontamination method.