• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.153-164
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• 2014

Currently, there are many programs which are now being developed or already developed to predict radionuclide and corrosion product at the stage of designing NPP. However, since there are not many developments in evaluating quantity of activation corrosion products occurring when disassembling a nuclear power plant there exist some difficulties in calculating accurately. In order to evaluate activation products inventory for the research of effect of neutron activation in the reactor vessel, component of nuclear reactor and adjacent structures, it should be evaluated by using operation history of nuclear reactor, material composition of structure and average neutron flux in every field representing fixed structure of nuclear reactor. In this study, CORA, PACTOLE, CRUDSIM, CREAT and ACE codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor which is used at the stage of designing. As a future study, the accuracy in calculating the quantity of product corrosion can be increase by finding out the possibility of use and improvement for evaluation of the decontamination.

• Nuclear Medicine and Molecular Imaging
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• v.40 no.3
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• pp.148-154
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• 2006

Purpose: Although radionuclide cisternography (RNC) is an useful study to detect cerebrospinal fluid (CSF) leakage in the patient with spontaneous intracranial hypotension (SIH), it sometimes fails to demonstrate the site of CSF leakage. The aim of the study is to improve the detection of leakage site of CSF and to reduce time for the study in RNC using modified protocol (m-RNC). Materials & methods : The study consists of 8 studies of 7 patients ($38{\pm}8$ years, M:F=2:5) with SIH, who underwent m-RNC following administration of 185-222 MBq of $^{99m}Tc$-DTPA into the lumbar subarachnoid space. Sequential images were obtained the whole spine with the head including urinary bladder at 10 minute, 30 minute, 1 hour, 2 hour, 4 hour and 6 hour. Radioactivity of extradural space and urinary bladder was evaluated. Results: Leakage site of CSF was identified in all 8 cases by m-RNC. Leakage site was cervicothoracic junction (CTJ, n=3), CTJ with C1-2 (n=2), CTJ with thoracic spine, thoracolumbar spine and lumbar spine (each n=1). All cases presented leakage sites within 1 hour and multiple sites, where CTJ was included in 6 cases. Only one case presented additional site in 6 hour image. Early radioactivity within the urinary bladder was noted in 6 cases, but that was fellowing after identification of the leakage site. Conclusion: Radionuclide cisternography is sensitive to detect the leakage site of CSF and is expected to improve the detection of CSF leakage site and reduce time for the study using modified protocol.

• The Korean Journal of Nuclear Medicine
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• v.28 no.1
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• pp.62-68
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• 1994

This study was aimed to compare the density of the functional microcirculation of hepatocellular carcinoma (HCC) with normal liver and to investigate the effect of hepatic-arterial oily chemoembolization (HAE) by radionuclide examination. Methods : Eight patients with HCC proven by biopsy in five, and clinically and radiologically in three were included. The mixture of 2 cc normal saline with three to four mCi of $^{99m}Tc$-MAA was infused through a hepatic-arterial catheter for a minute. Dynamic images were obtained at a rate of 4 sec per frame for a minute, and static images and SPECT were followed. Results : In three patients who underwent hepatic arterial angiography (HAA) alone, radioactivity was markedly increased in tumors compared to the adjacent liver immediately after infusion of $^{99m}Tc$-MAA. The ratios of tumoral and extratumoral up-take (T/E ratio) were above 6.5 (range; $6.5{\sim}l9$, mean; 12.5). In four of the five patients who under-went superselective HAE, T/E ratio were remark-ably decreased ($0.5{\sim}1.3$). The areas of embolization were better delineated in radionuclide study than in postembolization HAA. In the other one who was considered to be embolized completely on HAA, strong radiouptake in the tumor was disclosed (T/E ratio; 7.0). Conclusions : Therefore hepatic-arterial flow study with radionuclide imaging using $^{99m}Tc$-MAA can be a valuable method to assess the accurate embolization effect in HCC.

• The Korean Journal of Nuclear Medicine
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• v.18 no.1
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• pp.9-17
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• 1984

The purpose of this study is to investigate significance of new method for quantification of left to right cardiac shunts using deconvolution analysis as an adjunct to gamma variate analysis in patients with bad bolus injection. In the present study, quantitative radionuclide angiocardiography (QRAC) was performed with and without deconvolution analysis (DA) in 37 patients with left to right shunt and 103 control patients without shunt. The results obtained were as follows; 1) The mean value of $Q_p/Q_s$ in 103 control patients was $1.10{\pm}0.12$ without DA, and was $1.01{\pm}0.03$ with DA. 2) Correlation(r) between oximetry and QRAC with DA was 0.87 and correlation(r) between oximetry and QRAC without DA was 0.61. 3) The 13 patients with left to right shunt, whose $Q_p/Q_s$ was greater than 3.0 by QRAC without DA, was studied by DA. Then the $Q_p/Q_s$ values were measurable in these 13 patients by DA and they showed significant correlation with oximetry. (r=0.68) The results indicate that deconvolution analysis technique for quantification of a left to right shunt provides more reliable and accurate shunt quantification, and reduces the influences of poor bolus injection.

• Journal of Korean Society of Environmental Engineers
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• v.38 no.5
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• pp.249-254
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• 2016

In this study, we investigated the main characteristics of off-gas generated from melting process and off-gas treatment system operation example to provide some primary data for commercial vitrification facility design. The purpose of vitrification facility operation is to treat hazardous materials in the radioactive wastes and harmful off-gas containing a variety of chemical species generated in the glass melting process. Constructing and operating vitrification facility essentially need to be licensed through safety analysis; it is very important to treat radionuclide and hazardous materials below the legal environment emissions regulation level. We must accurately understand the characteristics of off-gas and apply an appropriate off-gas treatment process accordingly. Thus, to design the appropriate off-gas treatment there must be a wide range of elements taken into account such as characteristics of waste and melter, regulation guidance of off-gas, characteristics of generated off-gas and off-gas treatment system performance assessment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.253-262
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• 2015

Radionuclide deposited on the surface of several internal and external systems in a nuclear power plant is created by the activation of corrosion products from nuclear reactor structural materials and fission products. Especially, the constant contact between water and the surface corrodes the inside where primary system makes coolants and corrosion products mixed. Also, these are circulated along the systems. For comparing models, CRUDTRAN, DISER, MIGA-RT and CPAIR codes are analyzed to predict the quantity of radionuclide and corrosion product of primary reactor that are used at the stage of designing. The corrosion products behavior of domestic PWR primary system was predicted by using CRUDTRAN. This study aims to increase the reliability of corrosion product evaluation model by comparing the actual values and calculated values with the data of a Westing House-type Nuclear Power Plant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.135-147
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• 2016

The buffer in geological disposal system is one of the major elements to restrain the release of radionuclide and to protect the container from the inflow of groundwater. The buffer material requires long-term stability, low hydraulic conductivity, low organic content, high retardation of radionuclide, high swelling pressure, and high thermal conductivity. These requirements could be determined by the quantitative analysis results. In case of South Korea, the bentonites produced in Gyeongju area have been regarded as candidate buffer/backfill materials at KAERI (Korea Atomic Energy Research Institute) since 1997. According to the study on several physical and chemical characteristics of domestic bentonite in the same district, this is the Ca-type bentonite with about 65% of montmorillonite content. Through this study, we present the criteria for the performance evaluation items and methods when collecting new buffer/backfill materials.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.847-856
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• 2014

Spent resin waste containing a high concentration of $^{14}C$ radionuclide cannot be disposed of directly. A fundamental study on selective $^{14}C$ stripping, especially from the IRN-150 mixed bed resin, was carried out. In single ion-exchange equilibrium isotherm experiments, the ion adsorption capacity of the fresh resin for non-radioactive $HCO_3{^-}$ ion, as the chemical form of $^{14}C$, was evaluated as 11mg-C/g-resin. Adsorption affinity of anions to the resin was derived in order of $NO_3{^-}$ > $HCO_3{^-}{\geq}H_2PO_4{^-}$. Thus the competitive adsorption affinity of $NO_3{^-}$ ion in binary systems appeared far higher than that of $HCO_3{^-}$ or $H_2PO_4{^-}$, and the selective desorption of $HCO_3{^-}$ from the resin was very effective. On one hand, the affinity of $Co^{2+}$ and $Cs^+$ for the resin remained relatively higher than that of other cations in the same stripping solution. Desorption of $Cs^+$ was minimized when the summation of the metal ions in the spent resin and the other cations in solution was near saturation and the pH value was maintained above 4.5. Among the various solutions tested, from the view-point of the simple second waste process, $NH_4H_2PO_4$ solution was preferable for the stripping of $^{14}C$ from the spent resin.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.4
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• pp.459-467
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• 2021

In the majority of countries, the upper limit of buffer temperature in a repository is set to below 100℃ due to the possible illitization. This smectite-to-illite transformation is expected to be detrimental to the swelling functions of the buffer. However, if the upper limit is increased while preventing illitization, the disposal density and cost-effectiveness for the repository will dramatically increase. Thus, understanding the characteristics and creating a database related to the buffer under the elevated temperature conditions is crucial. In this study, a strategy to investigate the bentonite found in Korea under the elevated temperatures from a mineral transformation and radionuclides retardation perspective was proposed. Certain long-term hydrothermal reactions generated the bentonite samples that were utilized for the investigation of their mineral transformation and radionuclide retardation characteristics. The bentonite samples are expected to be studied using in-situ synchrotron-based X-Ray Diffraction (XRD) technique to determine the smectite-to-illite transformation. Simultaneously, the 'high-temperature and high-pressure mineral alteration measurement system' based on the Diamond Anvil Cell (DAC) will control and provide the elevated temperature and pressure conditions during the measurements. The kinetic models, including the Huang and Cuadros model, are expected to predict the time and manner in which the illitization will become detrimental to the performance and safety of the repository. The sorption reactions planned for the bentonite samples to evaluate the effects on retardation will provide the information required to expand the current knowledge of repository optimization.

• Journal of Radiation Protection and Research
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• v.48 no.2
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• pp.84-89
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• 2023

Background: Food consumption is one of the most important routes for radionuclide intake for the public; therefore, there is the need to have a comprehensive understanding of the amount of radioactivity in food products. Consumption of radionuclide-contaminated food could increase potential health risks associated with exposure to radiation such as cancers. The present study aims to determine radioactivity levels in some food products (milk, rice, sugar, and wheat flour) consumed in Mali and to evaluate the radiological effect on the public health from these radionuclides. Materials and Methods: The health impact due to ingestion of radionuclides from these foods was evaluated by the determination of activity concentration of radionuclides ²³⁸U, ²³²Th, ⁴⁰K, and ¹³⁷Cs using gamma spectrometry system with high-purity germanium detector and radiological hazards index in 16 samples collected in some markets, mall, and shops of Bamako-Mali. Results and Discussion: The average activity concentrations were 9.8±0.6 Bq/kg for ²³⁸U, 8.7±0.5 Bq/kg for ²³²Th, 162.9±7.9 Bq/kg for ⁴⁰K, and 0.0035±0.0005 Bq/kg for ¹³⁷Cs. The mean values of radiological hazard parameters such as annual committed effective dose, internal hazard index, and risk assessment from this work were within the dose criteria limits given by international organizations (International Commission on Radiological Protection and United Nations Scientific Committee on the Effects of Atomic Radiation) and national standards. Conclusion: The results show low public exposure to radioactivity and associated radiological impact on public health. Nevertheless, this study stipulates vital data for future research and regulatory authorities in Mali.