• Journal of Radiation Industry
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• v.17 no.4
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• pp.457-469
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• 2023

The clearance level by nuclide is announced by the Nuclear Safety and Security Commission. However, the clearance level of uranium existing in nature has not been announced, and research is needed. Therefore, the purpose of this study was to evaluate the clearance level of uranium nuclides appropriate to domestic conditions preliminary. For this purpose, this study selected major processes for recycling metal wastes and analyzed the exposure scenarios and major input factors by investigating the characteristics of each process. Then, the radiation dose to the general public and workers was evaluated according to the selected scenarios. Finally, the results of the radiation dose per unit radioactivity for each scenario were analyzed to derive the clearance level of uranium in metal waste. The results of the radiation dose assessment for both the general public and workers per unit radioactivity of uranium isotopes were shown to meet the allowable dose (individual dose of 10 µSv y^-1 and collective dose of 1 Man-Sv y^-1) regulated by the Nuclear Safety and Security Commission. The most conservative scenarios for volumetric and surface contamination were evaluated for the handling of the slag generated after the melting of the metal waste and the direct reuse of the contaminated metal waste into the building without further disposal. For each of these scenarios, the radioactivity concentration by uranium isotope was calculated, and the clearance level of uranium in metal waste was calculated through the radioactivity ratio by enrichment. The results of this study can be used as a basic data for defining the clearance level of uranium-contaminated radioactive waste.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2317-2323
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• 2024

Medical linear accelerators with an energy of 8 MV or higher are radiated owing to photonuclear reactions and neutron capture reactions. It is necessary to quantitatively evaluate the concentration of radioactive isotopes when replacing or disposing them. HPGe detectors are commonly used to identify isotopes and measure radioactivity. However, because the detection efficiency is generally calibrated using a standard material with a density of 1.0 g/cm³, a self-absorption effect occurs if the density of the measured material is high. In this study, self-absorption correction factors were calculated for tungsten, lead, copper, and SUS-303, which are the main materials of medical linear accelerator head parts, for each gamma-ray energy using MCNP 6.2 code. The self-absorption effect was more pronounced as the energy of the emitted gamma rays decreased and the density of the measured materials increased. These correction factors were applied to the radioactivity measurements of the in-built and portable HPGe detectors. Furthermore, compared to the surface dose rate measured by the survey meter, the accuracy of the measurements of radioactivity improved by an average of 124.31 and 100.53 % for inbuilt and portable HPGe detectors, respectively. The results showed a good agreement, with an average difference of 3.70 and 5.24 %.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2444-2451
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• 2024

This work demonstrates a new sol-gel approach for synthesizing PbO₂-doped zirconia using zircon mineral precursors. The streamlined methodology enables straightforward fabrication of the doped zirconia composites. Comprehensive materials characterization was performed using XRD, SEM, and TEM techniques to analyze the crystal structure, microstructure, and morphology. Quantitative analysis of the XRD data provided insights into the nanoscale crystallite sizes achieved, along with their relationship to lattice imperfections. Furthermore, the gamma-ray shielding capacity for the PbO₂-doped zirconia samples was estimated by the Monte Carlo simulation, which proves an increase in the gamma ray shielding properties by raising the Pb concentration. The linear attenuation coefficient increased between 0.467 and 0.499 cm^-1 (at 0.662 MeV) by increasing the Pb content between 11 and 21 wt%. By increasing the Pb content to 21 wt%, the synthesized composites' lead equivalent thickness reaches 2.49 cm. The radiation shielding properties for the synthesized composites revealed a remarkable performance against low and intermediate γ-ray photons, with radiation shielding capacity of 37.3 % and 21.4 % at 0.662 MeV and 2.506 MeV, respectively. As a result, the developed composites can be employed as an alternative shielding material in hospitals and radioactive zones.

• Korean Journal of Mineralogy and Petrology
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• v.35 no.2
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• pp.101-109
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• 2022

The mobility and transport of radioactive cesium are crucial factors to consider for the safety assessment of high-level radioactive waste disposal sites in granite. The retardation of radionuclides in the fractured crystalline rock is mainly controlled by the hydrochemical condition of groundwater and surface reactions with minerals present in the fractures. This paper reports the experimental results of cesium sorption to the Wonju Granite, a typical Mesozoic granite in Korea, performed in an anaerobic chamber that mimics the anoxic environment of a deep disposal site. We measured the rates and amounts of cesium (¹³³Cs) removed by crushed granite samples in different electrolyte (NaCl, KCl, and CaCl₂) solutions and a synthetic groundwater solution, with variations in the initial cesium concentration (10^-5, 5×10^-6, 10^-6, 5×10^-7 M). The cesium sorption kinetic and isotherm data were successfully simulated by the pseudo-second-order kinetic model (r²= 0.99) and the Freundlich isotherm model (r²= 0.99), respectively. The sorption distribution coefficient of granite increased almost linearly with increasing biotite content in granite samples, indicating that biotite is an effective cesium scavenger. The cesium removal was minimal in KCl solution compared to that in NaCl or CaCl₂ solution, regardless of the ionic strength and initial cesium concentration that we examined, showing that K⁺ is the most competitive ion against cesium in sorption to granite. Because it is the main source mineral of K⁺ in fracture fluids, biotite may also hinder the sorption of cesium, which warrants further research.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.85-90
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• 2007

Fenton's Reagent is applied to directly dissolve the cation-exchange resin, IRN-77. The characteristics of the experimental procedure is to dry the resin first and $FeSO_4$ solution is completely absorbed into the resin, and then $H_2O_2$ is introduced later for an effective reaction between the reagents within the resin. An a characteristic of the dissolution, the lag time is needed for about 1 hour until the main reaction is occurred, which was more affected with the less concentration of $FeSO_4$ and the less initial dose of $H_2O_2$. The dose of $H_2O_2$ was equally divided into the early stage and the later stage after the initial reaction to provide an effective and safe reaction condition. The optimum conditions is appeared that the concentration of $FeSO_4$ is 0.9M and the dose of 15% $H_2O_2$ solution is 6-7 volume for the dissolution of unit weight of IRN-77. The effect of the heating on the lag time was checked and the time could be reduced within 5 minutes at $50^{\circ}C$, which is a relatively low temperature. The large amount of the resin, 5g and 10g, was also completely decomposed by increasing the dose of $H_2O_2$ to 9-10 volume ratio.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.221-227
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• 2007

Fenton's reagent is applied to directly decompose the ion-exchange resins, IRN-78 and the mixed resin with IRN-77. The newly applied procedures is to dry the resin first and the catalyst solution is completely absorbed into the resin, then a limited dose of $H_2O_2$ is introduced for an effective reaction between the reagents within the resin. As a characteristic on the decomposition of IRN-78, the resin mixture should be heated to $40^{\circ}C$ to induce the initial reaction and lag time is also needed for about 20 minutes until the main reaction occurs. The effectiveness of the decomposition is investigated using $CuSO_4,\;Cu(NO_3)_2\;and\;FeSO_4$ as a catalyst and the decomposition rate is compared depending on the concentration of each catalyst and the amount of $H_2O_2$. The most effective catalyst was found to be $FeSO_4$ for IRN-78 alone and the mixed resin with IRN-77, and $FeSO_4$ showed a special effect that the reaction was initiated without heating and a lag time. Furthermore, the optimum concentration of the catalyst for each resin and the mixed one is suggested in the view point of the amount of $H_2O_2$ needed and the stability of the decomposition reaction.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.63-68
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• 2013

To see if slaked lime and organic fertilizer applications to soil are useful as countermeasures for reducing $^{99}Tc$ concentrations in rice seeds after $^{99}Tc$ contamination of paddy fields, pot experiments were performed for two different paddy soils in a greenhouse. The upper soils for a depth of about 20 cm were treated with the agricultural materials and $^{99}Tc$ 15 d before transplanting. The effects were compared using the transfer factor (TF) defined as the ratio of the plant concentration to the soil concentration. In the case of control plants, TF values for brown rice in the two soils were $4.1{\times}10^{-4}$ and $4.3{\times}10^{-4}$. Of various types of the application, only the application of slaked lime at a lower dose (about 0.6 kg $m^{-2}$), which led to a 60% reduction in the TF value for one soil, seemed to be worth using as a countermeasure. Little effect of the same application was found in the other soil so it is important to determine the effect averaged for a number of soils. Organic fertilizer applications at both of two different doses increased the TF value. It is considered necessary to perform experiments for slake lime applications at doses lower than the above.

• Applied Biological Chemistry
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• v.45 no.1
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• pp.30-36
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• 2002

A bioconcentration experiment was performed for killifish using nonradioactive and radioactive butachlor. At 0.036 ppm concentration, the highest bioconcentration ratio $(C_f/C_w)$ and BCF at steady state recorded as 296 and 87 respectively. And at 0.0036 ppm concentration, the highest $C_f/C_w$ ratio was 169 and the BCF was 51 at steady state. Considering the experimental variation of the BCF's, the BCF of butachlor was tentatively determined to be $69{\pm}28$. And the $^{14}C-butachlor$ and its metabolites depurated about 50% within 12 hours and 90% within 30 hours after depuration experiment started. And in vivo metabolites, designated as M-I, M-II, and M-III, were found in killifish and the excretes as butachlor was metabolised.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.3
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• pp.163-170
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• 2008

The reductive stripping of Np using a n-butyraldehyde (NBA) from loaded organic solution containing Np, which was oxidative-extracted in a system of a 30 % TBP/NDD-2M $HNO_3$ and O/A=2 containing 0.005 M $K_2Cr_2O_7$ as an oxidant of Np, was studied. The stripping yields of Np was increased with an increasing the NBA concentration, with a decreasing the nitric acid concentration of stripping solution and with a decreasing the reaction temperature. The apparent reductive stripping rate equation was shown by the following equation : $-d[Np]_{Org.}/dt$ = 1,524 exp(-2,906/T) $[NBA]^{0.91}\;[H^+]^{-0.92}[Np]_{Org.}$. At 1.04 M NBA and 2 M $NHO_3$, the stripping yield of Np and U was 70.1 %, and 7.1 %, respectively, and the separation factor of U over Np ($=D_U/D_{Mp}$) was about 30.4. Therefore, it was found that U and Np co-extracted in a system of TBP-$HNO_3$ could be effectively mutual-separated by the NBA.

• Journal of Korean Society of Environmental Engineers
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• v.22 no.10
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• pp.1765-1775
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• 2000

The removal of radioactive organic iodide generated from high temperature process in nuclear facility was generally performed by silver ion-exchanged synthetic zeolite (AgX). The purpose of this study is to obtain fundamental data for the substitution of natural zeolite(NZ) in stead of synthetic zeolite as supporter for the removal of methyl iodide in high temperature conditions. Therefore, NZ was modified with NaCl, $NaNO_3$ solution, and the analysis of the physical or surface characteristics through XRD, SEM-EDAX, and BET analysis was performed. In order to obtain the optimal surface-modification condition of NZ, adsorption capacities at $150^{\circ}C$ on surface-modified silver ion-exchanged NZ prepared with the variation of solution concentration were evaluated. The optimal condition of surface modification is that concentration of $NaNO_3$ and $AgNO_3$ are 1N and 1.2N, respectively(namely Ag-SMNZ). The adsorption isotherm of methyl iodide on Ag-SMNZ in a range of $100^{\circ}C$ to $300^{\circ}C$ was obtained, which is similar to that of 13X, and the maximum adsorption amount of Ag-SMNZ reached approximately 50% that of AgX. It would be evaluated that the adsorption capacity at $150{\sim}200^{\circ}C$ is relatively higher than other temperature, and the chemisorption between silver and iodide is attributed to a strong binding even after desorption test.