• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.51-59
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• 2005

The long-lived fission product $^{99}Tc$ is present in large quantities in nuclear wastes and its chemical behavior in aqueous solution is of considerable interest. Under oxidizing conditions technetium exists as the anionic species $TcO_4^-$ whereas under the reducing conditions it is generally predicted that technetium will be present as $TcO_2{\cdot}nH_2O$. Technetium oxide was prepared by reduction of a technetate solution with $Sn^{2+}$. The concentration of total technetium and Tc(IV) species in the solutions were periodically determined by separating the oxidized and reduced technetium species using a solvent extraction procedure and counting the beta activity of the $^{99}Tc$ with a liquid scintillation counter. The experimental results show that the rate of oxidation of Tc(IV) in simulated groundwater and redistilled water is about $(1.49{\~}1.86){\times}10^{-9} mol/(L{\cdot}d$) under aerobic conditions, but Tc(IV) in simulated groundwater and redistilled water is not oxidized under anaerobic conditions. Under aerobic or anaerobic conditions the solubility of Tc(IV) oxide in simulated groundwater and redistilled water is equal on the whole.

• Nuclear Engineering and Technology
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• v.45 no.1
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• pp.61-66
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• 2013

Among the radioactive wastes generated from the nuclear power plant, a radioactive nuclide such as $^{129}I$ is classified as a difficult-to-measure (DTM) nuclide, owing to its low specific activity. Therefore, the establishment of an analytical procedure, including a chemical separation for $^{129}I$ as a representative DTM, becomes essential. In this report, the adsorption and recovery rate were measured by adding $^{125}I$ as a radio-isotopic tracer ($t_{1/2}$ = 60.14 d) to the simulation sample, in order to measure the activity concentration of $^{129}I$ in a pressurized-water reactor primary coolant. The optimum condition for the maximum recovery yield of iodine on the anion exchange resins (AG1 x2, 50-100 mesh, $Cl^-$ form) was found to be at pH 7. In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of $^{129}I$ was examined, as was the effect of $^3H$ on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous $^3H$ presence was found with activity concentrations of $^3H$ lower than 50 Bq/mL, and with a boron concentration of less than 2,000 ${\mu}g/mL$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.3
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• pp.323-330
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• 2021

Silver tellurite glasses with melting temperature of approximately 700℃ were developed to immobilize ¹²⁹I wastes. Long-term dissolution tests in 0.1 M acetic acid and disposability assessment were conducted to evaluate sustainability of the glasses. Leaching rate of Te, Bi and I from the glasses decreased for up to 16 d, then remained stable afterwards. On the contrary, tens to tens of thousands of times more of Ag was leached in comparison to the other elements; additionally, Ag leached continuously for all 128 d of the test owing to the exchange of Ag⁺ and H⁺ ions between the glasses and solution. The I leached much lower than those of other elements even though it leached ~10 times more in 0.1 M acetic acid than in deionized water. Some TeO₄ units in the glass network were transformed to TeO₃ by ion exchange and hydrolysis. These silver tellurite glasses met all waste acceptance criteria for disposal in Korea.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.90-95
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• 2003

Chemical management of annulus gas system is carried out for the purpose of ensuring the safety and reliability of the system via securing the integrity of the system, detecting the D$_2$O in-leakage of coolant and/or moderator, and reducing the radiation dose. Since the quality of CO_2$ gas, which is used as a filling gas for annulus gas system at CANDU plants, has a propound effect on the integrity of the system material and the radiation dose, CO_2$ gas of high quality is needed. If the quality of CO_2$ gas does not meet the specification, it may give rise to undesirable effect not only on the annulus gas system, but also on the environment due to the production of radioactive nuclei. Therefore, it is very important to check the impurities of CO_2$ gas. Based on this background, the inventories of C-14 and Ar-41 in CO_2$ gas that is supplied as annulus gas were estimated using the data on concentrations of the impurities of $CO_2$ such as C, N_2$ and Ar. The results of this study is expect to give useful information on optimization of CO_2$ impurities maintenance and management of gaseous radioactive wastes produced at CANDU plants.

• Nuclear Engineering and Technology
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• v.9 no.2
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• pp.75-81
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• 1977

This experiment has been carried out to determine the pH dependent cation exchange capacity concerning the sorption phenomenon of long-lived radionuclides contained in low-level liquid radioactive waste on various clay minerals. The pH dependent cation exchange capacity determined by Sawhney's method are used to the analysis of sorption phenomenon. About 70 percent of the total cation exchange capacity is contributed by the pH dependent CEC due to the negative charge originated naturally in clays in case of clinoptilolite, vermiculite and sodalite. It is sugested in this test that the high neutral salt CEC, that is, highly charged clays would show good fixation yield. The removal of radionuclides at the pH range more than pH 9 is considered the hydroxide precipitation of metal ion rather than the cation exchange. The Na-clay prepared by the method of successive isomorphic substitution with electrolyte showed a considerable improvement in removal efficiency for the decontamination.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.33-40
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• 2006

The characteristics of the aluminum waste melting and the distribution of the radioactive nuclides have been investigated for the estimation on the volume reduction and the decontamination of the aluminum wastes from the decommissioning of the TRIGA MARK it and III research reactors at the Korea Atomic Energy Research Institute(KAERI). The aluminum wastes were melted with the use of the fluxes such as flux $A:NaCl-KCl-Na_3AlF_6$, flux B:NaCl-NaF-KF, flux $C:CaF_2$, and flux $D:LiF-KCl-BaCl_2$ in the DC graphite arc furnace. For the assessment of the distribution of the radioactive nuclides during the melting of the aluminum, the aluminum materials were contaminated by the surrogate nuclides such as cobalt(Co), cesium(Cs) and strontium(Sr). The fluidity of aluminum melt was increased with the addition of the fluxes, which has slight difference according to the type of fluxes. The formation of the slag during the aluminum melting added the flux type C and D was larger than that with the flux A and B. The rate of the slag formation linearly increased with increasing the flux concentration. The results of the XRD analysis showed that the surrogate nuclide was transferred to the slag, which can be easily separated from the melt and then they combined with aluminum oxide to form a more stable compound. The distribution ratio of cobalt in ingot to that in slag was more than 40% at all types of fluxes. Since vapor pressures of cesium and strontium were higher than those that of the host metals at the melting temperature, their removal efficiency from the ingot phase to the slag and the dust phase was by up to 98%.

• 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.4 no.2
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• pp.133-138
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• 2006

Chemical wastes containing small amounts of uranium can not be disposed of them after treatment as an industrial waste, because the uranium concentration in the final dry cake exceeds the exemption level. Especially for the removal of uranium in this study, the method for immobilizing Diphosil powder within alginate beads is adopted to make a bead form from a powdered resin. Sodium alginate bead itself showed a capability to uptake uranium to above 60%, but the value was decreased to below 30% after equilibrium. The adsorption rate of uranium increased with the increasing content of Diphosil in the sodium alginate bead. Diphosil resin itself showed very fast uptake of uranium from early stages, and then the rates were leveled off. Diphosil bead showed an improved capability to uptake uranium considering the pure Diphosil content in the composite bead, and provide a considerable potential for further applications of a continuous process by using Diphosil as a bead form.

• Applied Chemistry for Engineering
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• v.17 no.6
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• pp.604-608
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• 2006

Polymer composites for measuring the radioactive contamination are prepared by coating ZnS(Ag) powders as a scintillator on polysulfone base layer. The composites consist of the active layer for a scintillation reaction with radioactive wastes and the transparent support layer for transmittance of light photons emitted by scintillation in the active layer. The binding of the active layer, including ZnS(Ag), on the support layer is proceeded via coating with polysulfone as a binder, without any extra adhesive. The coating was obtained by either casting via a Doctor Blade as applicator or screen printing. The prepared composites feature a monolithic structure, resulting in the complete adhesion between two layers. The composite prepared by the casting technique using an applicator holds a good detection efficiency in measuring the alpha radionuclide, but its structure becomes fragile because of warping in morphology. On the contrary, the composite prepared by the screen printing shows a good detection capacity as well as a good stability in a mechanical shape.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.4
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• pp.267-272
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• 2007

Various kinds of RI wastes are discharged from licensed organizations of radioisotopes les such as hospitals and clinic organizations, educational organizations, research institutions, and public organizations. Radioiodines such as $^{125}I\;and\;^{131}I$ are radioisotopes mainly used in nuclear medicine and industry. A method for the determination of radioiodines in RI wastes has been applied to measure low level activity using acid decomposition method and HPGe gamma ray spectrometer. Prior to analysis of real samples, $^{131}I$ reference solution and 10 g of yellow tissue paper was added to flask in mantle and was heated in 100 mL of 0.4 N $K_2Cr_2O_7$ and 100 mL of 9 M $H_2SO_4$, and then distilled after adding 10 mL of 30% $H_2PO_3$ and 1 mL of 30% $H_2O_2$. The condensed iodine by circulator was extracted into $CCl_4$, then back-extracted into the aqueous phase with 10 mL of 5% $K_2SO_2$ solution. Finally, $^{131}I$ was measured at 364.48 keV using HPGe gamma ray spectrometer after precipitation and filtration. Chemical yield of three steps such as acid decomposition process, chemical separation process, and precipitation and filtration process was more han 94% respectively, MDA(Minimum Detectable Activity) of $^{131}I$ at this analytical condition was 0.6 Bq/g.