• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.421-429
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• 2018

Hydrazine based reductive dissolution applied on magnetite oxide was investigated. Dissolution of Fe(II) and Fe(III) from magnetite takes place either by protonation, surface complexation, or reduction. Solution containing hydrazine and sulfuric acid provides hydrogen to break bonds between Fe and oxygen by protonation and electrons for the reduction of insoluble Fe(III) to soluble Fe(II) in acidic solution of pH 3. In terms of dissolution rate, numerous transition metal ions were examined and Cu(II) ion was found to be the most effective to speed up the dissolution. During the cycle of Cu(I) ions to Cu(II) ions, the released electron promoted the reduction of Fe(III) and Cu(II) ions returned to Cu(I) ion due to the oxidation of hydrazine. In the experimental results, the addition of a very low amount of cupric ion (about 0.5 mM) to the solution increased the dissolution rate about 40% on average and up to 70% for certain specific conditions. It is confirmed that even though the coordination structure of copper ions with hydrazine is not clear, the $Cu(II)/H^+/N_2H_4$ system is acceptable regarding the dissolution performance as a decontamination reagent.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.397-410
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• 2018

When considering the long-term safety assessment of spent-nuclear fuel management, americium is one of the most radio-toxic actinides. Although spectroscopic methods are widely used for the study of actinide chemistry, application of those methods to americium chemistry has been limited. Herein, we purified $^{241}Am$ to obtain a highly pure stock solution required for spectroscopic studies. Quantitative and qualitative analyses of purified $^{241}Am$ were carried out using liquid scintillation counting, and gamma and alpha radiation spectrometry. Highly sensitive absorption spectrometry coupled with a liquid waveguide capillary cell and time-resolved laser fluorescence spectroscopy were employed for the study of Am(III) hydrolysis and oxalate (Ox) complexation. $Am^{3+}$ ions under acidic conditions exhibit maximum absorbance at 503 nm, with a molar absorption coefficient of $424{\pm}8cm^{-1}{\cdot}M^{-1}$. $Am(OH)_3(s)$ colloidal particles formed under near neutral pH conditions were identified by monitoring the absorbance at around 506-507 nm. The formation of ${Am(Ox)_3}^{3-}$ was detected by red-shifts of the absorption and luminescence spectra of 4 and 5 nm, respectively. In addition, considerable enhancements of the luminescence intensities were observed. The luminescence lifetime of ${Am(Ox)_3}^{3-}$ increased from 23 to 56 ns, which indicates that approximately six water molecules are replaced by carboxylate ligands in the inner-sphere of the Am(III). These results suggest that ${Am(Ox)_3}^{3-}$ is formed through the bidentate coordination of the oxalate ligands.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.167-181
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• 2019

Currently, the development of evaluation technology for vibration and shock load characteristics and spent nuclear fuel structural integrity under normal conditions of transport is being conducted in the Republic of Korea. This is the first such research conducted in the Republic of Korea and, thus, previous international studies need to be investigated and will be referred to in the ongoing project. Before 2000, several studies related to measurement of vibration and shock loads on spent nuclear fuel were conducted in the US. US national research institutes conducted uniaxial fuel assembly shaker tests, concrete block tests, and multi-axis fuel assembly tests between 2009 and 2016. In 2017, multi-modal transportation tests including road, sea, and rail transport were also performed by research institutes from the US, Spain and the Republic of Korea. Therefore, test preparation procedures, acceleration and strain measurement results, and finite-element and multi-body dynamics analysis were investigated. Based on the measured strain data, the preliminary conclusion was obtained that the measured strain was too small to cause damage to spent nuclear fuel rods. However, this conclusion is a preliminary conclusion that only reviews part of the results; a detailed review is being conducted in the US. The investigation of international studies on spent nuclear fuel structural integrity tests considering vibration and shock loads under normal conditions of transport in the US will be useful data for the project being conducted in the Republic of Korea.

• Journal of the Korea institute for structural maintenance and inspection
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• v.23 no.2
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• pp.107-115
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• 2019

The many countries are facing the shortage of natural resources, and the supply of aggregates are being exhausted. To consider this situation a variety of studies were performed for the development of alternative resources. In particular, high density filler material was used for shielding radioactive waste, large amount of natural aggregates are required in order to produce filler material. Also, in order to improve the shielding performance of filler material, it is required to increase the density of the filler material. Therefore, in this study was carried out to provide basic data for expanding the feasibility of high density industrial waste resource as aggregate in heavyweight concrete. From the test results, OPC case, concrete strength decreased by using heavyweight waste glass as fine aggregate, however, it is improved by using mineral admixture as binder. Therefore, when the heavyweight waste glass and steel slag are applied to heavyweight concrete, it is desirable to use mineral admixture, especially to use BFS than FA. Meanwhile, when the steel slag was replaced as coarse aggregate of heavyweight concrete, elasticity of modulus and radiation shielding performance can be improved owing to high density of steel slag.

• The Korean Journal of Nuclear Medicine Technology
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• v.23 no.1
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• pp.50-53
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• 2019

Purpose Yttrium-90 (Y-90) is high-energy beta emitters ($E{\beta}$, max = 2.28 MeV) with the mean penetration depth of 2.5 mm in tissue. Radioactive microspheres containing Y-90 is widely used for the transarterial radioembolization of hepatocellular carcinoma. However, bremsstrahlung radiation from Y-90 can cause the external radiation exposure to medical staff who handle the Y-90 microspheres. In this study, shielding device for Y-90 microspheres was developed to minimize the external radiation exposure. Materials and Methods Y-90 microsphere shielding device was made from 6 mm thicknesses of tungsten including the lead glass window. Radiation shielding ability of Y-90 microsphere shielding device was evaluated using 4 GBq of $SIR-Spheres^{(R)}$ Y-90 microspheres. The bremsstrahlung radiation was measured using radiation survey meter. Results The mean radiation dose of Y-90 microspheres in acrylic shield was $261.7{\pm}2.3{\mu}Sv/h$ (n=5) at 10 cm away from the shield. With the additional tungsten shielding device, it was $23.7{\pm}1.3{\mu}Sv/h$ (n=5). Thus, the bremsstrahlung radiation dose was decreased by 90.9%. At 50 cm away from the shield, bremsstrahlung radiation was reduced by 89.2% after using tungsten shielding device. Conclusion During the preparation and radioembolization of Y-90 microsphere, medical staff are exposed to external radiation. In this study, we demonstrated that the use of tungsten shielding device devices significantly reduced the amount of bremsstrahlung radiation. Y-90 microsphere tungsten shielding device can be highly effective in reducing the bremsstrahlung radiation.

• Journal of the Korean Society of Radiology
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• v.13 no.3
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• pp.433-438
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• 2019

The medical institutions use radiation generating devices and radioactive isotopes to diagnose and treat patients. The patient transporter performs work in an environment that is more likely to be exposed to radiation when compared with the general public, such as inevitably entering the radiation management area for patient transfer, or transferring the isotope-administered patient at a short distance. For this reason, we conducted a study to determine the degree of exposure of the patient transporter. The 12 patient transporters working at Incheon A General Hospital are eligible. From April 1, 2019 to April 30, 2019, the dosimeter was used in the chest for one month and the accumulated dose was measured. The dosimeter used was a Optically Stimulated Luminescence Dosimetry (OSLD) and the dose reading was OSLD Microstar Reading System. As a result of cumulative dose measurement for one month, the average of the deep dose was 0.13 mSv and the surface dose was 0.13 mSv, and the cumulative dose for one month was multiplied by 12 to estimate the cumulative dose expectation As a result, the average of the deep dose and the surface dose were 1.52 mSv and 1.51 mSv, respectively. It is necessary to classify the patient transporter as a frequent visitor in order to measure and manage the exposure dose, increase the knowledge of protection against radiation through education and training, and prevent radiation trouble through medical examination.

• Resources Recycling
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• v.28 no.4
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• pp.15-22
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• 2019

Coal combustion products are generated during coal combustion and can be grouped into fly ash and bottom ash depending on collection methods. Fly ash and bottom ash can be recycled for various purposes based on their characteristics. Australia is the fourth largest coal production country in the world and reuses coal ash as cement, concrete, mine filler, and agricultural soil amendment. When fly ash is used as a supplement for cement and concrete, strength of the cement and the durability of the concrete can be improved. Use of coal combustion product for mine backfill stabilizes underground mine voids and stores a large amount of coal ash in the voids. Because of alkalinity of coal combustion products, it can neutralize acid mine drainage when used for mine backfill. In addition, it can be used as an agricultural soil amendment to improve acidity and physical properties of the soil and to supply plant nutrients. Recycling of fly ash in Australia will be further expanded because of its low trace element contents that can be toxic to plants and low radioactive element contents existing within soil background concentrations. The characteristics of coal combustion products are related to the characteristics of the coal used for combustion, and since Korea imports coal from Australia, Korean coal combustion products also can be recycled for various purposes.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.347-353
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• 2019

The composition of the reduced fuel produced in the electrolytic reduction process of pyroprocessing affects the concentration change of $UCl_3$, an important operating variable of the electrorefining process. In this study, we examined the concentration change of $UCl_3$ in the electrorefiner according to the content of TRU and RE elements in the reduced fuel and the concentration of $Li_2O$ introduced in the electrorefiner accompanied with the reduced fuel. Considering only the TRU and RE elements, the concentration of $UCl_3$ decreased with increasing the number of electrorefining operation batch. In order to operate one campaign (20 batches) of electrorefining process, it was found that additional injection of $UCl_3$ should be conducted more than 3 times. On the other hand, the concentration of $UCl_3$ in the electrorefiner changed significantly depending on the concentration of $Li_2O$ and, accordingly the number of operable electrorefining batches decreased rapidly, showing that the concentration of $Li_2O$ is an important operating variable in electrorefining. Therefore, the results of this study show that to maintain the concentration of $UCl_3$ in the electrorefiner, the operation mode should be set by taking into account the effect of $Li_2O$ as well as the TRU and RE elements contained in the reduced fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.279-298
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• 2019

This study comprehensively addresses recent progress at KAERI in waste treatment technology to cope with waste produced by pyroprocessing, which is used to effectively manage spent fuel. The goal of pyroprocessing waste treatment is to reduce final waste volume, fabricate durable waste forms suitable for disposal, and ensure safe packaging and storage. KAERI employs grouping of fission products recovered from process streams and immobilizes them in separate waste forms, resulting in product recycling and waste volume minimization. Novel aspects of KAERI approach include high temperature treatment of spent oxide fuel for the fabrication of feed materials for the oxide reduction process, and fission product concentration or separation from LiCl or LiCl-KCl salt streams for salt recycling and higher fission-product loading in the final waste form. Based on laboratory-scale tests, an engineering-scale process test is in progress to obtain information on the performance of scale-up processes at KAERI.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.321-328
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• 2019

The detection performances of the NaI(Tl), $LaBr_3$(Ce) and $CeBr_3$ scintillation detectors, which can be used to rapidly evaluate the major artificial radionuclides deposited on the soil surface in a nuclear accident or radiological emergency, were compared. Detection performance was assessed by calculating the minimum detectable activity (MDA). The detection efficiency of each detector for artificial radionuclides was semi-empirically determined using mathematical modelling and point-like sources having certified radioactivity. The background gamma-ray energy spectrum for MDA evaluation was obtained from relatively wide and flat grassland, and the MDA values of each detector for the major artificial radionuclides that could be released in nuclear accidents were calculated. As a result, the relative MDA values of each detector regarding surface deposition distribution at normal environmental radiation level were evaluated as high in the order of the NaI(Tl), $LaBr_3$(Ce), and $CeBr_3$ detectors. These results were compared based on each detector's intrinsic and measurement environment background, detection efficiency, and energy resolution for the gamma-ray energy region of the radionuclide of interest.