• Analytical Science and Technology
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• v.18 no.1
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• pp.5-12
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• 2005

It has been established that soft error of high precision electronic circuits can be induced by alpha particles emitted from the naturally occurring radioactive impurities such as U, and Th. As the electronic circuits have recently become lower dimension and higher density, these alpha-particle emitting radioactive impurities have to be strictly controlled. The aim of this study is to develop of NAA (Neutron Activation Analysis) and gamma-spectrometry to improve the analytical sensitivity and precision of U and Th. A new NAA method has been established using the HTS (Hydrulic transfer system) irradiation facility which has been used to produce radioisotopes for industries and medicines instead of the PTS (pneumatic transfer system) irradiation facility which has been used in general NAA. When the ultratrace impurities have to be analyzed by NAA, background gamma-ray spectra induced from $^{222}Rn$ and its progenies in air is serious problem. This unstable background has been eliminated or stabilized by the use of a nitrogen purging system. Ultra trace amounts of U (0.1 ng/g) and Th (0.01 ng/g) in high purity silica used for EMC could be analyzed by the use of HTS-NAA and low background gamma-spectrometry.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1756-1763
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• 2020

The paper studies spark plasma sintering (SPS) of industrially used UO₂-based fuel containing integral fuel burnable absorber (IFBA) of neutrons Gd₂O₃. Densification dynamics of pristine UO₂ powder and the one added with 2 and 8 wt% of Gd₂O₃ under ultrasonication in liquid has been studied under SPS conditions at 1050, 1250, and 1450 ℃. Effect of sintering temperature on phase composition as well as on O/U stoichiometry has been investigated for UO₂ SPS ceramics. Sintering of uranium dioxide added with Gd₂O₃ yields solid solution (U,Gd)O₂, which is isostructural to UO₂. SEM with EDX and metallography were implemented to analyze the microstructure of the obtained UO₂ ceramics and composite UO₂-Gd₂O₃ one, particularly, open porosity, defects, and Gd₂O₃ distribution were studied. Microhardness, compressive strength and density were shown to reduce after addition of Gd₂O₃. Obtained results prove the hypothesis on formation of stable pores in the system of UO₂-Gd₂O₃ due to Kirkendall effect that reduces sintering efficiency. The paper expands fundamental knowledge on pros and cons of fuel fabrication with IFBA using SPS technology.

• Journal of Radiation Protection and Research
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• v.4 no.1
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• pp.14-20
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• 1979

For accurate and easily shielding irregular shaped organ, its minimized penumbra region and a low melting point alloy 'Lead Y' and synchronizing instrument have been developed. The 'Lead Y' is the quaternary eutectic alloy and it is composed of Lead 30.0% Tin 11.5% Bismuth 48 5% Cadmium 10.0% The density of its at $22^{\circ}C$ is $9.8g/cm^3$ and the melting temperature has $40^{\circ}C\;to\;68^{\circ}C$. The thickness of 'Lead Y' for perfect shielding of Co-60 gamma ray and LINAC 10MeV x-ray is 6cm and 7cm respectively. The 'Lead Y' shielding block is casted directly on the styrofoam from which is cut with hot wire of synchronizer device. The special features and advantages of the Lead Y shielding block could be summarized as follows; 1. The shielding block for radiotherapy is rapidly processed only with boiling water and styrofoam. 2. It is not injure one's health and not danger of a fire, because of not generating of any metals vapor and evil smelling. 3. It is very effective to minimize secondary penumbra for the protection of healthy tissue from unnecessary ionizing radiation regardless of the magnification source to skin distance. 4. The HVL of the Lead Y is 1.2cm for Co-60 gamma ray and it's shielding effect is almost same as the pure lead block. 5. The hardness of Lead Y is 1.5 times higher than lead block. 6. It's reavailability is higher than lead block and then one block of Lead Y is reavailable about 30 to 40 times. 7. It is usefull for shielding of x-ray, gamma ray, beta-ray, electron and neutron radiation. 8. The materials for Lead Y are easy to acquire with reasonable price and tractable.

• Journal of the Korean Chemical Society
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• v.27 no.4
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• pp.279-286
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• 1983

An apparatus was made for the electrodeposition of alpha emitting actinide nuclides, $^{207}Bi$ and $^{210}Po$. The electrodeposition was made on a polished stainless steel plate cathode. The anode was made of platinum wire and to stir the solution. With the ammonium chloride as electrolyte initial pH = 4, chloride concentration = 0.6M and solution volume = 15ml, a current of 1.5 ampere(current density = 0.59A/$cm^2$) was flowed for 100 minutes for the quantitative recovery of electrodeposition and on average recovery of 98.3% was obtained within ${\pm}$0.7% uncertainty. Alpha spectrometry of the electrodeposited sample showed alpha peaks from $^{210}Po, ^{234}U$ and $^{239}Pu$ having energy resolution (FWHM) of 18.3, 21.8 and 36.0 keV respectively. The electrodeposition and alpha spectrometry for a natural uranium sample of domestic origin gave $^{238}U : ^{234}U = 1 : 6.1{\times}10^{-5}$ and for a neutron-irradiated uranium sample did $^{238}U : ^{239}Pu : ^{241}Am = 100 : 0.0263 : 5.20{times}10^{-5}$. The result of $^{238}U$ determination in the irradiated sample by electrodeposition-alpha spectrometry was in accord within ${\pm}1.6%$ of relative error with the results of solid fluorimetry and mass spectrometry. For $^{239}Pu$ the result of electrodeposition-alpha spectrometry was in accord within ${\pm}$4.0% of relative error with the results of anion exchange separation and the thenoyltrifluoroacetone(TTA) extraction both followed by alpha spectrometries.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.243-260
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• 2018

A literature review on the effects of high temperature and radiation on radiation shielding concrete in Spent Fuel Dry Storage is presented in this study with a focus on concrete degradation. The general threshold is $95^{\circ}C$ for preventing long-term degradation from high temperature, and it is suggested that the temperature gradient should be less than $60^{\circ}C$ to avoid crack generation in concrete structures. The amount of damage depends on the characteristics of the concrete mixture, and increases with the temperature and exposure time. The tensile strength of concrete is more susceptible than the compressive strength to degradation due to high temperature. Nuclear heating from radiation can be neglected under an incident energy flux density of $10^{10}MeV{\cdot}cm^{-2}{\cdot}s^{-1}$. Neutron radiation of >$10^{19}n{\cdot}cm^{-2}$ or an integrated dose of gamma radiation exceeding $10^{10}$ rads can cause a reduction in the compressive and tensile strengths and the elastic moduli. When concrete is highly irradiated, changes in the mechanical properties are primarily caused by variation in water content resulting from high temperature, volume expansion, and crack generation. It is necessary to fully utilize previous research for effective technology development and licensing of a Korean dry storage system. This study can serve as important baseline data for developing domestic technology with regard to concrete casks of an SF (Spent Fuel) dry storage system.