• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.273-278
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• 1998

The phase stability of atomized U-10wt. ％Mo powder and the thermal compatibility of dispersed fuel meats at 40$0^{\circ}C$ and 50$0^{\circ}C$ have been characterized. Atomized U-10Mo powder has a good \ulcorner-U phase stability, and excellent thermal compatibility with aluminum matrix in a dispersion fuel. It is thought that the good phase stability is related to th large supersaturation of Mo atoms in the atomized particles. The reasons for the excellent thermal compatibility have been considered to be as follows. Before thermal decomposition of ${\gamma}$-U in particle, supersaturated Mo atoms at ${\gamma}$-U grain boundaries inhibit the diffusion of Al atoms. After thermal decomposition of ${\gamma}$-U into ${\gamma}$-U and U$_2$Mo, the intermetallic compound of U$_2$Mo seems to retard the penetration of Al atoms. The penetration mechanisms of aluminum atoms in the atomized particles are assumed be classified as (a) diffusion through the reacted layer between fuel particles and Al matrix leaving a kernel-like unreacted island and (b) diffusion along grain boundaries showing several unreacted islands and more reacted regions.

• Korean Chemical Engineering Research
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• v.57 no.3
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• pp.368-371
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• 2019

Using first principles calculations we investigated the thermomechanical stability of spent nuclear fuels (SNF), especially how mechanical properties of $UO_2$, such as, bulk, shear and Young's moduli and Poisson's ratio vary through alpha-decay of U into Th with generation of He gas. Our results indicate that substitution of U by Th through alpha decay ($U_{1-x}Th_xO_2$) does not significantly affect the stability of the grain in a fuel matrix. In addition, we studied the transport properties of He in and boundaries of the $U_{1-x}Th_xO_2$ grain. Helium preferentially resides at the grain boundaries through diffusion. Our study can contribute to substantial reduction of environmentally risk and enhancement of our sustainability by safe control of radioactive materials.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.56.2-56.2
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• 2003

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.453-460
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• 1994

The migration equation is modified for a three-member decay chain in the fracture and porous matrix Analytical solutions are obtained by utilizing Laplace transform the initial conditions of Delta function and Bateman equation. The concentrations for each nuclide of Np$^{237}$ -U$^{233}$ -Th$^{229}$ and U$^{234}$ -Th$^{230}$ -Ra$^{226}$ chains selected from the 4n＋1 and 4n＋2 chains are plotted by utilizing analytical solutions in the fracture. Retardation coefficient of the nuclides are obtained using those of the granite. The results indicate that the daughter nuclides such as U$^{233}$ , Th$^{229}$ , Th$^{230}$ and Ra$^{226}$ become important at the far field from the repository though there is very small initial inventory in the waste solid or spent fuel, for they are produced by the mother nuclides decayed in the fracture and porous matrix.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.64.2-64.2
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• 2003

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.377-383
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• 2019

Targeting at simulating the application of thorium-uranium (TU) fuel in the CANDU-6 reactor, this paper analyzes the process using the code DRAGON/DONJON where the discrete TU fuel pins are applied in the CANDU-6 reactor under the time-average equilibrium refueling. The results show that the coolant void reactivity of the assembly analyzed in this paper is lower than that of 37-element bundle cell with natural uranium and 37-element bundle cell with mixed TU fuel pins; that the max time-average channel/bundle power of the core meets the limits - less than 6700kW/860 kW; that the fuel conversion ratio is higher than that of the CANDU-6 reactor with natural uranium; and that the exit burnup increases to 13400 MWd/tU. Thus, the simulation in this paper with the fuel in the 37-element bundle cell using discrete TU fuel pins can be considered to be applied in CANDU-6 reactor with adequate modifications of the core structure and operating modes.

• Korean Chemical Engineering Research
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• v.55 no.6
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• pp.846-853
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• 2017

Cold bed and $30kW_{th}$ pilot bed tests using circulating fluidized bed (CFB) were conducted to apply oxy-fuel technology for waste sludge combustion as a carbon capture and storage technology. In cold bed test, the minimum fluidization velocity ($u_{mf}$) and superficial velocity for fast fluidization was determined as 0.120 m/s and 2.5 m/s, respectively. In the pilot test, air and oxy-fuel combustion experiments for waste sludge were conducted using CFB unit. The flue-gas temperature in 21~25% oxy-fuel combustion was higher than that of air and up to 30% oxy-fuel combustion. In addition, the concentration of carbon dioxide was more than 80% with the oxygen injection range from 21% to 25% in oxy-fuel CFB waste sludge combustion.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.551-556
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• 1998

The correlation of isotope composition of Zr with the turnup and some heavy isotopes in PWR uranium dioxide fuel has been investigated. The total and partial ($^{235}$ U) burnup were determined by $^{148Nd}$ and by U and Pu mass spectrometric method, respectively. After separating Zr from the fuel samples, its isotope composition was measured by mass spectrometry. In addition, the quantities of the U and Pu in the spent fuel were determined by isotope di lution mass spectrometric method using $^{233}$ U and $^{242}$ Pu as spikes. The content of some heavy isotopes, $^{235}$ U, $^{239}$ Pu and $^{241}$ Pu, and the Pu Contribution to total turnup were expressed by the correlation with Zr isotope ratios, $^{91}$ Zr/$^{96}$ Zr and $^{93}$ Zr/$^{96}$ Zr The correlations by isotope compositions measured were compared wi th those calculated from ORIGEN2 code.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1460-1467
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• 2023

A robust approach was conducted to determining the absolute oral bioavailable (f_ab) fractions of ²³⁸U and ²³²Th in rats exposed to contaminated soil along with their hematotoxicity and nephrotoxicity. The soil sample is the International Atomic Energy Agency-312 (IAEA-312) certified reference material, whereas blood, bones, and kidneys of in vivo female Sprague-Dawley (SD) rats estimate ²³⁸U- and ²³²Th-f_ab fractions post-exposure. We predict the bioavailable concentration (C_ab) and f_ab values of ²³⁸U and ²³²Th after acute soil ingestion. The blood ²³⁸U (0.750%) and ²³²Th (0.028%) reach their maximum f_ab values after 48 h. The ²³⁸U (f_ab: 0.169-0.652%) accumulates mostly in the kidney, whereas the ²³²Th (f_ab: 0.004-0.021%) accumulates primarily in the bone. Additionally, ²³⁸U is more bioavailable than ²³²Th. Post 48 h acute ingestion demonstrates noticeable histopathological and hematological alterations, implying that intake of ²³⁸U in co-contaminated soil can lead to erythrocytes and proximal tubules damage, whereas, ²³²Th intake can harm erythrocytes. Our study provides new directions for future research into the health implications of acute oral exposures to ²³⁸U and ²³²Th in co-contaminated soils. The findings offer significant insight into the utilization of in vivo SD rat testing to estimate ²³⁸U and ²³²Th bioavailability and toxicity in exposure assessment.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.837-846
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• 2014

A lead slowing down spectrometer (LSDS) is under development for analysis of isotopic fissile material contents in pyro-processed material, or spent fuel. Many current commercial fissile assay technologies have a limitation in accurate and direct assay of fissile content. However, LSDS is very sensitive in distinguishing fissile fission signals from each isotope. A neutron spectrum analysis was conducted in the spectrometer and the energy resolution was investigated from 0.1eV to 100keV. The spectrum was well shaped in the slowing down energy. The resolution was enough to obtain each fissile from 0.2eV to 1keV. The detector existence in the lead will disturb the source neutron spectrum. It causes a change in resolution and peak amplitude. The intense source neutron production was designed for ~E12 n's/sec to overcome spent fuel background. The detection sensitivity of U238 and Th232 fission chamber was investigated. The first and second layer detectors increase detection efficiency. Thorium also has a threshold property to detect the fast fission neutrons from fissile fission. However, the detection of Th232 is about 76% of that of U238. A linear detection model was set up over the slowing down neutron energy to obtain each fissile material content. The isotopic fissile assay using LSDS is applicable for the optimum design of spent fuel storage to maximize burnup credit and quality assurance of the recycled nuclear material for safety and economics. LSDS technology will contribute to the transparency and credibility of pyro-process using spent fuel, as internationally demanded.