• Journal of Powder Materials
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• v.15 no.2
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• pp.87-94
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• 2008

In order to investigate a nitriding process of spent oxide fuel and the subsequent change in thermal properties after nitriding, simulated spent fuel powder was converted into a nitride pellet with simulated fission product elements through a carbothermic reduction process. Nitriding rate of simulated spent fuel was decreased with increasing of the amount of fission products. Contents of Ba and Sr in simulated spent fuel were decreased after the carbothermic reduction process. The thermal conductivity of the nitride pellet was decreased by an addition of fission product element but was higher than that of the oxide fuel containing fission product elements.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1208-1212
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• 2020

Knowledge on fuel enthalpy and its temperature derivative, the heat capacity, are important quantities in determination of fuel behavior in normal reactor operation and reactor transients. The aim of this study is to compare the heat capacity of oxide and nitrite fuels by using Einstein-Debye approximation. A simple analytical expression was performed to calculate the heat capacity of fuels. To test the validity and reliability, the calculated formulas were compared to published results for various nuclear fuels including UO₂, ThO₂, PuO₂ and UN. Calculated formulas yielded results in consistent with literature.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.1097-1104
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• 2023

A preliminary criticality analysis for novel pyrochemical apparatuses for the reprocessing of mixed uranium-plutonium nitride spent nuclear fuel from the BREST-OD-300 reactor was performed. High-temperature processing apparatuses, "metallization" electrolyzer, refinery remelting apparatus, refining electrolyzer, and "soft" chlorination apparatus are considered in this work. Computational models of apparatuses for two neutron radiation transport codes (MCU-FR and MCNP) were developed and calculations for criticality were completed using the Monte Carlo method. The criticality analysis was performed for different loads of fissile material into the apparatuses including overloading conditions. Various emergency situations were considered, in particular, those associated with water ingress into the chamber of the refinery remelting apparatus. It was revealed that for all the considered computational models nuclear safety rules are satisfied.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.489-494
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• 1997

A comparative transient analyses were performed for oxide and nitride cores or a large (3000 MWt), pool-type, liquid-metal-cooled reactor (LMR). The study was focused on three representative accident initiators with failure to scram : the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected fast transient overpower (UFTOP). The margins to fuel melting and sodium boiling have been evaluated for these representative transients. The results show that there is an increase in safety margin with nitride core which maintains the physical dimensions of the oxide core.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.68-68
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.133-134
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• 2005

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.660-672
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• 2016

In Part I of this paper, the two-temperature homogenized model for the fully ceramic microencapsulated fuel, in which tristructural isotropic particles are randomly dispersed in a fine lattice stochastic structure, was discussed. In this model, the fuel-kernel and silicon carbide matrix temperatures are distinguished. Moreover, the obtained temperature profiles are more realistic than those obtained using other models. Using the temperature-dependent thermal conductivities of uranium nitride and the silicon carbide matrix, temperature-dependent homogenized parameters were obtained. In Part II of the paper, coupled with the COREDAX code, a reactor core loaded by fully ceramic microencapsulated fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure is analyzed via a two-temperature homogenized model at steady and transient states. The results are compared with those from harmonic- and volumetric-average thermal conductivity models; i.e., we compare $k_{eff}$ eigenvalues, power distributions, and temperature profiles in the hottest single channel at a steady state. At transient states, we compare total power, average energy deposition, and maximum temperatures in the hottest single channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized model for Doppler temperature feedback lead to significant differences.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.847-858
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• 2013

Since 2001, a series of five irradiation test campaigns for atomized U-Mo dispersion fuel rods, KOMO-1, -2, -3, -4, and -5, has been conducted at HANARO (Korea) in order to develop high performance low enriched uranium dispersion fuel for research reactors. The KOMO irradiation tests provided valuable information on the irradiation behavior of U-Mo fuel that results from the distinct fuel design and irradiation conditions of the rod fuel for HANARO. Full size U-Mo dispersion fuel rods of 4-5 $g-U/cm^3$ were irradiated at a maximum linear power of approximately 105 kW/m up to 85% of the initial U-235 depletion burnup without breakaway swelling or fuel cladding failure. Electron probe microanalyses of the irradiated samples showed localized distribution of the silicon that was added in the matrix during fuel fabrication and confirmed its beneficial effect on interaction layer growth during irradiation. The modifications of U-Mo fuel particles by the addition of a ternary alloying element (Ti or Zr), additional protective coatings (silicide or nitride), and the use of larger fuel particles resulted in significantly reduced interaction layers between fuel particles and Al.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.495-501
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• 1997

A preliminary and parametric sensitivity study on recriticality risk of fast reactor cores after a hypothetical total core meltdown accident was performed. Only neutronic aspects of the accident were considered, independent of the accident scenario, and efforts were made to estimate the quantity of molten fuel which must be ejected out of the core to assure a sub-critical state after the accident. Two types of parameters were examined : characteristic parameters of molten core such as geometry, molten pool type (homogenized or stratified), fuel temperature, environment, and relative parameters to normal core such as core size(small or large), and fuel type (oxide, nitride, metal). The first type of parameters was found to intervene more directly in the recriticality risk than the second type of parameters.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.724-730
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• 2018

Chromium and chromium nitride were selected as potential barriers to prevent fuel-clad chemical interaction (FCCI) between the cladding and the fuel material. In this study, ferritic/martensitic HT-9 steel and misch metal were used to simulate the reaction between the cladding and fuel fission product, respectively. Radio frequency magnetron sputtering was used to deposit Cr and CrN films onto the cladding, and the gas flow rates of argon and nitrogen were fixed at certain values for each sample to control the deposition rate and the crystal structure of the films. The samples were heated for 24 h at 933 K through the diffusion couple test, and considerable amount of interdiffusion (max. thickness: $550{\mu}m$) occurred at the interface between HT-9 and misch metal when the argon and nitrogen were used individually. The elemental contents of misch metal were detected at the HT-9 through energy dispersive X-ray spectroscopy due to the interdiffusion. However, the specimens that were sputtered by mixed gases (Ar and $N_2$) exhibited excellent resistance to FCCI. The thickness of these CrN films were only $4{\mu}m$, but these films effectively prevented the FCCI due to their high adhesion strength (frictional force ${\geq}1,200{\mu}m$) and dense columnar microstructures.