• Nuclear Engineering and Technology
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• v.31 no.4
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• pp.365-374
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• 1999

A nuclear design of fission moly production targets for a research reactor, HANARO was peformed. It was found that the use of MCNP-4A, ORIGEN-2 code was reliable for the analysis of production characteristics of $^{99}$ Mo in a target fuel at an irradiation holes. A parametric study was done for the optimization of target location, target dimension, target shape and fuel materials. It was shown that a fuel thickness was the most sensitive parameters and electro-deposited target gave the highest 99Mo yield ratio. A pellet target with vibro-compaction powder, however, showed the largest production capacity and better engineering feasibility even with less yield ratio. Ten kinds of optimized target design for both LEU and HEU satisfied all the given design constraints. The most favorable design was the HEU ring-shaped electro-deposited target, considered the safety limit, production yield, chemical process easiness, yield ratio, and amount of radioactive waste.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.734-743
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• 2017

In order to investigate the swelling behavior and fuel-cladding interaction mechanism of U-10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel-cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal-hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2005.11a
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• pp.25-26
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• 2005

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.220.1-220.1
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• 2005

• Korean Journal of Materials Research
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• v.10 no.2
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• pp.166-170
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• 2000

Nuclear fuel $UO_{2+x}$ power was produced from concentrated $UF_6$ by the DCP(Dry Conversion Process). The characterstics of $UO_{2+x}$ powder, prepared with respect to steam flowing conditions and temperature variations in a rotary kiln reactor, have been investigated with a uranium analyzer, water vapor measurement, and SEM. Fluorine content of the powder could be reduced to 8ppm. The moisture content was found to be optimized.

• 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.

• Nuclear Engineering and Technology
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• v.29 no.3
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• pp.196-200
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• 1997

The variation of sintered density and fain size in ex-AUC, ex-ADU and granulated ex-ADU UO$_2$ pellets in which 0.1~1.0wt% Nb$_2$O$_{5}$ were doped were examined. Pellets were sintered in an atmosphere of H$_2$ at 1$700^{\circ}C$ for 4h. All the specimens tested shooed more than 94% T.D.(Theoretical Density). Sintered density decreased with increasing the amount of Nb$_2$O$_{5}$. Powder types had little influence on the sintered density. Pore size distribution was shifted to the larger ones as Nb$_2$O$_{5}$ was added. The increase of total pore volume and grain growth due to the addition of Nb$_2$O$_{5}$ were thought to be the cause of the sintered density decrease. The largest grain size was seen in the 1. 0wt% Nb$_2$O$_{5}$ doped ex-AUC UO$_2$ pellets. Their average size was 13.9 ${\mu}{\textrm}{m}$.m}$.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3525-3534
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• 2023

Non-uniformity of Pu and Cm composition in used nuclear fuel was analyzed to determine its effect on Pu accountancy in pyroprocessing, while employing the Pu-to-²⁴⁴Cm ratio method. Burnup simulation of a typical pressurized water reactor fuel assembly, required for the analysis, was carried out using MCNP code. Used fuel nuclide composition, as a function of nine axial and two radial meshes, were evaluated. The axial variation of neutron flux and self-shielding effects were found to affect the uniformity of Pu and Cm compositions and in turn the Pu-to-²⁴⁴Cm ratio. However, the results of the study showed that these non-uniformities do not affect the use of Pu-to-²⁴⁴Cm ratio method for Pu accountancy, if the measurement samples are drawn from the voloxidized powder at the feed step of pyroprocessing. 'Material Unaccounted For' and its uncertainty estimates are also presented for a pyrprocessing facility to verify safeguards monitoring requirements of the IAEA.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.899-906
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• 2018

The irradiation performance of $U_3Si$ dispersion fuel in an Al matrix, $U_3Si-Al$, under the Hi-Flux Advanced Neutron Application Reactor (HANARO) design full-power condition of 30 MW was tested for full-power qualification of the fuel. A test assembly was fabricated containing 18 fuel rods made with atomized $U_3Si$ powder manufactured at the Korea Atomic Energy Research Institute. The test assembly was irradiated for 188 full-power operation days in the HANARO subject to the normal fuel-loading scheme and achieved about 60 at% U-235 average burnup and 75 at% U-235 peak burnup. The maximum linear power of the test assembly was 98 kW/m. Nondestructive and destructive postirradiation examinations were conducted. The measured postirradiation examination data were compared with data from previous irradiations and the design criteria required for HANARO fuel. Consequently, it was concluded that in-pile performance was acceptable and fuel integrity was maintained, and the behavior satisfied the fuel design requirements.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.23-32
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• 2022

Pyroprocessing is a promising technology for managing spent nuclear fuel. The nuclear material accounting of feed material is a challenging issue in safeguarding pyroprocessing facilities. The input material in pyroprocessing is in a solid-state, unlike the solution state in an input accountability tank used in conventional wet-type reprocessing. To reduce the uncertainty of the input material accounting, a double-stage homogenization process is proposed in considering the process throughput, remote controllability, and remote maintenance of an engineering-scale pyroprocessing facility. This study tests two types of mixing equipment in the proposed double-stage homogenization process using surrogate materials. The expected heterogeneity and accounting uncertainty of Pu are calculated based on the surrogate test results. The heterogeneity of Pu was 0.584% obtained from Pressurized Water Reactor (PWR) spent fuel of 59 WGd/tU when the relative standard deviation of the mass ratio, tested from the surrogate powder, is 1%. The uncertainty of the Pu accounting can be lower than 1% when the uncertainty of the spent fuel mass charged into the first mixers is 2%, and the uncertainty of the first sampling mass is 5%.