• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1398-1405
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• 2019

We use a molecular dynamics simulation to explore thermal transport in oxide nuclear fuels with point defects. The effect of vacancy and substitutional defects on the thermal conductivity of plutonium dioxide and uranium dioxide is investigated. It is found that the thermal conductivities of these fuels are reduced significantly by the presence of small amount of vacancy defects; 0.1% oxygen vacancy reduces the thermal conductivity of plutonium dioxide by more than 10%. The missing of larger atoms has a more detrimental impact on the thermal conductivity of actinide oxides. In uranium dioxide, for example, 0.1% uranium vacancies decrease the thermal conductivity by 24.6% while the same concentration of oxygen vacancies decreases the thermal conductivity by 19.4%. However, uranium substitution has a minimal effect on the thermal conductivity; 1.0% uranium substitution decreases the thermal conductivity of plutonium dioxide only by 1.5%.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.751-757
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• 2016

The present work discusses two different models of boiling water reactor (BWR) bundle to compare the neutronic characteristics of uranium dioxide ($UO_2$) and uranium zirconium hydride ($UZrH_{1.6}$) fuel. Each bundle consists of four assemblies. The BWR assembly fueled with $UO_2$ contains $8{\times}8$ fuel rods while that fueled with $UZrH_{1.6}$ contains $9{\times}9$ fuel rods. The Monte Carlo N-Particle Transport code, based on the Mont Carlo method, is used to design three dimensional models for BWR fuel bundles at typical operating temperatures and pressure conditions. These models are used to determine the multiplication factor, pin-by-pin power distribution, axial power distribution, thermal neutron flux distribution, and axial thermal neutron flux. The moderator and coolant (water) are permitted to boil within the BWR core forming steam bubbles, so it is important to calculate the reactivity effect of voiding at different values. It is found that the hydride fuel bundle design can be simplified by eliminating water rods and replacing the control blade with control rods. $UZrH_{1.6}$ fuel improves the performance of the BWR in different ways such as increasing the energy extracted per fuel assembly, reducing the uranium ore, and reducing the plutonium accumulated in the BWR through burnup.

• 대한기계학회논문집A
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• 제31권4호
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• pp.472-482
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• 2007

This study deals with the design and implementation results for a high-capacity vol-oxidizer that can convert Uranium Dioxide pellets to $U_3O_8$ powder for up to several tens of kg HM/batch. We developed two versions of the $1^{st}$ vol-oxidizer and the $2^{nd}$ vol-oxidizer. Through an experiment with the $1^{st}$ vol-oxidizer, we deduced some problems concerning the design considerations such as the recovery rate of $U_3O_8$, the oxidation time of the Uranium Dioxide pellets, the exothermic reaction, and the powder dispersion. From the analyses of the drawbacks of the $1^{st}$ vol-oxidizer, we devised some novel items such as a folding type mesh, vibrators, and mixing blades. Also, we used the Stokes and Density ratio Eq. to determine the most reasonable flux for preventing a powder dispersion. Compared with the results of the $1^{st}$ vol-oxidizer, we showed that both the permeability of the $U_3O_8$ powders and the oxidation rate of the Uranium Dioxide pellets of the $2^{nd}$ vol-oxidizer were remarkably increased, and the temperature of the reactor was controlled well in spite of an exothermic reaction. Also, the powder was not entirely dispersed through the outlet of the voloxidizer. The experimental results of this work can help in the design of a novel and efficient vol-oxidizer with a higher capacity.

• Nuclear Engineering and Technology
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• 제13권4호
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• pp.237-244
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• 1981

이산화우라늄중의 미량불순물원소들의 중성자 방사화분석을 위하여 이들 원소들의 사전농축에 음이온교환수지를 사용하였다. 황산에 녹인 우라늄 시료용액의 산도를 pH 2.7에 맞춘다음 음이온교환수지통에 얹었다. 이 음이온교환수지통으로 부터 미량분순물 원소만을 정량적으로 회수할 수 있었다. 이 사전농축법과 알미늄중 미량불순물 원소의 정량을 위하여 개발한 방사화학적인 분리법을 이용하므로서 원자로급 이산화우라늄 중의 21개 미량불순물을 정량할 수 있었다.

• 한국세라믹학회지
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• 제37권9호
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• pp.833-838
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• 2000

기존의 핵연료재료 습식처리 공정 대체를 위한 건식 처리 공정 기초 연구로서 산소 플라즈마 기체에 의한 금속우라늄과 이산화우라늄의 산화 연구를 수행하였다. 연구결과 산소 플라즈마를 사용할 경우 $UO_2$는 40$0^{\circ}C$에서 약 300% 정도, 50$0^{\circ}C$에서는 70% 정도의 산화율 증가가 일어났으며 금속우라늄의 경우에도 35$0^{\circ}C$에서 50% 정도의 증가를 확인할 수 있었다. 이들 산화율은 플라즈마 출력이 증가함에 따라 비례적으로 증가하였는데 이는 출력 증가에 따른 플라즈마내 산소 원자의 발생과 일치하여 이러한 산화율 증가 현상은 플라즈마내 산소 원자가 주도하는 것으로 드러났다. 이들 실험 결과는, 기존의 실험 결과와 길이, 시간에 따라 산화량이 선형적으로 증가하는 것으로 나타나 산소 플라즈마 산화 반응은 표면 반응이 주요 반응이라는 것이 밝혀졌다.

• 분석과학
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• 제21권5호
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• pp.397-402
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• 2008

이산화우라늄을 질산 및 미량의 불산으로 녹인 다음 우라늄 매트릭스로부터 규소를 분리하지 않고 분광광도계(spectrophotometer)로 측정할 수 있는 분석조건을 검토하였다. 분광광도계로 미량의 규소를 측정할 때 미량의 불산이 규소 분석에 미치는 영향을 조사하였으며, 불산의 간섭을 방지하기 위하여 붕산을 사용하였다. 우라늄 용액에서 미량의 불산이 존재할 경우 포화붕산 사용 유무에 따른 규소의 회수율은 각각 $103.3{\pm}0.8$ 및 $76.6{\pm}6.8%$이었다. 포화붕산의 양은 규소의 회수율에 크게 영향을 미치지 않았다. 따라서 본 방법으로 이산화우라늄 분말 중의 불순물로 존재하는 미량의 규소를 분리 과정없이 바로 UV-VIS 분광광도법으로 정량할 수 있었다.

• Advances in materials Research
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• 제2권2호
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• pp.93-98
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• 2013

Objective of this study is to calculate gamma buildup factors for pointed and isotropic gamma sources in depleted uranium, uranium dioxide, natural uranium, tin, water and concrete using MCNP4C code. The thickness of the media ranges from 0.5 to 10 mean-free-path (mfp) and gamma energy ranges from 0.5 to 10 MeV. Owing to the outstanding accuracy of MCNP in calculation involving gamma interaction, results fairly match those reported previously. The maximum relative error is 2%.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.271-271
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• 2002

• Nuclear Engineering and Technology
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• 제49권8호
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• pp.1711-1716
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• 2017

Uranium tetrafluoride ($UF_4$) is the most used nuclear material for producing metallic uranium by reduction with Ca or Mg. Metallic uranium is a raw material for the manufacture of uranium silicide, $U_3Si_2$, which is the most suitable uranium compound for use as nuclear fuel for research reactors. By contrast, ammonium uranyl carbonate is a traditional uranium compound used for manufacturing uranium dioxide $UO_2$ fuel for nuclear power reactors or $U_3O_8-Al$ dispersion fuel for nuclear research reactors. This work describes a procedure for recovering uranium and ammonium fluoride ($NH_4F$) from a liquid residue generated during the production routine of ammonium uranyl carbonate, ending with $UF_4$ as a final product. The residue, consisting of a solution containing high concentrations of ammonium ($NH_4^+$), fluoride ($F^-$), and carbonate ($CO_3^{2-}$), has significant concentrations of uranium as $UO_2^{2+}$. From this residue, the proposed procedure consists of precipitating ammonium peroxide fluorouranate (APOFU) and $NH_4F$, while recovering the major part of uranium. Further, the remaining solution is concentrated by heating, and ammonium bifluoride ($NH_4HF_2$) is precipitated. As a final step, $NH_4HF_2$ is added to $UO_2$, inducing fluoridation and decomposition, resulting in $UF_4$ with adequate properties for metallic uranium manufacture.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1714-1723
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• 2020

This study aims at characterizing the rheological properties of uranium oxide powders for nuclear fuel pellets manufacturing. The flowability of these powders must be compatible with a reproducible filling of press molds. The particle size distribution is known to have an impact on the rheological properties and fine particles (<100 ㎛) are suspected to have a detrimental effect. In this study, the impact of the particle size distribution on the rheological properties of UO₂ powders was quantified, focusing on the influence of fine particles. Two complementary approaches were used. The first approach involved characterizing the powder in a static state: density, compressibility and shear test measurements were used to understand the behavior of the powder when it is transitioned from a static to a dynamic state (i.e., incipient flow conditions). The second approach involved characterizing the behavior of the powder in a dynamic state. Two zones, corresponding to two characteristic behaviors, were demonstrated for both types of measurements. The obtained results showed the amount of fines should be kept below 10 % wt to ensure a robust mold filling operation (i.e., constant mass and production rate).