• 방사선산업학회지
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• 제17권4호
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• pp.457-469
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• 2023

The clearance level by nuclide is announced by the Nuclear Safety and Security Commission. However, the clearance level of uranium existing in nature has not been announced, and research is needed. Therefore, the purpose of this study was to evaluate the clearance level of uranium nuclides appropriate to domestic conditions preliminary. For this purpose, this study selected major processes for recycling metal wastes and analyzed the exposure scenarios and major input factors by investigating the characteristics of each process. Then, the radiation dose to the general public and workers was evaluated according to the selected scenarios. Finally, the results of the radiation dose per unit radioactivity for each scenario were analyzed to derive the clearance level of uranium in metal waste. The results of the radiation dose assessment for both the general public and workers per unit radioactivity of uranium isotopes were shown to meet the allowable dose (individual dose of 10 µSv y^-1 and collective dose of 1 Man-Sv y^-1) regulated by the Nuclear Safety and Security Commission. The most conservative scenarios for volumetric and surface contamination were evaluated for the handling of the slag generated after the melting of the metal waste and the direct reuse of the contaminated metal waste into the building without further disposal. For each of these scenarios, the radioactivity concentration by uranium isotope was calculated, and the clearance level of uranium in metal waste was calculated through the radioactivity ratio by enrichment. The results of this study can be used as a basic data for defining the clearance level of uranium-contaminated radioactive waste.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.6-16
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• 2017

This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the $2{\times}2$ checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

• Mass Spectrometry Letters
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• 제3권2호
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• pp.54-57
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• 2012

Mass spectrometric analysis was carried out using multicollector inductively coupled plasma mass spectrometry (MC-ICP-MS) for the precise and accurate determination of the isotope ratios of ultratrace levels of uranium dissolved in 3% $HNO_3$. We used the certified reference material (CRM) 112-A at a trace level of 100 pg/mL for the uranium isotopic measurement. Multiple collectors were utilized for the simultaneous measurement of uranium isotopes to reduce the signal uncertainty due to variations in the ion beam intensity over time. Mass bias correction was applied to the measured U isotopes to improve the precision and accuracy. Furthermore, elemental standard solution with certified values of platinum, iridium, gold, and thallium dissolved in 3% $HNO_3$ were analyzed to investigate the formation rates of the polyatomic ions of $Ir^{40}$ $Ar^+$, $Pt^{40}$ $Ar^+$, $Tl^{40}$ $Ar^+$, $Au^{40}$ $Ar^+$ for the concentration range of 50-400 pg/mL. Those polyatomic ions have mass-to-charge ratios in the 230-245 m/z region that it would contribute to the increase of background intensity of uranium, thorium, plutonium, and americium isotopes. The effect of the polyatomic ion interference on uranium isotope measurement has been estimated.

• 한국표면공학회지
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• 제32권2호
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• pp.165-171
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• 1999

The etching reaction of $UO_2$ in $CF_4/O_2$ gas plasma is examined as functions of $CF_4/O_2$ ratio, plasma power, and substrate temperature at up to $370^{\circ}C$ under the total pressure of 0.30 Torr. It is found that the highest etching rate is obtained at 20% $O_2$ mole fraction, regardless of r. f. power and substrate temperature. The existence of the optimum $CF_4/O_2$ ratio is confirmed by SEM, XPS and XRD analysis. The highest etching reaction rate at $370^{\circ}C$ under 150W exceeds 1000 monolayers/min., which is equivalent to 0.4$\mu\textrm{m}$/min. The mass spectrometry analysis results reveal that the major reaction product is uranium hexa-fluoride $UF_6$. Based on the experimental findings, dominant overall reaction of uranium dioxide in $CF_4/O_2$ plasma is determined : $8UO_2+12CF_4+3O_2=8UF_6+12CO_{2-x}$ where $CO_{2-x}$ represents the undetermined mix of $CO_2$ and CO.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.1-12
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• 2019

One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than $UO_2$ pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of $^{14}N$. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to $^{235}U$ enrichment ratio compared to the $UO_2/Zr$ fuel.

• Nuclear Engineering and Technology
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• 제6권2호
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• pp.80-88
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• 1974

Amiel의 지발중성자 계측법을 사용한 분석과정을 다소 수정하여 지질학적 시료중의 우라륨 및 토륨의 동시 정량을 시도하였다. 본 분석과정은 넓은 범위의 두원소의 함량비에서 정화하게 적용할 수 있었다. 개발과정중 수행한 세력사항을 기술하였으며 나아가 본 분석법을 평가하였던 바 우라늄의 감도는 0.1$\mu\textrm{g}$이하였고 토륨의 감도는 약 5$\mu\textrm{g}$이었다.

• 자원리싸이클링
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• 제30권4호
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• pp.64-74
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• 2021

원자력 발전 프로그램에 의해 세계 우라늄 메탈의 수요가 증가함에 따라 우라늄 메탈의 중요도는 날로 증가하고 있다. 이러한 우라늄 수요의 높은 증가로 인해 주로 바나듐을 비롯한 다른 원소들로부터 우라늄을 회수하고자 하는 새로운 추출법들이 더욱 중요하게 되었다. 본 연구에서는 kerosene에 희석된 상업용 아민 기반 추출제를 사용하여 황산 용액에서 우라늄(VI)과 바나듐(V)을 분리하는 방법을 다루었다. 0.005 ~ 5.0 mol/L농도의 황산 용액과 0.005 ~ 0.2 mol/L농도의 추출제를 사용하였다. 모든 실험은 25℃에서 30분동안 동일한 수상과 유기상의 비율(A:O = 1:1)로 수행하였다. 실험 결과 값으로부터 계산된 분리 계수(SF's)를 제시하고 모든 실험 값과 비교하였다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.124-129
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• 1998

A sensitivity study has been done to determine the composition of DUPIC fuel from the viewpoint of neutronics fuel design. The spent PWR fuel compositions were generated and fissile contents adjusted by blending fresh uranium after mixing two spent PWR fuel assemblies. The $^{239}$ Pu and $^{235}$ U enrichments of DUPIC fuel were adjusted by controlling the amount of fresh uranium feed and the ratio of slightly enriched and depleted uranium in the fled uranium. Based on the material balance calculation, it is recommended that DUPIC fuel composition be such that spent PWR fuel utilization is more than 90%.. A sensitivity study on the temperature reactivity coefficient of DUPIC fuel has shown that it is desirable to increase the $^{239}$ Pu and $^{235}$ U contents to reduce both the fuel and coolant temperature coefficients. On the other hand, refueling simulations of the DUPIC core have shown that the channel power peaking factor, which is a measure of the reactor trip margin, increases with the total fissile content. Considering these neutronic characteristics of the DUPIC fuel, il is recommended to have enrichments of 0.45 and 1.00 wt% for $^{239}$ Pu and $^{235}$ U, respectively.

• Nuclear Engineering and Technology
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• 제51권2호
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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.

• 방사성폐기물학회지
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• 제18권4호
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• pp.457-464
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• 2020

In this study, extraction of uranium(VI) from an aqueous nitric acid solution was investigated using tri-n-butyl phosphate (TBP) as an extractant in an ionic liquid, 1-alkyl-3-methylimidazolium bis (trifluoromethylsulfonyl)imide ([Cnmim][Tf2N]). The distribution ratio of U(VI) in 1.1 M TBP/[Cnmim][Tf2N] was significantly high when the concentration of nitric acid was low. The value of the distribution ratio decreased as the concentration of the nitric acid increased at lower acidities, and then increased with a nitric acid concentration of up to 8 M. This can be attributed to the different extraction mechanisms of U(VI) based on nitric acid concentrations. Thus, a cation exchange at low acidity levels and an ion-pair extraction at high acidity levels were suggested as the extraction mechanism of U(VI) in the TBP/[Cnmim][Tf2N] system.