• Journal of the Korean Chemical Society
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• v.31 no.4
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• pp.315-321
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• 1987

The ligand, 2,10-dibenzyl-4,6,8-trioxo-3,9-diaza undecane dioic acid(DTDA) for the extraction of uranium was synthesized under dry nitrogen from phenylalanine and 3-oxoglutaric acid. Extraction was performed by stirring a solution of DTDA in dichloromethane for 1 hour with an aqueous solution of $UO_2(ClO_4)_2{\cdot}6H_2O$ at various pH values and at different $DTDA/UO_2{^{2+}}$ molar ratios. Extraction efficiency reaches a maximum when the pH of the aqueous phase was ca 8.0. The extraction percentage was affected by concentration of DTDA and increases with the $DTDA/UO_2{^{2+}}$ molar ratio to complete extraction with a 4 fold excess of DTDA. The high selectivity of the DTDA for uranium was ascertained by competition experiments with other cations. The bound uranyl ion was quantitatively liberated within few minutes from the organic phase by treatment with an aqueous 1M HCI solution and DTDA was recovered very satisfactorily from the organic phase. The values of the over-all formation constants of the complex between uranyl ion and DTDA were determined to be the following : ${\beta}_1=1.20{\times}10^5\;,\;{\beta}_2=1.01{\times}10^8$.

• Nuclear Engineering and Technology
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• v.17 no.2
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• pp.116-128
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• 1985

Considering vacancy generation and migration in grain and sink at grain boundary, a mechanistic densification model which is dependent on UO$_2$ temperature and microstructure has been developed. This densification model is a function of time, fission rate, temperature, density, pore size distribution and grain size. The resultant equation derived in this model which is different from Assmann and Stehle's resultant equations for four temperature regions, can be applied directly for all the pellet temperatures. The predictions of the present densification model very well agreed with the experimental data. This model well predicts absolute magnitude and trend in comparison with the empirical algorithm used in KFEDA code.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.951-960
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• 2013

Energy group structure has a significant effect on the results of multigroup transport calculations. It is known that $UO_2-PuO_2$ (MOX) is a recently developed fuel which consumes recycled plutonium. For such fuel which contains various resonant nuclides, the selection of energy group structure is more crucial comparing to the $UO_2$ fuels. In this paper, in order to improve the accuracy of the integral results in MOX thermal lattices calculated by WIMSD-5B code, a swarm intelligence method is employed to optimize the energy group structure of WIMS library. In this process, the NJOY code system is used to generate the 69 group cross sections of WIMS code for the specified energy structure. In addition, the multiplication factor and spectral indices are compared against the results of continuous energy MCNP-4C code for evaluating the energy group structure. Calculations performed in four different types of $H_2O$ moderated $UO_2-PuO_2$ (MOX) lattices show that the optimized energy structure obtains more accurate results in comparison with the WIMS original structure.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.29 no.1
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• pp.62-67
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• 2016

In this paper, the IES file was measured by applying a secondary optical lens to a 21 W LED engine, and the lighting calculation software RELUX was used to perform simulations with the data file of this measurement. For two-lane (two way) concrete paved roads, six LED engine are applied to each streetlight and simulation results show that Uo (uniformity) 0.56, UI (longitudinal uniformity) 0.86 and TI (threshold iIncrement) 9% which satisfies the required standards. RELUX was also used to LED streetlights by designing them in three dimensions, that is ${\pm}25%$ of the arm length of 2.8 m standardized by the road lighting standards of the Korea Expressway Corporation. Comparative analysis was carried out on adjustments were made in increments of 0.1 m that Uo, UI, and TI values in the range of arm lengths from 2.1 m~3.5 m. For the arm length range of 2.1 m~2.4 m, Uo was high, whereas UI was low. Therefore, we present the optimal light distribution values designed for an arm length of 2.5 m.

• Nuclear Engineering and Technology
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• v.35 no.4
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• pp.299-308
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• 2003

Fabrication technique of $(Th,U)O_2$ pellets has been investigated. Powder mixtures of $ThO_2\;and\;UO_2$ were milled in two different ways-dry and wet milling. Milled powder was compacted and sintered to $(Th,U)O_2$ pellets. The wet-milled powder leads to a $(Th,U)O_2$ pellet having a high sintered density and uniform distribution of U and Th, compared to the dry-milled powder. The sintered density of a $(Th,U)O_2$ pellet tends to decrease by increasing the content of $ThO_2$. The thermal conductivity of $ThO_2\;and\;(Th,U)O_2$ pellets was measured by the laser flash method. The thermal conductivity of the $ThO_2$ pellet is higher than that of the $UO_2$ pellet, and the thermal conductivities of $(Th,U)O_2$ pellets containing $65wt\%\;and\;35wt\%\;ThO_2$ pellets are lower than that of the $UO_2$ pellet.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.923-928
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• 2018

We developed a compact gamma camera based on a modified uniformly redundant array coded aperture to investigate the position of a $UO_2$ pellet emitting characteristic X-rays (98.4 keV) and ${\gamma}-rays$ (185.7 keV). Experiments using an only-mask method and an antimask subtractive method were conducted, and the maximum-likelihood expectation maximization algorithm was used for image reconstruction. The images obtained via the antimask subtractive method were compared with those obtained using the only-mask method with regard to the signal-to-noise ratio. The reconstructed images of the antimask subtractive method were superior. The reconstructed images of the characteristic X-rays and the ${\gamma}-rays$ were combined with the obtained image using the optical camera. The combined images showed the precise position of the $UO_2$ pellet. According to the self-absorption ratios of the nuclear material and the minimum number of effective events for image reconstruction, we estimated the minimum detection time depending on the amount of nuclear material.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.344-357
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• 2002

The present paper reports spontaneous steam explosions observed in fuel coolant interaction experiments using prototypic reactor materials. Pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$ are used. A high temperature molten material in the form of a jet is poured into a subcooled water pool located in a pressure vessel. An induction skull melting technique is used for the melting of the reactor material. In both tests using pure ZrO$_2$ and a mixture of UO$_2$ and ZrO$_2$, either a quenching or a spontaneous steam explosion was observed. The morphology of debris and pressure profile clearly indicate the differences between the qunching cases and explosion cases. The dynamic pressure. dynamic impulse, water temperature, melt temperature, and static pressure Inside the containment chamber were measured . As the spontaneous steam explosion for the reactor material is firstly observed in the present experiments, the results of present experiments could be a siginificant step forward the understanding the explosion of the reactor material.

• Resources Recycling
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• v.11 no.4
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• pp.3-10
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• 2002

The characteristics of dry and wet milled powder prepared by 1 cycle OREOX (oxidation and reduction of oxide fuels) treatment were investigated using the simulated spent fuel pellet. Sintered pellets simulating spent nuclear fuel burned in reactor were fabricated from $UO_2$ powder using as a starting material in fabrication of nuclear fuel. The 1 cycle OREOX-treated powder was prepared by only one path of oxidation md reduction of the simulated pellet. Powder having average particle size of less than 1 $\mu\textrm{m}$ could be easily obtained by dry milling, but not be achieved by wet milling. And, specific surface area of dry milled pow-der was higher than that of wet milled powder. Dry milled powder formed loose agglomerate, while wet milled powder showed the shape of irregular and angular particles. Dry milled powder provided higher green density, resulting in higher sintered density of higher than 95% TD and average grain size of larger than 8 $\mu\textrm{m}$ satisfying the standard specification of sintered pellets.

• Nuclear Engineering and Technology
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• v.33 no.4
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• pp.386-396
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• 2001

Mixtures Of AUC-UO$_2$and Gd$_2$O$_3$ Powders doped With Cr$_2$O$_3$ or Cr$_2$O$_3$-SiO$_2$ were Pressed and sintered at 1730 t in hydrogen gas witk various water-vapor contents. The density of UO$_2$- 6wt% Gd$_2$O$_3$ pellets can be increased from 91% TD to 94.5% TD in 1 vol% $H_2O$-H$_2$ gases by the addition of 0.02wt% Cr$_2$O$_3$-(0.01~0.04) wt% SiO$_2$. The magnitude of density increase is much larger in (1~3 vol%) $H_2O$-H$_2$ gases than in 0.05 vol% $H_2O$-H$_2$ gas. The densification of U0$_2$- Gd$_2$O$_3$ compact is significantly delayed in the temperature range between 1300 and 1500 t , but that of compacts with Cr$_2$O$_3$-SiO$_2$ is not. The role of Cr$_2$O$_3$ and SiO$_2$ in densification is discussed.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.91-100
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• 2005

Kyung-hee Thorium Fuel (KTF), a heterogeneous thorium-based seed and blanket design concept for pressurized light water reactors, is being studied as an alternative to enhance proliferation resistance and fuel cycle economics of PWRs. The proliferation resistance characteristics of the KTF assembly design were evaluated through parametric studies using neutronic performance indices such as Bare Critical Mass (BCM), Spontaneous Neutron Source rate (SNS), Thermal Generation rate (TG), and Radio-Toxicity. Also, Fissile Economic Index (FEI), a new index for gauging fuel cycle economy, was suggested and applied to optimize the KTF design. A core loaded with optimized KTF assemblies with a seed-to-blanket ratio of 1: 1 was tested at the Korea Next Generation Reactor (KNGR), ARP-1400. Core design characteristics for cycle length, power distribution, and power peaking were evaluated by HELIOS and MASTER code systems for nine reload cycles. The core calculation results show that the KTF assembly design has nearly the same neutronic performance as those of a conventional $UO_2$ fuel assembly. However, the power peaking factor is relatively higher than that of conventional PWRs as the maximum Fq is 2.69 at the M$9^{th}$ equilibrium cycle while the design limit is 2.58. In order to assess the economic potential of a heterogeneous thorium fuel core, the front-end fuel cycle costs as well as the spent fuel disposal costs were compared with those of a reference PWR fueled with $UO_2$. In the case of comprising back-end fuel cycle cost, the fuel cycle cost of APR-1400 with a KTF assembly is 4.99 mills/KWe-yr, which is lower than that (5.23 mills/KWe-yr) of a conventional PWR. Proliferation resistance potential, BCM, SNS, and TG of a heterogeneous thorium-fueled core are much higher than those of the $UO_2$ core. The once-through fuel cycle application of heterogeneous thorium fuel assemblies demonstrated good competitiveness relative to $UO_2$ in terms of economics.