• 산업기술연구
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• 제21권A호
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• pp.273-278
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• 2001

In this study, we prepared nano-sized $TiO_2$ particles for various process variables by the diffusion flame reactor and we collected $TiO_2$ particles by thermophoresis. It is found that the size of $TiO_2$ particles increases, as the flame temperature or the inlet $TiCl_4$ concentration increase or the total gas flow rate decreases. We investigated the photo-degradation of phenol wish the prepared $TiO_2$ particles. We found the optimum amounts of $TiO_2$ photocatalysts for our experimental apparatus and investigated the photo-degradation efficiencies of phenol, changing the process variables such as size of $TiO_2$ photocatlysts, phase ratio of rutile/anatase, concentration of phenol, input ratio of $O_2$. Degradation efficiencies of phenol were almost 95% in 15 minutes for the standard conditions of our experiments.

• 한국물환경학회지
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• 제32권4호
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• pp.343-348
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• 2016

A chemical sequencing batch reactor was operated to test the feasibility of nutrient recovery from a biological livestock wastewater treatment plant. Both phosphate and ammonia could be successfully recovered as magnesium ammonium phosphate (MAP) crystals. The contents of TP and TN in the recovered MAP crystals were 26.2% and 4.0%, respectively. Zn, Cr and Ti were identified in the crystals, but the contents remained below the Korean standard for an organic fertilizer. Chemical analyses confirmed that the MAP crystals could be useful phosphate fertilizers. On the other hand, the results of physical analyses using an X-ray diffractometer and an energy dispersive X-ray spectrometer strongly suggested that crystalline materials like magnesium potassium phosphate (KMP) and hydroxyapatite (HAP) were also formed during the MAP crystallization, depending on the availability of K⁺ and Ca²⁺.

• 전기전자학회논문지
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• 제18권4호
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• pp.566-571
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• 2014

A safety class field programmable gate array based system in nuclear power plant has been developed to improve the diversity. Testbench is necessary to satisfy the technical reference, IEC-62566, for verification and validation of register transfer level code. We use the open verification methodology(OVM) developed by standard body. We show that our testbench can use random input for test. And also we show that reusability of block level testbench for the integration level testbench, which is very efficient for large scale system like nuclear reactor protection system.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2002년도 춘계 학술대회논문집
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• pp.79-87
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• 2002

The present work analyzed the effect of mixing vane shape on the flow structure and heat transfer downstream of mixing vane in a subchannel of fuel assembly, by obtaining velocity and pressure fields, turbulent intensity, flow-mixing factors, heat transfer coefficient and friction factor using three-dimensional RANS analysis. NJl5, NJ25, NJ35, NJ45, which were designed by the authors, were tested to evaluate the performances in enhancing the heat transfer. Standard $\kappa-\epsilon$ model is used as a turbulence closure model, and, periodic and symmetry conditions are set as boundary conditions. The flow blockage ratio is kept constant, but the twist angle of mixing vane is changed. The results with three turbulence models( $\kappa-\epsilon$, $\kappa-\omega$, RSM) were compared with experimental data.

• Nuclear Engineering and Technology
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• 제45권2호
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• pp.165-178
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• 2013

A comprehensive review on the technical standards about human factors (HF) design and software reliability maintenance for digital instrumentation and control (I&C) and human-machine interface technology (HMIT) in Japanese light water reactor nuclear power plants (NPPs) was given in this paper mainly by introducing the relevant activities at the Japan Electric Association to set up many industrial standards within the traditional framework of nuclear safety regulation in Japan. In Japan, the Fukushima Daiichi accident that occurred on March 11, 2011 has great impact on nuclear regulation and nuclear industries where concerns by the general public about safety have heightened significantly. However for the part of HF design and software reliability maintenance of digital I&C and HMIT for NPP, the author believes that the past practice of Japanese activities with the related technical standards can be successfully inherited in the future, by reinforcing the technical preparedness for the prevention and mitigation against any types of severe accident occurrence.

• Nuclear Engineering and Technology
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• 제28권3호
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• pp.331-338
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• 1996

In the thermal-hydraulic design, the effect of fuel rod bow is quantified tv the rod bow DNBR penalty which is a key design parameter to assure the coolability of fuel assembly in the pressurized water reactor. In this work, a computer program for the evaluation of the rod bow DNBR penalty based on Westinghouse methodology is developed and its application procedure is proposed. The computer simulation is based on the Monte-Carlo method. The qualification of developed computer program is performed by a comparison of calculational result with that given by Westinghouse's document. A new application procedure is built using batch mean and batch standard deviation. The normality of sample population generated by the batch calculation is confirmed by means of a chi-square test for goodness of fit. On the view point of statistics it is effected that the more reliable design value may be produced by the new application procedure.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 1999년도 추계 학술발표회 논문집 Proceedings of EESK Conference-Fall
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• pp.137-146
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• 1999

The SSI(Soil-Structure Interaction) analyses are being performed for the KNGR(Korean Next Generation Reactor) design because the KNGR is developed as a standard nuclear power plant concept enveloping various soil conditions. the SASSI program which adopts the flexible volume method is used for the SSI analyses. The soil curves used in the three dimensional SSI analyses of KNGR Nuclear Island(NI) structures are based on the upper bound shear modulus curve and lower bound damping degradation on SSI response the average shear modulus curve with average damping curve was used for two soil cases. This study presents the results of the variances by using different soil nonlinearity parameters based on the paametric SSI analyses. The results include the maximum member forces(shear and axial force) at the base of the NI structures and the 5% damping Floor Response Spectra (FRS) at some representative locations at the top of the NI superstructures. They are also compared together with the enveloped SSI results for eight soil cases and fixed-base analysis for rock case by using two control motions.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1349-1354
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• 2018

Inductively Coupled Plasma Atomic Emission Spectroscopy (ICP-AES) is widely used for the determination of trace elements in geological samples in Iran. In this paper, we have calculated the detection limits of neutron activation analysis (NAA) for some of the common elements in such samples utilizing the ORIGEN and MCNP codes and verified the simulations using the experimental results of three soil standard reference materials, namely, G02.SRM, G18.SRM, and G28.SRM. The results show that while the detection limit of ICP-AES method is usually in the mg/kg range, it is represented to the ${\mu}g/kg$ range for most of the elements of interest using the NAA method, and the simulations can be verified in a tolerance range of 20%.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2803-2811
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• 2020

With the aging of nuclear reactors, embrittlement of the reactor pressure vessel (RPV) steel, as a consequence of routine operations, is highly probable. To ensure operational integrity and safety, prediction and mitigation of compromising damage, brought on by pressurized thermal shock (PTS) following an emergency procedure, is of utmost importance. Computational fluid dynamics (CFD) codes can be employed to predict these events and have therefore been an acceptable method for such assessments. In this paper, CFD simulations of a density driven ECC state in the ROCOM facility are analyzed. Obtained numerical results are validated with the experimental measurements. Considerable attention is attributed to the boundary conditions and their influence, specifically outlet definitions, in order to determine and adequately replicate the non-active pumps in the facility. Consequent analyses focused on initial conditions as well as on the temporal discretization and inner iterations. Disparities due to different turbulent modelling approaches are investigated for standard RANS models. Based on observed trends for different cases, a definitive simulation setup has been established, results of which have been ultimately compared to the measurements.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.410-418
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• 2024

Study of flow behavior within rod bundles has been an active topic. Surface modification technologies are important parts of the design of the fourth generation reactor, which can increase the strength of the secondary flow within the rod bundle lattices. Quasi-periodic large-scale vortex structure (QLVS) is introduced by arranging micro ribs on the surface of rod bundles, which enhanced the scale of the secondary flow between the rod bundle lattices. Using computational fluid dynamics (CFD) and water experiments, the flow field distribution and drag coefficient of the rod-bundle lattices are studied. The secondary flow between the micro-ribbed rod-bundle lattice is significantly enhanced compared to the standard rod-bundle lattice. The numerical simulation results agree well with the experimental results.