• Journal of The Korean Astronomical Society
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• v.47 no.4
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• pp.159-161
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• 2014

The jet production efficiency of radio galaxies can be quantified by comparison of their kinetic jet powers $P_{jet}$ and Bondi accretion powers $P_B$. These two parameters are known to be related linearly, with the jet power resulting from the Bondi power by multiplication with an efficiency factor of order 1%. Using a recently published (Nemmen & Tchekhovskoy 2014) high-quality sample of 27 radio galaxies, I construct a $P_B$ - $P_{jet}$ diagram that includes information on optical AGN types as far as available. This diagram indicates that the jet production efficiency is a function of AGN type: Seyfert 2 galaxies seem to be systematically (with a false alarm probability of $4.3{\times}10^{-4}$) less efficient, by about one order of magnitude, in powering jets than Seyfert 1 galaxies, LINERs, or the remaining radio galaxies. This suggests an evolutionary sequence from Sy 2s to Sy 1s and LINERs, controlled by an interplay of jets on the one hand and dust and gas in galactic nuclei on the other hand. When taking this effect into account, the $P_B$ - $P_{jet}$ relation is probably much tighter intrinsically than currently assumed.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.291-298
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• 2014

This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor ($k_{eff}$), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

• Proceedings of the Korean Reliability Society Conference
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• 2011.06a
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• pp.319-328
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• 2011

The failure of LCC analysis is recognized as a serious risk for companies in fast-paced business environment. LCC analysis has been mentioned and analyzed only in accounting perspectives, but recently engineering perspectives of LCC analysis based on the execution of appropriate procedures become more important than the accounting perspectives. Especially, the practical use of reliability engineering related methodologies is recognized as a key factor for the LCC analysis. For the practical use of reliability methods, LCC analysis for unexposed problems is a key issue, and utilizing FMEA and FTA techniques is needed to solve the unexposed problems. Reliability, maintainability, availability, and safety should be evaluated by the LCC analysis with the reliability methods, so we study methodologies for the LCC analysis. Present Worth can be calculated by multiplication of Annual Equivalent Cost and PWAF. Reliability engineering related methods are needed for the process of dividing Present Worth into PWAF, and the practical use of reliability methods can improve accuracy of LCC analysis.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.870-876
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• 1995

In this paper, a new three-dimensional nodal diffusion code which is based on the AFEN methodology is described and tested. The method expands the homogeneous flux within a node in ter-ms of eighteen analytic basis functions satisfying the diffusion equation at any point of the node. And the nodal coupling equations are derived such that nodal balance, current continuity and leakage balance within an infinitesimally small box around the edge are satisfied. To verify its accuracy, the code was applied to the well-known static LMW benchmark problem and a small core benchmark problem that has the same material properties as the three-dimensional IAEA benchmark problem and compared with two other codes (QUANDRY, VENTURE). The results show that the code provides good accuracy both in the power distribution and in the effective multiplication factor.

• Nuclear Engineering and Technology
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• v.48 no.3
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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.

• Journal of Energy Engineering
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• v.12 no.2
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• pp.101-108
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• 2003

The purpose of this study is to develop an entrainment model of the Pressurized Water Reactor (PWR) U-tube Steam Generator (SG) for Main Steam Line Break (MSLB) analyses. Generally, the temperature of the inside containment vessel at MSLB is decreased by introducing the liquid entrainment effect. This effect makes a profit on the aspect of integrity evaluation for Equipment Environmental Qualification (EEQ) in the containment. However, the target plant, Kori unit 1 does not have the entrainment data. Therefore, this study has been performed. RETRAN-3D and LOFTRAN computer programs are used for the model development. There are several parameters that are used for the initial benchmark, such as Combustion Engineerings (CE) experimental data and the RETRAN-3D model which describes the test leg. A sensitivity study is then performed with this model in which the model parameters are varied until the calculated results provide reasonable agreement with the measured results for the entire test set. Finally, a multiplication factor has been obtained from the 95/95 values of the calculated (best-estimate) quality data relative to the measured quality data. With this new methodology, an additional temperature margin of about 40$^{\circ}C$ can be obtained. So, the new methodology is found to have an explicit advantage to EQ analyses.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2007.06a
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• pp.179-183
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• 2007

This paper is to implement a low power frequency Shift Keying(FSK) receiver using Xilinx System Generator. The receiver incorporates a 16 point Fast Fourier Transform(FFT) for symbol detection. The design units of the receiver are digital designs for better efficiency and reliability. The receiver functions on one bit data processing and supports data rates 10kbps. In addition CORDIC algorithm is used for avoiding complex multiplications while computing FFT, multiplication of twiddle factor is substituted by rotators. The design and simulation of the receiver is carried out in Simulink, then the simulink model is translated to a hardware model to implement FPGA using Xilinx System Generator and to verify performance.

• Food Science and Biotechnology
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• v.18 no.1
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• pp.12-17
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• 2009

The effect of kimchi, traditional Korean fermented vegetables, on inactivating food-borne pathogens and the kimchi factors affecting the antimicrobial activity were investigated. More cells of Listeria monocytogenes, Staphylococcus aureus, Escherichia coli O157:H7, and Salmonella typhimurium were inactivated in the kimchi that had low pH and high titratable acidity. Of the raw ingredients in kimchi, raw garlic showed the strongest antimicrobial activity against the pathogens. When kimchi was fermented at 0, 4, 10, or $20^{\circ}C$ to pH 4.4, higher kimchi fermentation temperature resulted in higher titratable acidity. The greatest inactivation of S. typhimurium occurred in kimchi fermented at $20^{\circ}C$, while L. monocytogenes were inactivated in kimchi fermented at $0^{\circ}C$ in situ. This study showed that appropriately fermented kimchi can inactivate various food-borne pathogens and that the fermentation temperature of the kimchi is an important factor in determining the ability of the kimchi to inactivate specific pathogens. Lactic acid bacteria (LAB) multiplication and organic acids produced according to LAB metabolism play a role in inactivating food-borne pathogens in kimchi.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.35-42
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• 2018

This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics) Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4) and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE) codes. Cores 133 and 134 were analyzed in 2-D (r, ${\theta}$) and 3-D (r, ${\theta}$, z), using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0), Joint Evaluated Fission and Fusion File (JEFF-3.1), Japanese Evaluated Nuclear Data Library (JENDL-3.2), and Joint Evaluated File (JEF-2.2) nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.43-54
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• 2016

In the present paper, development of the three-dimensional (3D) computational code based on Galerkin finite element method (GFEM) for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA)-3D and Water-Water Energetic Reactor (VVER)-1000 reactor cores. In addition, validation of the calculations against the $P_1$ approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.