• Development and Reproduction
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• v.15 no.3
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• pp.239-247
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• 2011

Spermatogenesis and taxonomic values of mature sperm morphology of in male Septifer (Mytilisepta) virgatus were investigated by transmission electron microscope observations. The morphologies of the sperm nucleus and the acrosome of this species are the cylinder shape and cone shape, respectively. Spermatozoa are approximately 45-50 ${\mu}m$ in length including a sperm nucleus (about 1.26 ${\mu}m$ long), an acrosome (about 0.99 ${\mu}m$ long), and tail flagellum (about 45-47 ${\mu}m$). Several electron-dense proacrosomal vesicles become later the definitive acrosomal vesicle by the fusion of several Golgi-derived vesicles. The acrosome of this species has two regions of differing electron density: there is a thin, outer electron-dense opaque region (part) at the anterior end, behind which is a thicker, more electron-lucent region (part). In genus Septifer in Mytilidae, an axial rod does not find and also a mid-central line hole does not appear in the sperm nucleus. However, in genus Mytilus in Mytilidae, in subclass Pteriomorphia, an axial rod and a mid-central line hole appeared in the sperm nucleus. These morphological differences of the acrosome and sperm nucleus between the genuses Septifer and Mytilus can be used for phylogenetic and taxonomic analyses as a taxonomic key or a significant tool. The number of mitochondria in the midpiece of the sperm of this species are five, as seen in subclass Pteriomorphia.

• Nuclear Engineering and Technology
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• v.31 no.4
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• pp.391-400
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• 1999

A conceptual core design of the 1,300MWe KNGR (Korean Next Generation Reactor) without using soluble boron for reactivity control was developed to determine whether it is technically feasible to implement SBF (Soluble Boron Free) operation. Based on the borated KNGR core design, the fuel assembly and control rod configuration were modified for extensive use of burnable poison rods and control rods. A new fuel rod, in which Pu-238 had been substituted for a small amount of U-238 in fuel composition, was introduced to assist the reactivity control by burnable poison rods. Since Pu-238 has a considerably large thermal neutron capture cross section, the new fuel assembly showed good reactivity suppression capability throughout the entire cycle turnup, especially at BOC (Beginning of Cycle). Moreover, relatively uniform control of power distribution was possible since the new fuel assemblies were loaded throughout the core. In this study, core excess reactivity was limited to 2.0 %$\delta$$\rho$ for the minimal use of control rods. The analysis results of the SBF KNGR core showed that axial power distribution control can be achieved by using the simplest zoning scheme of the fuel assembly Furthermore, the sufficient shutdown margin and the stability against axial xenon oscillations were secured in this SBF core. It is, therefore, concluded that a SBF operation is technically feasible for a large sized LWR (Light Water Reactor).

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1125-1134
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• 2024

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19-and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ~10 % and maximum errors of ~20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

• Nuclear Engineering and Technology
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• v.17 no.2
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• pp.105-115
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• 1985

The axial height-dependent transverse buckling is derived from 3-dimensional depletion file in steadystate conditions. For transient conditions a physical correlation is developed based on the linear relationship existing between the responses of in-core and ex-core detectors. The use of this model greatly improves the reliability of a 1-dimensional diffusion theory program in Predicting the axial power transients accompanying large variations of control rod positions.

• Journal of the Korean Geotechnical Society
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• v.19 no.6
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• pp.15-22
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• 2003

In the conventional load transfer analysis for a steel pipe drilled shaft, it was assumed that the concrete's strain is the same as the measured steel's strain and the elastic modulus of the steel and the concrete calculated by the formular as prescribed by specification is used in the calculation of pile axial load. But, the pile axial load calculation by conventional method differed to some extent from the actual pile load. So, the behavior of a steel pipe drilled shaft could not be analyzed exactly. Thus, the necessity to measure the strain for each pile component was proposed. In this study, a new approach for load transfer measurement of large diameter drilled shafts was suggested ; the strain of each pile component(i. e., steel and concrete) was measured by DRS(Deformable Rod Sensor), the elastic modulus was determined by the uniaxial compression test for concrete specimens made at test site and a value of elastic modulus was evaluated as average tangential modulus corresponding to the stress level of the (0.2-0.6)$f_{ck}$. Field application was confirmed by the results of load transfer measurement tests for 3 drilled shafts. The errors for calculated pile head load were -11 ∼16% and 3.4% separately.

• Journal of Korean Association for Spatial Structures
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• v.2 no.3 s.5
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• pp.81-92
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• 2002

The objective of this paper is to help to make decision of the appropriate structural types in long span structured building due to range of span. For the intention, based on 7 forces of structural element, it is analized the relationships among 6 configurations of structural element(d/1), 25 structural types, 4 materials, and span-length known with 186 sample from 1850 to 1996. 1) bending forces: $club(1/100{\sim}1/10),\;plate(1/100{\sim}1/10),\;rahmen(steel,\;10{\sim}24m)\;simple\;beam(PC,\;10{\sim}35m)$ 2) shearing forces: $shell(1/100{\sim}1/1000)\;hyperbolic\;paraboloids(RC,25{\sim}97m)$ 3) shearing+bending forces: plate, folded $plate(RC21{\sim}59m)$ 4) compression axial forces: club, $arch(RC,\;32{\sim}65m)$ 5) compression+tension forces: shell, braced dome $shell(RC,\;40{\sim}201m),\;vault\;shell(RC,\;16{\sim}103m)$ 6) compression+tension axial forces: $rod(1/1000{\sim}1/100)$, cable(below 1/1000)+rod, coble+rod+membrane(below 1/1000), planar $truss(steel,\;31{\sim}134m),\;arch\;truss(31{\sim}135m),\;horizontal\;spaceframe(29{\sim}10\;8m),\;portal\;frame(39{\sim}55m),\;domical\;space\;truss(44{\sim}222m),\;framed\;\;membrane(45{\sim}110m),\;hybrid\;\;membrane\;(42{\sim}256m)$ 7) tension forces: cable, membrane, $suspension(60{\sim}150m),\;cable\;\;beam(40{\sim}130m),\;tensile\;membrane(42{\sim}136m),\;cable\;-slayed(25{\sim}90m),\;suspension\;membrane(24{\sim}97m),\;single\;layer\;pneumatic\;structure(45{\sim}231m),\;double\;layer\;pneumatic\;structures(30{\sim}44m)$

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.162-174
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• 1996

The local hydraulic characteristics in a single phase flow of a 6$\times$6 rod bundle with neighboring different spacer grids were measured by using a LDV(Laser Doppler Velocimeter) system. 6$\times$6 rod bundle is formed by two 3$\times$6 rod bundles with different spacer grids. The objective of this study in a rod bundle is to investigate the thermal-hydraulic interactions between different spacer grids with different configurations and resistance. By using a LDV system, the velocity and turbulent intensity in axial and horizontal directions ore measured. Pressure drop measurements ore also performed to evaluate the loss coefficient for the spacer grid and the friction factor for rod bundles. Implications concerning thermal mining due to spacer grids were investigated based on the hydraulic test results. Swirl factor, which is assumed as a qualitative criteria for DNB(departure from nucleate boiling), was defined and estimated from the horizontal velocity result.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.41 no.1
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• pp.53-62
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• 2017

The forced convection heat transfer performance of a twist-vane spacer grid for a dual-cooled annular fuel assembly was examined experimentally. The twist-vane spacer grid was uniquely designed to enhance mixing inside subchannels and mixing between adjacent subchannels. For testing, a $4{\times}4$ square-arrayed rod bundle with narrow gaps between rods was prepared as the dual-cooled annular fuel assembly to be simulated. The pitch-to-rod diameter ratio of simulated dual-cooled annular fuel assembly was 1.08. The experiments were performed under the following conditions: axial bulk velocity, 1.5 m/s and heat flux, $26kW/m^2$. With regard to the circumferential temperature distribution, the lowest rod-wall temperatures upstream and downstream were measured at the subchannel center and the position toward the tip of twist-vane, respectively. With regard to the axial temperature distribution, behind the twist-vane spacer grid, the rod-wall temperature decreased drastically, and the Nusselt number was enhanced by up to 56 %. The present measured data indicate that the twist-vane spacer grid can effectively improve the forced convection heat transfer in the dual-cooled annular fuel assembly with narrow gaps.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.97-106
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• 2018

Improved load-following capability is one of the most important technical tasks of a pressurized water reactor. Controlling the nuclear reactor core during load-following operation leads to some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking: the core is subjected to sharp and large variation of local power density during transients. Axial offset (AO) is the parameter usually used to represent the core power peaking. One of the important local power peaking components in nuclear reactors is axial power peaking, which continuously changes. The main challenge of nuclear reactor control during load-following operation is to maintain the AO within acceptable limits, at a certain reference target value. This article proposes a new robust approach to AO control of pressurized water reactors during load-following operation. This method uses robust feedback-linearization control based on the multipoint kinetics reactor model (neutronic and thermal-hydraulic). In this model, the reactor core is divided into four nodes along the reactor axis. Simulation results show that this method improves the reactor load-following capability in the presence of parameter uncertainty and disturbances and can use optimum control rod groups to maneuver with variable overlapping.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2004.11a
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• pp.199-203
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• 2004

In this paper, a simplified model is studied to predict analytically the radiated noise from the helical gear system due to an axial excitation of helical gear. The simplified model describes gear, shaft, bearing, and housing. To obtain the axial force of helical gear, mesh stiffness is calculated in the load deflection relation. The axial force is obtained from the solution of the equation of motion, using the mesh stiffness. It is used as a longitudinal excitation of the shaft, which in turn drives the gear housing through the bearing. In this study, the shaft is modeled as a rod, while the bearing is modeled as a parallel spring and damper only supporting longitudinal forces. The gear housing is modeled as a clamped circular plate with viscous damping. For the modeling of this system, transfer function from the shaft to the clamped plate are used, using a spectral method with four pole parameters. Out-of-plane displacement for the thin circular plate with viscous damping is derived and sound pressure radiated from the plate is also derived. Using the model, parameter studies are carried out.