• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.677-690
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• 2009

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.

• Journal of the Korean Society for Precision Engineering
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• v.29 no.2
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• pp.139-146
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• 2012

A multi-tasking machine tool for large scale marine engine crankshafts has been developed together with design technologies for its special devices. Since work pieces, that is, crankshafts to be machined are big and heavy; weight of over 100 tons, length of 10 m long, and diameter of over 3.5 m, several special purpose core devices are necessarily developed such as PTD (Pin Turning Device) for machining eccentric pin parts, face place and steady rest for chucking and resting heavy work pieces. PTD is a unique special purpose device of open-and-close ring typed structure equipped with revolving ring spindle for machining eccentric pins apart from journal. In order to achieve high rigidity of the machine tool, structural design optimization using TMSA (Taguch Method based Sequential Algorithm) has been completed with FEM structural analysis, and a hydrostatic bearing system for the PTD has been developed with theoretical hydrostatic analysis.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1079-1087
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• 2021

NTP-ERSN is a package developed for solving the multigroup form of the discrete ordinates, characteristics and collision probability of the Boltzmann transport equation in one-dimensional cartesian geometry, by combining pin cells. In this work, C5G7 MOX benchmark is used to verify the accuracy and efficiency of NTP-ERSN package, by treating reactor core problems without spatial homogenization. This benchmark requires solutions in the form of normalized pin powers as well as the vectors and the eigenvalue. All NTP-ERSN simulations are carried out with appropriate spatial and angular approximations. A good agreement between NTP-ERSN results with those obtained with OpenMC calculation code for seven energy groups. In addition, our studies about angular and mesh refinements are carried out to produce better quality solution. Moreover, NTP-ERSN GUI has also been updated and adapted to python 3 programming language.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.237-241
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• 2001

The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of $r-{\theta}$ geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.

• Journal of the Korean Society for Precision Engineering
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• v.29 no.2
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• pp.129-138
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• 2012

In this study, three types of large scale multi-tasking machine tools together with core technologies involved have been developed and introduced; a multi-tasking machine tool for large scale marine engine crankshafts, a multi-tasking vertical lathe for windmill parts, and a large scale 5-axis machine tool of gantry type. Several special purpose devices has been necessarily developed for the purpose of handling and machining big and heavy workpieces accurately, such as PTD (Pin Turning Device) with revolving ring spindle for machining eccentric crankshaft pins, hydrostatic rotary table and steady rest for supporting and resting heavy workpieces, and 2-axis automatic swiveling head for high-quality free surface machining. Core technologies have been also developed and adopted on their detail design stage; 1) structural design optimization with FEM structural analysis, 2) theoretical hydrostatic analysis for the PTD and rotary table bearings, 3) box-in-box type cross-rail and octagonal ram design to secure machine rigidity and accuracy, 4) constant spindle rpm control against gravitational torque due to unbalanced workpiece.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1291-1302
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• 2016

This paper presents a new and efficient scheme to determine the optimal neutron source position in a model near-equilibrium pressurized water reactor, which is based on the OPR1000 Hanul Unit 3 Cycle 7 configuration. The proposed scheme particularly assigns importance of source positions according to the local adjoint flux distribution. In this research, detailed pin-by-pin reactor adjoint fluxes are determined by using the Monte Carlo KENO-VI code from solutions of the reactor homogeneous critical adjoint transport equations. The adjoint fluxes at each allowable source position are subsequently ranked to yield four candidate positions with the four highest adjoint fluxes. The study next simulates ex-core detector responses using the Monte Carlo MAVRIC code by assuming a neutron source is installed in one of the four candidate positions. The calculation is repeated for all positions. These detector responses are later converted into an inverse count rate ratio curve for each candidate source position. The study confirms that the optimal source position is the one with very high adjoint fluxes and detector responses, which is interestingly the original source position in the OPR1000 core, as it yields an inverse count rate ratio curve closest to the traditional 1/M line. The current work also clearly demonstrates that the proposed adjoint flux-based approach can be used to efficiently determine the optimal geometry for a neutron source and a detector in a modern pressurized water reactor core.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.407-412
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• 2014

In this research, we investigated the burnup characteristics and the conversion of fertile $^{232}Th$ into fissile $^{233}U$ in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning $^{232}Th$ fuel (fuel pin 1) and $^{233}U$ fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

• Proceedings of the IEEK Conference
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• 2003.07b
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• pp.859-862
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• 2003

This paper proposes an emulation environment for SoC designs using small number of large gate-count FPGA's and a PC system. To overcome the pin limitation problem in partitioning the design when the design size overwhelms the FPGA gate count, we use bus splitter modules that replicate on-chip bus signals in one FPGA to arbitrary number of other FPGA's with minimal pin count. The proposed scheme is applied to the emulation of 2 million gate multimedia processing chip using two Xilinx Viretex-2 6000 FPGA devices in 6.6MHz operating frequency. An ARM core, memories, camera and LCD display are modeled in software using dual 2GHz Pentium-III processors. This scheme can be utilized for more than 2 FPGA's in the same ways as two FPGA case without losing emulation speed.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3207-3216
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• 2021

Three-dimensional structures of a vortical flow field and heat transfer characteristics in a partially blocked 7-pin fuel assembly mock-up of sodium-cooled fast reactor have been investigated through a numerical analysis using a commercial computational fluid dynamics code, ANSYS CFX. The simulation with the SST turbulence model agrees well with the experimental data of outlet and cladding wall temperatures. From the analysis on the limiting streamline at the wall, multi-scale vortexes developed in axial direction were found around the blockage. The vortex core has a high cladding wall temperature, and the attachment line has a low cladding wall temperature. The small-scale vortex structures significantly enhance the convective heat transfer because it increases the turbulent mixing and the turbulence kinetic energy. The large-scale vortex structures supply thermal energy near the heated cladding wall surface. It is expected that control of the vortex structures in the fuel assembly plays a significant role in the convective heat transfer enhancement. Furthermore, the blockage plate and grid spacer increase the pressure drop to about 36% compared to the bare case.

• Journal of the Korean Society for Precision Engineering
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• v.28 no.5
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• pp.623-631
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• 2011

Metallic sandwich panels based on a truss core structure have been developed for a wide range of potential applications with their lightweight and multi-functionality. Structural performance of sandwich panels can be predicted from the studies on mechanical behavior of a unit cell of truss core structures. Analytical investigations on the unit cell provide approximated guidelines for the design of overall core structures for a specific application in short time. In this study, the effects of geometrical parameters on mechanical behavior of a pyramidal shape of unit cell were investigated with analytical models. The unit cell with truss member angle of 45 degree was considered as reference model and other models were designed to have the same weight and projected area but different truss member angle. All truss members were assumed to be connected with pin joint in analytical models. Under the assumptions, the equivalent strength and stiffness of the unit cell under compressive and shear loads were predicted and compared. And finally, the optimum core member angle to have maximum mechanical property could be calculated and verified with FE analysis results.