• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.41 no.4
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• pp.275-283
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• 2013

In this paper, the structural optimum design method of composite rotor blade cross-section was investigated with the genetic algorithm. An auto-mesh generation program was developed for iterative calculations of optimum design, and stresses in the blade cross-section were analyzed by VABS (variational asymptotic beam sectional analysis) program. Minimum mass of rotor blade was defined as an object function, and stress failure index, center mass and blade minimum mass per unit length were chosen as constraints. Finally, design parameters such as the thickness and layup angles of a skin, and the thickness, position and width of a torsion box were determined through the structural optimum design method of composite rotor blade cross-section presented in this paper.

• Wind and Structures
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• v.23 no.1
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• pp.45-57
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• 2016

Open-cross-section composite beam (OCB) tends to suffer vortex-induced vibration (VIV) due to its bluff aerodynamic shape. A cable-stayed bridge equipped with typical OCB is taken as an example in this paper to conduct sectional model wind tunnel test. Vortex-induced vibration is observed and maximum vibration amplitudes are obtained. CFD approach is employed to calculate the flow field around original cross sections in service stage and construction stage, as well as sections added with three different countermeasures: splitters, slabs and wind fairings. Results show that flow separate on the upstream edge and cause vortex shedding on original section. Splitters can only smooth the flow field on the upper surface, while slabs cannot smooth flow field on the upper or lower surface too much. Thus, splitters or slabs cannot serve as valid aerodynamic means. Wind tunnel test results show that VIV can only be mitigated when wind fairings are mounted, by which the flow field above and below the bridge deck are accelerated simultaneously.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2733-2742
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• 2020

This paper presents an assessment of applicability of the multigroup cross sections generated with Monte Carlo tools to the fast reactor analysis based on transport calculations. 33-group cross section sets were generated for simple one- (1-D) and two-dimensional (2-D) sodium-cooled fast reactor problems using the SERPENT code and applied to deterministic steady-state and depletion calculations. Relative to the reference continuous-energy SERPENT results, with the transport corrected P₀ scattering cross section, the k-eff value was overestimated by 506 and 588 pcm for 1-D and 2-D problems, respectively, since anisotropic scattering is important in fast reactors. When the scattering order was increased to P₅, the 1-D and 2-D problem errors were increased to 577 and 643 pcm, respectively. A sensitivity and uncertainty analysis with the PERSENT code indicated that these large k-eff errors cannot be attributed to the statistical uncertainties of cross sections and they are likely due to the approximate anisotropic scattering matrices determined by scalar flux weighting. The anisotropic scattering cross sections were alternatively generated using the MC²-3 code and merged with the SERPENT cross sections. The mixed cross section set consistently reduced the errors in k-eff, assembly powers, and nuclide densities. For example, in the 2-D calculation with P3 scattering order, the k-eff error was reduced from 634 pcm to -223 pcm. The maximum error in assembly power was reduced from 2.8% to 0.8% and the RMS error was reduced from 1.4% to 0.4%. The maximum error in the nuclide densities at the end of 12-month depletion that occurred in ²³⁷Np was reduced from 3.4% to 1.5%. The errors of the other nuclides are also reduced consistently, for example, from 1.1% to 0.1% for ²³⁵U, from 2.2% to 0.7% for ²³⁸Pu, and from 1.6% to 0.2% for ²⁴¹Pu. These results indicate that the scalar flux weighted anisotropic scattering cross sections of SERPENT may not be adequate for application to fast reactors where anisotropic scattering is important.

• Wind and Structures
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• v.21 no.1
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• pp.63-84
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• 2015

The shape of a tall building has significant impact on wind force generation and wind-induced dynamic response. To study the effect of recessed cavities, wind excitations on a wind-tunnel model of an H-section tall building were compared with those on a square-section building model. Characteristics of the fluctuating wind pressures on the side faces of the two tall buildings and their role in the generation of crosswind forces on the buildings were investigated with the space-time statistical tool of proper orthogonal decomposition (POD). This paper also compares the use of different pressure data sets for POD analysis in situations where pressures on two different surfaces are responsible for the generation of a wind force. The first POD mode is found to dominate the generation of crosswind excitation on the buildings.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.340-355
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• 2018

The methods and performance of a fast reactor multigroup cross section (XS) generation code EXUS-F are described that is capable of directly processing Evaluated Nuclear Data File format nuclear data files. RECONR of NJOY is used to generate pointwise XS data, and Doppler broadening is incorporated by the Gauss-Hermite quadrature method. The self-shielding effect is incorporated in the ultrafine group XSs in the resolved and unresolved resonance ranges. Functions to generate scattering transfer matrices and fission spectrum matrices are realized. The extended transport approximation is used in zero-dimensional calculations, whereas the collision probability method and the method of characteristics are used for one-dimensional cylindrical geometry and two-dimensional hexagonal geometry problems, respectively. Verification calculations are performed first for various homogeneous mixtures and cylindrical problems. It is confirmed that the spectrum calculations and the corresponding multigroup XS generations are performed adequately in that the reactivity errors are less than 50 pcm with the McCARD Monte Carlo solutions. The nTRACER core calculations are performed with the EXUS-F-generated 47 group XSs for the two-dimensional Advanced Burner Reactor 1000 benchmark problem. The reactivity error of 160 pcm and the root mean square error of the pin powers of 0.7% indicate that EXUF-F generates properly the broad-group XSs.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1789-1803
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• 2022

RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate the feasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal code verification results for the MET-1000 and CAR-3600 benchmark cores. Three different multi-group cross-section calculation schemes are employed to improve the agreement between the nodal and reference solutions. The reference solution is obtained by the MCS code using continuous-energy nuclear data. Additionally, the MCS/RAST-F nodal solution is verified with results based on cross-section generated by collision probability code TULIP. A good agreement between MCS/RAST-F and reference solution is observed with less than 120 pcm discrepancy in k_eff and less than 1.2% root-mean-square error in power distribution. This study confirms the two-step approach MCS/RAST-F as a reliable tool for the three-dimensional simulation of reactor cores with fast spectrum.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3450-3463
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• 2023

In the two-step analysis of Pebble-Bed type High-Temperature Gas-Cooled Reactor (PB-HTGR), the lattice physics calculation for the generation of homogenized cross-sections is based on the fuel pebble. However, the randomly-dispersed fuel particles in the fuel pebble introduce double heterogeneity and randomness. Compared to the deterministic method, the Monte Carlo method which is flexible in geometry modeling provides a high-fidelity treatment. Therefore, the Monte Carlo code NECP-MCX is extended in this study to perform the lattice physics calculation of the PB-HTGR. Firstly, the capability for the simulation of randomly-dispersed media, using the explicit modeling approach, is developed in NECP-MCX. Secondly, the capability for the generation of the homogenized cross-section is also developed in NECP-MCX. Finally, simplified PB-HTGR problems are calculated by a two-step neutronics analysis tool based on Monte Carlo homogenization. For the pebble beds mixed by fuel pebble and graphite pebble, the bias is less than 100 pcm when compared to the high-fidelity model, and the bias is increased to 269 pcm for pebble bed mixed by depleted fuel pebble. Numerical results show that the Monte Carlo lattice physics calculation for the two-step analysis of PB-HTGR is feasible.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.245-258
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• 1989

A 69-group cross section library consisting of more than 130 materials was generated for thermal reactor applications using the NJOY nuclear data processing system and the recent version of evaluated nuclear data files available from IAEA Nuclear Data Section. The multigroup library was validated through the analysis of various criticality experiments and depletion results of PWR. When used with the WIMS-KAERI code, the average $K_{eff}$ obtained for 47 uranium-oxide and 41 uranium metal fueled critical configurations is 0.9997 with a standard deviation of 0.69 percent. The calculated burnup dependent isotopic inventories of uranium and plutonium generally show good agreement with measured values obtained from depleted PWR pins.s.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.177-198
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• 2012

This paper reviews the fast reactor physics and computational methods. The basic reactor physics specific to fast spectrum reactors are briefly reviewed, focused on fissile material breeding and actinide burning. Design implications and reactivity feedback characteristics are compared between breeder and burner reactors. Some discussions are given to the distinct nuclear characteristics of fast reactors that make the assumptions employed in traditional LWR analysis methods not applicable. Reactor physics analysis codes used for the modeling of fast reactor designs in the U.S. are reviewed. This review covers cross-section generation capabilities, whole-core deterministic (diffusion and transport) and Monte Carlo calculation tools, depletion and fuel cycle analysis codes, perturbation theory codes for reactivity coefficient calculation and cross section sensitivity analysis, and uncertainty analysis codes.

• Proceedings of the Korean Society of Machine Tool Engineers Conference
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• 1999.10a
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• pp.231-235
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• 1999

Cooling characteristics of cooling jacket for spindle system with built-in motor are studied. for the analysis, three dimensional model for the cooling jacket is built by using finite volume method. The three dimensional model includes the estimation on the amount of heat generation of bearing and built-in motor and the thermal characteristic values such as heat transfer coefficients on the boundary. The temperature distributions and the cooling characteristics are analyzed by using the commercial software FLUENT. Numerical results show that stream-wise cross section area and flow rate are important factors for cooling characteristics of cooling jacket. Cooling performance of cooling jacket is good in condition that stream-wise cross section's horizontal length is close to its vertical one and flow rate is high. This results show that heat transfer is dominated by velocity profile and heat transfer area.