• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.225-230
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• 1996

A three-dimensional reactor core simulation code, MASTER has been developed as a part of ADONIS which is the Korean core design package in KAERI. CASMO-3 is used as a precedent lattice code for two-group microscopic cross section and heterogeneous formfunctions. The pin power reconstruction capability of CASMO-3/MASTER was evaluated for a validation and verification Five B&W critical experiments were selected as benchmark problems. These problems included two experiments for CE-type and three for WH-type fuel assemblies. Two of them contained gadolinia rods as burnable absorber. Comparison of the calculated pin power distributions with the measured ones demonstrate that CASMO-3/MASTER can predict the pin power distribution as well as CASMO-3/SIMULATE-3.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2009년도 춘계 학술논문 발표대회 학계
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• pp.65-68
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• 2009

Recently, as architectural concrete structures become high-rise and megastructured, concrete become high-strengthened and, by ensuring products of more stability, and rationalization of construction are required.large cross-sectional precast concrete members such as columns show large temperature increase in manufacturing process not only by external heating but also by concrete itself's hydration heating. Therefore, it is expected that specimen for management to predict strength and compression strength of precast concrete member shows different strength characteristics. Concerning this, in order to suggest strength characteristics of high strength mass concrete suitable for precast concrete application, this study comprises the inclusive investigations on the relations between core strength and the strength characteristics per member cross-section dimensional value and per water-bonding material ratio value.

• Nuclear Engineering and Technology
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• 제51권5호
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• pp.1209-1217
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• 2019

Due to the limitation of the computers, the conventional homogenization method is based on many assumptions and approximations, and some tough problems such as energy spectrum and boundary condition are faced. To deal with those problems, the Monte Carlo global homogenization is adopted. The Reactor Monte Carlo code RMC is used to study the global homogenization method, and the one-step global homogenization method is proposed. The superimposed mesh geometry is also used to divide the physical models, leading to better geometric flexibility. A set of multigroup homogenization cross sections is online generated for each mesh under the real neutron energy spectrum and boundary condition, the cross sections are adjusted by the superhomogenization method, and no leakage correction is required. During the process of superhomogenization, the author-developed reactor core program NLSP3 is used for global calculation, so the global flux distribution and equivalent homogenization cross sections could be solved simultaneously. Meanwhile, the calculated homogenization cross section could accurately reconstruct the non-homogenization flux distribution and could also be used for fine calculation. This one-step global homogenization method was tested by a PWR assembly and a small reactor model, and the results show the validity.

• Steel and Composite Structures
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• 제39권5호
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• pp.583-598
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• 2021

The composite steel reinforced concrete (SRC) columns have been widely used in Structural Engineering due to their good performances. Many studies have been done on the SRC columns' performances, but they focused on the ordinary types with conventional configurations and materials. In this study, nine new types of steel reinforced lightweight aggregate concrete (SRLAC) short columns with cross-shaped (+shaped and X-shaped) steel section were tested under monotonically axial compressive load; the studied parameters included steel section ratio, steel section configuration, ties spacing, lightweight aggregate concrete (LWAC) strength, and longitudinal bars ratio. From the results, it could be found that the specimens with larger ties ratio, concrete strength, longitudinal bars ratio, and steel section ratio achieved great strength and stiffness due to the excellent interaction between the concrete and steel. The well-confined concrete core could strengthen the steel section. The ductility and toughness of the specimens were influenced by the LWAC strength, steel section ratio, and longitudinal bars ratio; in addition, larger ties ratio with smaller LWAC strength led to better ductility and toughness. The load transfer between concrete and steel section largely depends on the LWAC strength, and the ultimate strength of the new types of SRLAC short columns could be approximately predicted, referring to the codes' formulas of ordinary types of steel reinforced concrete (SRC) columns. Among the used codes, the BS-5400-05 led to the most conservative results.

• International Journal of Naval Architecture and Ocean Engineering
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• 제6권2호
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• pp.245-256
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• 2014

I-core sandwich panel that has been used more widely is assembled using high power $CO_2$ laser welding. Kim et al. (2013) proposed a circular cone type heat source model for the T-joint laser welding between face plate and core. It can cover the negative defocus which is commonly adopted in T-joint laser welding to provide deeper penetration. In part I, a volumetric heat source model is proposed and it is verified thorough a comparison of melting zone on the cross section with experiment results. The proposed model can be used for heat transfer analysis and thermal elasto-plastic analysis to predict welding deformation that occurs during laser welding. In terms of computational time, since the thermal elasto-plastic analysis using 3D solid elements is quite time consuming, shell element model with multi-layers have been employed instead. However, the conventional layered approach is not appropriate for the application of heat load at T-Joint. This paper, Part II, suggests a new method to arrange different number of layers for face plate and core in order to impose heat load only to the face plate.

• 한국농공학회지
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• 제31권2호
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• pp.125-134
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• 1989

This study was conducted to investigate the flexural behaviour of sandwich constructions with cement concrete core and polymer concrete facings. Six different cross-sectional shapes using epoxy based polymer concrete facings were investigated. Some of the results from the static tests are given including the load-deflection responses, load-strain relationships, ultimate moment, and mode of failure. From the. results the following conclusions can be made. 1. The various strengths of polymer concrete were very high compared to the strengths for portland cement concrete, while modulus of elasticity assumed an aspect of contrast. 2. The thickness of core and facing exerted a great influence on the deflection and ultimate strenght of polymer concrete sandwich constructions. 3. The variation shape of deflection and strain depend on loading were a very close approximation to the straight line. The ultimate strain of polymer concrete at the end of tensile side were ranged from 625x10-6 to 766x10-6 and these values increased in proportion to the decrease of thickness of core and facings. 4. The ultimate moments of polymer sandwich constructions were 3 to 4 times that of cement concrete constructions which was transformed same section. It should he noted that polymer concrete have an effect on the reinforcement of weak constructions. 5. Further tests are neede to investigate the shear strain of constructions, and thermal expansion, shrinkage and creep of cement and polymer concrete which were composite materials of sandwich constructions.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1721-1728
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• 2019

The advanced node core code CORCA-3D is one of the independent developed codes of NPIC for the nuclear reactor core design. CORCA-3D code can calculate the few-group cross section, solve the 3D diffusion equations, consider the thermal-hydraulic feedback, reconstruct the pin-by-pin power. It has lots of functions such as changing core status calculation, critical searching, control rod value calculation, coefficient calculation and so on. The main theory and functions of CORCA-3D code are introduced and validated with a lot of reactor measured data and the SCIENCE system. Now, CORCA-3D code has been applied in ACP type reactor nuclear cores design.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2635-2649
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• 2022

CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect. The neutronic simulation of the start-up experiments conducted at the CEFR have been performed with a deterministic code system RAST-F, which is based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo (MC) code and a multi-group nodal diffusion solver. The RAST-F results were compared against the measurement data. Moreover, the characteristic of neutron spectrum in the fuel rings, and adjacent reflectors was evaluated using different models for generation of accurate nuclear libraries. The numerical solution of RAST-F system was verified against the full core MC solution MCS at all control rods fully inserted and withdrawn states. A good agreement between RAST-F and MCS solutions was observed with less than 120 pcm discrepancies and 1.2% root-mean-square error in terms of k_eff and power distribution, respectively. Meanwhile, the RAST-F result agreed well with the experimental values within two-sigma of experimental uncertainty. The good agreement of these results indicating that RAST-F can be used to neutronic steady-state simulations for small core-size SFR, which was challenged to deterministic code system.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.209-214
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• 2023

The effect of thermal annealing on defect recovery of in-core neutron-irradiated 4H-SiC was investigated. Au/SiC Schottky diodes were manufactured using a 4H-SiC epitaxial wafer that was neutron-irradiated at the HANARO research reactor. The electrical characteristics of their epitaxial layers were analyzed under various conditions, including different neutron fluences (1.3 × 10¹⁷ and 2.7 × 10¹⁷ neutrons/cm²) and annealing times (up to 2 h at 1700 ℃). Capacity-voltage measurements showed high carrier compensation in the neutron-irradiated samples and a recovery tendency that increased with annealing time. The carrier density could be recovered up to 77% of the bare sample. Deep-level-transient spectroscopy revealed intrinsic defects of 4H-SiC with energy levels 0.47 and 0.68 eV below the conduction-band edge, which were significantly increased by in-core neutron irradiation. A previously unknown defect with a high electron-capture cross-section was discovered at 0.36 eV below the conduction-band edge. All defect concentrations decreased with 1700 ℃ annealing; the decrease was faster when the defect level was shallow.

• Structural Engineering and Mechanics
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• 제53권1호
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• pp.17-26
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• 2015

The cross sections of hollow cylindrical tubes ovalise under a pure bending condition, and this reduces their flexural stiffness as their curvatures increase. It is important to accurately evaluate this phenomenon, known as the 'Brazier effect', to understand the bending behaviour of the systems considered. However, if the tubes are supported by an elastic medium or foundation, the ovalisation displacements of their cross sections may decrease. From this point of view, the purpose of this research is to analytically investigate the bending characteristics of single- and double-walled elastic tubes contacted by an elastic material by considering the Brazier effect. The Brazier moment, which is the maximum moment-carrying capacity of the ovalised cross section, can be calculated by introducing the strain energy per unit length of the tube in terms of the degree of ovalisation for the tube and the curvature. The total strain energy of the double-walled system is the sum of the strain energies of the outer and inner tubes and that of the compliant core. Results are comparatively presented to show the variation in the degree of ovalisation and the Brazier moment for single- and double-walled tubes.