• Journal of Korean Association for Spatial Structures
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• v.15 no.4
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• pp.49-56
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• 2015

The roof grid of single-layer space frame structure, for Energy Core of Incheon Airport Second Terminal, is very simple and aesthetic, but it is apt to buckle under external force because of mild curvature and complex shape. The object of this study is to estimate the stability of single-layer space frame structures for Energy Core of Incheon Airport Second Terminal with the analytical conditions of structural design. The results show that the buckling load of model(pin-pin, uniform load, rigid joint), that is, the most similar model to the analytical conditions of structural design. was $10.7kN/m^2$.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.635-641
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• 2016

Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3902-3909
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• 2021

SACOS series of subchannel analysis codes have been developed by XJTU-NuTheL for many years and are being used for the thermal-hydraulic safety analysis of various reactor cores. To achieve fine whole core pin-level analysis, the input preprocessing and parallel capabilities of the code have been developed in this study. Preprocessing is suitable for modeling rectangular and hexagonal assemblies with less error-prone input; parallelization is established based on the domain decomposition method with the hybrid of MPI and OpenMP. For domain decomposition, a more flexible method has been proposed which can determine the appropriate task division of the core domain according to the number of processors of the server. By performing the calculation time evaluation for the several PWR assembly problems, the code parallelization has been successfully verified with different number of processors. Subsequent analysis results for rectangular- and hexagonal-assembly core imply that the code can be used to model and perform pin-level core safety analysis with acceptable computational efficiency.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.345-355
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• 2019

Use of the reconstructed fuel assemblies from partially burnt nuclear fuel pins is analyzed. This reutilization option is a potential candidate technique to make better use of the nuclear resources. Standard two step method is used to calculate node i.e. fuel assembly average burnup and then pin by pin ${\eta}$ values are reconstructed to ascertain the residual reactivity in the used fuel pins. Fuel pins with ${\eta}$ > 1:0 are used to reconstruct to-be-reused fuel assemblies. These reconstructed fuel assemblies are burnt during the cycle 3, 4, 5 and 6 of a 1000 MW PWR core by replacing fresh, once burnt and twice burnt fuel assemblies of the reference core configurations. It is concluded that using reconstructed fuel assemblies for the fresh fuel affect dearly on the cycle length (>50 EFPD) when more than 16 fresh fuel assemblies are replaced. However, this loss is less than 20 days if the number of fresh fuel assemblies is less than eight. For the case of replacing twice burned fuel, cycle length could be increased slightly (10 days or so) provided burnt fuel pins from other reactors were also available. Reactor safety parameters, like axial off set (< ${\pm}10%$), Doppler temperature coefficient (<0), moderator temperature coefficient at HFP (<0) are always satisfied. Though, 2D and 3D pin peaking factors are satisfied (<1:55) and (<2:52) respectively, for the cases using eight or less reconstructed fuel assemblies only.

• KIEE International Transactions on Electrophysics and Applications
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• v.3C no.5
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• pp.165-170
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• 2003

This paper presents development of solid core suspension disc insulators (cap and cap suspension disc insulator) for replacing cap and pin suspension disc insulators in overhead transmission and distribution lines which expose to lightning discharges. By this means the punctured problem caused by steep front surge voltage created by lightning discharge on the lines can be solved. The solid core suspension insulator was designed and constructed based on the dimensions of conventional suspension disc insulators (cap and pin insulators). The insulators are made of alumina porcelain. The electrical and mechanical characteristics of the solid core suspension insulators were carried out. The puncture test was performed in the air by applying steep front impulse voltage with amplitude about 2.5 per unit of 50% flashover (CFO) of the insulator unit at negative standard lightning impulse $1.2/50\;\mu\textrm{s}$ with steepness up to $9200\;kV/\mu\textrm{s}$. The testing results show that solid core suspension disc insulators are not punctured eventhough the steepness of the steep front impulse voltage was increased up to $9200\;kV/\mu\textrm{s}$.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2003.06a
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• pp.1630-1633
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• 2003

Hair pin bending machine is pipe forming machine consisting of heat exchanger product system. Hair pin produced by these machine is pathway of refrigerant and play a important role improving the performance and productivity of heat exchanger. The core technology of hair pin bending machine is forming the straight pipe into U-type without any defaults. Therefore, this paper study the relation between the pipe bending forming and the shape and position of mandrel using the elastic-plastic finite element analysis and provide a foundation technology for which developing the hair pin bending machine. The results are followed 1. Mandrel located in front of rotating center of bending die minimized the circular shape variation of copper pipe. 2. Diameter change of mandrel hardly effect the pipe shape.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1339-1345
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• 2017

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• The Journal of the Korean dental association
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• v.18 no.4 s.133
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• pp.279-286
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• 1980

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.283.1-283
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• 2002