• 한국전산유체공학회지
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• 제21권3호
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• pp.89-97
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• 2016

In this study, direct code coupling, in which two codes share a single flow field, was conducted using 3-dimensional high resolution thermal hydraulics code, CUPID and 1-dimensional system analysis code, MARS. This approach provide the merit to use versatile capability of MARS for nuclear power plants and 3-dimensional T/H analysis capability of CUPID. Numerical Method to directly couple CUPID and MARS was described in this paper. The straight flow and manometer flow oscillation were calculated to verify conservation of coupled CUPID/MARS code in mass, momentum, and energy. This verification calculations indicates that the CUPID/MARS is coupled appropriately in numerical aspect and the coupled code can be applied to nuclear reactor thermal hydraulics after validation against integral transient experiments.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1603-1610
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• 2020

Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1204-1212
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• 2024

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4048-4056
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• 2023

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.

• 대한토목학회논문집
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• 제37권6호
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• pp.973-980
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• 2017

본 연구는 해수침투를 모의하는 연산의 바탕이 되는 밀도 결합 방정식을 2차원으로 유한차분한 VDFT (Visual Basic based Density-coupled Flow and Transport) 기법을 개발한 것이다. VDFT 코드는 직관적이고 간단하게 사용할 수 있다는 장점을 가지고 있으며 일반 업무용으로 널리 사용되고 있는 EXCEL Visual Basic 플랫폼을 활용할 수 있다는 장점이 있다. 일반적으로 수치모의를 위해 개발한 코드는 벤치마크 라는 대표적인 수치예제를 통하여 검증을 할 수 있다. 본 연구에서는 2가지 실내실험 결과로서 얻어진 데이터와 Henry Problem 및 Modified Henry Problem을 수치예제로 활용하여 VDFT 기법을 검증하였다. 마지막으로 결론에서는 VDFT 코드의 활용가능성을 진단하고 향후 연구의 방향성을 제시하였다.

• Earthquakes and Structures
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• 제9권2호
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• pp.391-413
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• 2015

Damage of torsionally coupled buildings situated on soil sites has been reported in literature, however no site-specific studies are available for torsionally coupled buildings having site characteristics as a parameter. Effect of torsion is being accounted in seismic codes by the provision of design eccentricity where the dynamic to static eccentricity ratio is a parameter. In this paper, a methodology to determine dynamic to static eccentricity ratio of torsionally coupled buildings has been demonstrated for Delhi region for two torsionally coupled buildings on three soil sites. The variations of average and standard deviations of frame shears for stiff and flexible edges are studied for four eccentricity ratios for the two buildings for the three sites. From the limited studies made, it is observed that the dynamic to static eccentricity ratios observed for site-specific earthquakes are different from Indian seismic code specified value, hence a proposal is made to include a comment in Indian seismic code. Methodology proposed in this paper can be adopted for any region, for the estimation of dynamic to static eccentricity ratio for site specific earthquake.

• 신재생에너지
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• 제7권4호
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• pp.50-57
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• 2011

A wind turbine simulation program for the coupled dynamics of aerodynamics, elasticity, multi-body dynamics and controls of turbine is newly developed by combining an aero-elastic code and a multi-body dynamics code. The aero-elastic code, based on the blade momentum theory and generalized dynamic wake theory, is developed by NREL(National Renewable Energy Laboratory, USA). The multi-body dynamics code is commercial one which is capable of accounting for geometric nonlinearity and twist deflection. A turbulent wind load case is simulated for the NREL 5-MW baseline wind turbine model by the developed program and FAST. As a result, the two results agree well enough to verify the reliability of the developed program.

• Journal of Power Electronics
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• 제14권5호
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• pp.827-833
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• 2014

This study proposes a coupled inductor method for the parallel operation of a power conditioning system (PCS). When primary and secondary currents flow in the same direction in a coupled inductor, total flux and inductance are cancelled; when currents flow in opposite directions, each flux becomes an individual inductor. These characteristics are applied in the parallel operation of a PCS. To connect at a grid code, abnormal current, which is barred under the grid connection code, is blocked by using a coupled inductor. A design based on the capacity and current duration time of a PCS is verified through hardware implementation. Experiment results show the effectiveness of variance reduction.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4280-4295
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• 2022

In this paper, a new multi-group neutron-gamma transport calculation code system STRAUM-MATXST for complicated geometrical problems is introduced and its development status including numerical tests is presented. In this code system, the MATXST (MATXS-based Cross Section Processor for S_N Transport) code generates multi-group neutron and gamma cross sections by processing MATXS format libraries generated using NJOY and the STRAUM (S_N Transport for Radiation Analysis with Unstructured Meshes) code performs multi-group neutron-gamma coupled transport calculation using tetrahedral meshes. In particular, this work presents the recent implementation and its test results of the Krylov subspace methods (i.e., Bi-CGSTAB and GMRES(m)) with preconditioners using DSA (Diffusion Synthetic Acceleration) and TSA (Transport Synthetic Acceleration). In addition, the Krylov subspace methods for accelerating the energy-group coupling iteration through thermal up-scatterings are implemented with new multi-group block DSA and TSA preconditioners in STRAUM.

• 항공우주기술
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• 제5권2호
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• pp.60-66
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• 2006

본 논문은 MSC/Nastran superelement를 이용한 연성하중해석에 대한 연구이다. 위성개발시 발사체가 선정되면, 발사체와 위성체간의 연성하중해석이 실시된다. 연성하중해석 결과로부터 위성구조체의 주요 부위에서의 하중과 변위를 도출하고, 이로부터 현 설계의 안정성을 판단하게 된다. 지금까지의 연성하중해석은 MSC/Nastran의 DMAP 코드를 이용하여 수행이 되었다. DMAP 코드의 경우 코드가 매우 복잡하고, 길기 때문에 코드 분석 및 수정에 어려움이 많았다. 이를 해결하기 위해서 MSC/Nastran 2005의 superelement를 이용하여 연성하중해석을 수행하였다. 던저 위성체 유한요소모델을 MSC/Nastran 2005 superelement를 이용하여 Craig-Bampton 모델로 변환하고 이를 정확성을 검증하었다. Craig-Bampton 모델로 변환된 위성체 모델을 발사체 모델과 접속하여 연성하중해석을 수행하였고, 이의 정확성을 검증하였다.