• Proceedings of the Korea Water Resources Association Conference
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• 2005.09b
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• pp.1217-1219
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• 2005

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.97-109
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• 2022

Multi-dimensional coupling analysis is a research hot spot in nuclear reactor thermal hydraulic study and both the full-scale system transient response and local key three-dimensional thermal hydraulic phenomenon could be obtained simultaneously, which can achieve the balance between efficiency and accuracy in the numerical simulation of nuclear reactor. A one-dimensional to three-dimensional (1D-3D) coupling platform for the nuclear reactor multi-dimensional analysis is developed by XJTU-NuTheL (Nuclear Thermal-hydraulic Laboratory at Xi'an Jiaotong University) based on the CFD code Fluent and system code RELAP5 through the Dynamic Link Library (DLL) technology and Fluent user-defined functions (UDF). In this paper, the International Standard Problem (ISP) No. 43 is selected as the benchmark and the rapid boron dilution transient in the nuclear reactor is studied with the coupling code. The code validation is conducted first and the numerical simulation results show good agreement with the experimental data. The three-dimensional flow and temperature fields in the downcomer are analyzed in detail during the transient scenarios. The strong reverse flow is observed beneath the inlet cold leg, causing the de-borated water slug to mainly diffuse in the circumferential direction. The deviations between the experimental data and the transients predicted by the coupling code are also discussed.

• 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 Korean Tunnelling and Underground Space Association
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• v.9 no.2
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• pp.133-141
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• 2007

The excavation damaged zone (EDZ) is an area around an excavation where in situ rock mass properties, stress condition, displacement, groundwater flow conditions have been altered due to the processes induced by the excavation. Various studies have been carried out on EDZ, but most studies have focused on the mechanical bahavior of EDZ by in situ experiment. Even though the EDZ could potentially form a high permeable pathway of groundwater flow, only a few studies were performed on the analysis of groundwater flow in EDZ. In this study, the 'hydraulic EDZ' was defined as the rock zone adjacent to the excavation where the hydraulic aperture has been changed due to the excavation by using H-M coupling analysis. Fundamental principles of distinct element method (DEM) were used in the analysis. In the same groundwater level, the behavior of hydraulic aperture near the cavern was analyzed for different stress ratios, initial apertures, fracture angles and fracture spacings by using a two-dimensional DEM program. We evaluate the excavation induced hydraulic aperture change. Using the results of the study, hydraulic EDZ was defined as an elliptical shape model perpendicular to the joint.

• Journal of Aerospace System Engineering
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• v.14 no.1
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• pp.16-20
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• 2020

The fatigue happening during the road riding of the vehicle and for the moment the aircraft lands on the runway is closely related to the life cycle of the landing gear, the airframe, the vehicle's suspension, etc. The multiple loads acting on the wheel are longitudinal, lateral, vertical, and braking forces. To study the dynamic characteristics and fatigue stiffness of the vehicle, the dynamic fatigue simulator generally has been used to represent the real road vibration in the lab. It can save time and cost. In hardware, the critical factor in the hydraulic fatigue simulator structure is to decouple each axis and to endure several load vibration. In this paper, the inverse kinematic analysis method derives the magnitude of movement of the hydraulic servo actuator by the coupling after rendering the maximum movement displacement in the axial direction at the center of the dummy wheel. The result of the analysis is that the coupling between the axes is weak to reproduce the real road vibrations precisely.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• Geomechanics and Engineering
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• v.15 no.6
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• pp.1207-1217
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• 2018

Geomechanics parameters are critical to numerical simulation, stability analysis, design and construction of geotechnical engineering. Due to the limitations of laboratory and in situ experiments, back analysis is widely used in geomechancis and geotechnical engineering. In this study, a hybrid back analysis method, that coupling numerical simulation, response surface (RS) and relevance vector machine (RVM), was proposed and applied to identify geomechanics parameters from hydraulic fracturing. RVM was adapted to approximate complex functional relationships between geomechanics parameters and borehole pressure through coupling with response surface method and numerical method. Artificial bee colony (ABC) algorithm was used to search the geomechanics parameters as optimal method in back analysis. The proposed method was verified by a numerical example. Based on the geomechanics parameters identified by hybrid back analysis, the computed borehole pressure agreed closely with the monitored borehole pressure. It showed that RVM presented well the relationship between geomechanics parameters and borehole pressure, and the proposed method can characterized the geomechanics parameters reasonably. Further, the parameters of hybrid back analysis were analyzed and discussed. It showed that the hybrid back analysis is feasible, effective, robust and has a good global searching performance. The proposed method provides a significant way to identify geomechanics parameters from hydraulic fracturing.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1540-1555
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.

• Journal of Advanced Marine Engineering and Technology
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• v.22 no.2
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• pp.241-247
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• 1998

The hydraulic coupling is a kind of power transmission device combined with pump turbine and casing as its main parts. The purpose of this research is to construct an experimental test set-up and to establish an available soft ware for th characteristics of two domestically developed hydraulic couplings. The test item is torgue rpm. and slip and or efficiency characteristic in accordance with variation of oil quantity. in this case the oil quantity was varied 55%, 67% and 77% of the inside capacity of hydraulic copuling.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1769-1785
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• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.