• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2162-2173
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• 2021

In this paper, the coupling of fine-mesh computational fluid dynamics (CFD) thermal-hydraulics (TH) code and neutronics code is achieved using the Ansys Fluent User Defined Function (UDF) for code development, including parallel meshing mapping, data computation, and data transfer. Also, some CFD schemes are designed for mesh mapping and data transfer to guarantee physical conservation in the coupling computation. Because there is no rigorous research that gives robust guidance on the various CFD schemes that must be obtained before the fine-mesh coupling computation, this work presents a quantitative analysis of the CFD meshing and mapping schemes to improve the accuracy of the value and location of key physical prediction. Furthermore, the effect of the sub-pin scale coupling computation is also studied. It is observed that even the pin-resolved coupling computation can also create a large deviation in the maximum value and spatial locations, which also proves the significance of the research on mesh mapping and data transfer for CFD code in a coupling computation.

• Proceedings of the KIEE Conference
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• 2002.07c
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• pp.1964-1965
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• 2002

This paper reports on the single mode fiber coupling efficiency. There is difference between simulation by CODE-V and theoretical result, we observed CEO which is taken by experiment in lab and found out that CEO taken by CODE-V was closer to real experiment result then theoretical result.

• 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.

• Coupled systems mechanics
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• v.5 no.2
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• pp.145-156
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• 2016

An efficient and general numerical strategy for fluid-structure interaction problems is presented where either the fluid or the structure part are represented by nonlinear models. This partitioned strategy is implemented under the form of code coupling that allows to (re)-use previous made developments in a more general multi-physics context. This strategy and its numerical implementation is verified on classical fluid-structure interaction benchmarks, and then applied to the impact of tsunamis waves on submerged structures.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4431-4446
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• 2023

A Multiphysics coupling framework, MPCORE, has been developed to analyze safety parameters using the best estimate codes. The framework contains neutron kinetics (NK), thermal hydraulics (TH), and fuel performance (FP) codes to analyze fuel burnup, radial power distribution, and coolant temperature (T_bc). Shuffling and rotation capabilities have been verified on the Watts Bar reactor for three cycles. This study focuses on two coupling approaches for TH and FP modules. The one-way coupling approach involves coupling the FP code with the NK code, providing no data to the TH modules but getting T_bc as boundary condition from TH module. The two-way coupling approach exchanges information from FP to TH modules, so that the simplified heat conduction solver of the TH module is not used. The power profile in both approaches does not differ significantly, but there is an impact on coolant and cladding parameters. The one-way coupling approach tends to over-predict the cladding hydrogen concentration (CHC). This research highlights the difference between one-way and two-way coupling on critical boron concentration, T_bc, CHC, oxide surface temperature, and pellet centerline temperature. Overall, MPCORE framework with two-way coupling provides a more accurate and reliable analysis of safety parameters for nuclear reactors.

• Journal of computational fluids engineering
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• v.21 no.3
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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.

• Journal of computational fluids engineering
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• v.19 no.3
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• pp.62-68
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• 2014

The aerodynamic and noise calculations were performed through the CFD-CSD loose coupling methodology. In the loose coupling process, the trimmed rotor airloads were predicted by the in-house CFD code based on unstructured overset meshes, and the trim of the rotorcraft and the aeroelastic deformation of rotor blades were accounted with the CAMRAD II rotorcraft comprehensive code. The set of codes was used to analyze the HART-II baseline test condition. The effect of grid resolution and time step was examined and the loose coupling approach was found to be stable and convergent for the case. Comparison of the resulting sectional airloads, structural deformations, the noise carpets and the wake geometry with experimentally measured data was presented and showed the good agreement.

• KIPS Transactions on Software and Data Engineering
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• v.6 no.5
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• pp.229-234
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• 2017

Because of the invisible nature of software and the bad coding habits (bad smell) of the existing developers, there are many redundant codes and unnecessary codes, which increases the complexity and makes it difficult to upgrade software. Therefore, it is required a code visualization so that developers can easily and automatically identify the complexity of the source code. To do this, it is necessary to construct SW visualization tool based on open source software and redefine the coupling and cohesion according to the object oriented viewpoint. Specially to identify a bad smell code pattern, we suggest how to plug-in diverse parsers within our tool. In this paper, through redefining coupling and cohesion from an object oriented perspective, we will extract bad smell code patterns within source code from inputting any pattern into the tool.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.661-673
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• 2020

The quarter-core simulation of BEAVRS Cycle 2 depletion benchmark has been conducted using the MCS/CTF coupling system. MCS/CTF is a cycle-wise Picard iteration based inner-coupling code system, which couples sub-channel T/H (thermal/hydraulic) code CTF as a T/H solver in Monte Carlo neutron transport code MCS. This coupling code system has been previously applied in the BEAVRS benchmark Cycle 1 full-core simulation. The Cycle 2 depletion has been performed with T/H feedback based on the spent fuel materials composition pre-generated by the Cycle 1 depletion simulation using refueling capability of MCS code. Meanwhile, the MCS internal one-dimension T/H solver (MCS/TH1D) has been also applied in the simulation as the reference. In this paper, an analysis of the detailed criticality boron concentration and the axially integrated assembly-wise detector signals will be presented and compared with measured data based on the real operating physical conditions. Moreover, the MCS/CTF simulated results for neutronics and T/H parameters will be also compared to MCS/TH1D to figure out their difference, which proves the practical application of MCS into the BEAVRS benchmark two-cycle depletion simulations.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.763-775
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• 2021

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.