Challenge Analysis and Schemes Design for the CFD Simulation of PWR

Chen, Guangliang; Zhang, Zhijian; Tian, Zhilong; Li, Lei; Dong, Xiaomeng

doi:10.1155/2017/5695809

Cited by 3 publications

(2 citation statements)

References 13 publications

(15 reference statements)

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“…It has been tested that the grid type has limit influence on the temperature distribution in the single-phase flow (Chen et al, 2017b). But in two-phase flow, this conclusion is doubtful for the complexity of wall boiling status.…”

Section: Grids Turbulence Models and Wall Functionsmentioning

confidence: 99%

Numerical Investigation of the Effect of Grids and Turbulence Models on Critical Heat Flux in a Vertical Pipe

et al. 2018

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Numerical simulation has been widely used in nuclear reactor safety analyses to gain insight into key phenomena. The Critical Heat Flux (CHF) is one of the limiting criteria in the design and operation of nuclear reactors. It is a two-phase flow phenomenon, which rapidly decreases the heat transfer performance at the rod surface. This paper presents a numerical simulation of a steady state flow in a vertical pipe to predict the CHF phenomena. The detailed Computational Fluid Dynamic (CFD) modeling methodology was developed using FLUENT. Eulerian two-phase flow model is used to model the flow and heat transfer phenomena. In order to gain the peak wall temperature accurately and stably, the effect of different turbulence models and wall functions are investigated based on different grids. Results show that O type grid should be used for the simulation of CHF phenomenon. Grids with Y+ larger than 70 are recommended for the CHF simulation because of the acceptable results of all the turbulence models while Grids with Y+ lower than 50 should be avoided. To predict the dry-out position accurately in a fine grid, Realizable k-ε model with standard wall function is recommended. These conclusions have some reference significance to better predict the CHF phenomena of vertical pipe. It can also be expanded to rod bundle of Boiling Water Reactor (BWR) by using same pressure condition.

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Section: Grids Turbulence Models and Wall Functionsmentioning

confidence: 99%

Numerical Investigation of the Effect of Grids and Turbulence Models on Critical Heat Flux in a Vertical Pipe

et al. 2018

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“…Considerable effort has been devoted to simplifying the various systems, significantly reducing the overall size of the installation, the number of auxiliary components, the length of the pipes and the necessary wiring. Likewise, its design allows a modular construction, facilitating its mass production, reducing the total construction time and reducing the initial investment [3].…”

Section: Introductionmentioning

confidence: 99%

Thermo-hydraulic Simulation of AP1000 Nuclear Reactor Fuel Assembly

Martínez

Hernández

García

et al. 2021

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One of the challenges of future nuclear power is the development of safer and more efficient nuclear reactor designs. The AP1000 reactor based on the PWR concept of generation III + has several advantages, which can be summarized as: a modular construction, which facilitates its manufacture in series reducing the total construction time, simplification of the different systems, reduction of the initial capital investment and improvement of safety through the implementation of passive emergency systems. Being a novel design it is important to study the thermohydraulic behavior of the core applying the most modern tools. To determine the thermohydraulic behavior of a typical AP1000 fuel assembly, a computational model based on CFD was developed. A coupled neutronic-thermohydraulic calculation was performed, allowing to obtain the axial power distribution in the typical fuel assembly. The geometric model built used the certified dimensions for this type of installation that appear in the corresponding manuals. The thermohydraulic study used the CFD-based program ANSYS-CFX, considering an eighth of the fuel assembly. The neutronic calculation was performed with the program MCNPX version 2.6e. The work shows the results that illustrate the behavior of the temperature and the heat transfer in different zones of the fuel assembly. The results obtained agree with the data reported in the literature, which allowed the verification of the consistency of the developed model.

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