Abstract
In nuclear reactor system research, the multiscale coupled thermal-hydraulic (T-H) system code and CFD code is one of the most prevalent research areas, and it could help improve simulation fidelity and optimize nuclear reactor design. Additionally, a new idea known as the function fitting method (FFM) for coupling parameter distribution has been newly proposed for exchanging data on the coupling interface, which uses math equations to present the velocity distribution characteristics at the coupling interface. This method could improve the simulation error and numerical instability. To verify and validate the abovementioned FFM, a comparison between the velocity function shape by FFM and real velocity distribution was completed. Besides, the Edwards pipe blowdown test results were used to verify the coupled code. The results showed good agreement with experiment results, and a better simulation accuracy compared to previous work. The current work will establish the ability to explore multiscale coupled thermal-hydraulic operation characteristics which permit precise local parameter distribution.
Highlights
Nuclear safety is a top priority for nuclear power application and expansion
As computational resources have dramatically developed, component scale analysis codes like COBRA (Stewart et al, 1977), RELAP5-3D (RELAP5-3D Code, 2012), VIPRE (Stewart et al, 1989), and local scale codes such as Fluent (Rohde et al, 2007), CFX (Höhne et al, 2010), and Star-CCM+ (Cardoni, 2011) have emerged. These computational fluid dynamic (CFD) codes can provide three-dimensional features, which have been applied in pressurized thermal shock (PTS) (Egorov et al, 2004), boron dilution and distribution in reactor vessels (Muhlbauer, 2003; Scheuerer et al, 2005), and so on
The RELAP5 code sets the boundary conditions through time-dependent control volume (TMDPVOL) and time-dependent junction (TMDPJUN); while Fluent has a predetermined velocity-inlet, pressure-outlet, and outflow boundary conditions, etc
Summary
Nuclear safety is a top priority for nuclear power application and expansion. Estimate tools for increasing safety analysis and evaluation requirements need to become better and more precise. In the past few decades, on a system scale, the best estimate codes such as RELAP5 (Allison et al, 1993), RETRAN (McFadden et al, 1981), CATHARE (Barre and Bernard, 1990), and MARS (Jeong et al, 1999) have dominated nuclear reactor operation, safety analysis, and severe accident analysis. These codes can only present one-dimensional transient system behaviors, which can not provide the local characteristics of the reactor.
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