Development and Verification of a Computational Fluid Dynamics Model for a PWR-type Small Modular Reactor Subchannel

• Published in: Nuclear engineering and design 2023-10, Vol.412, p.112455, Article 112455
• Main Authors: Souto, Normando Perazzo Barbosa, de Oliveira, Claudio Luiz, Vellozo, Sergio de Oliveira, Cabral, Ronaldo Glicério
• Format: Article
• Language: English
• Subjects: ANSYS Fluent; Computational fluid dynamics; Finite Volume Method; Small Modular Reactors; Thermal-hydraulics
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/j.nucengdes.2023.112455

•CFD can improve safety by providing detailed analyses of reactor components.•Verification procedure shows good agreement between numerical and analytical outcomes.•CFD can serve as a valuable tool for future investigation of the proposed SMR concept. With regard to nuclear technology, innovation is critical not only in reducing the current and future costs of deploying Gen III reactors in the short to medium term, but also in enabling the advancement of Gen IV reactors and Small Modular Reactors (SMRs). Considering this aspect, Computational Fluid Dynamics (CFD) can be viewed as an innovative technology for nuclear thermal-hydraulics, increasing the range of possible analyzes with detailed results that are progressively enhanced by advances in processing power, seen in the last decades. This study presents the construction of a CFD model of a subchannel in a pressurized water reactor (PWR) using ANSYS Fluent. The data used for the analysis were obtained from a conceptual design of a Small Modular Reactor, and the development of the model was based on best practice guidelines from existing literature. Turbulence was modeled using the Standard k-ε Model, while a cosine-shaped heat generation profile was used for the fuel. Convergence and mesh studies were conducted to ensure the stability and accuracy of the numerical solution. The results indicated the development of a turbulent profile in the coolant flow, consistent with the inlet conditions. The temperature distribution and pressure drop along the subchannel were also examined. Verification performed by the comparison between the numerical and analytical outcomes revealed low errors. Maximum temperatures in each component remained below the safety limits prescribed for nuclear reactor projects. Based on these results, it can be concluded that the CFD model developed in this study is suitable for thermal-hydraulics analysis of PWR reactors, particularly when examining preliminary assessments of conceptual SMR configurations.