# CFD Analysis of Subcooled Flow Boiling in 4 x 4 Rod Bundle
## 1 Introduction  
Computational fluid dynamics (CFD) is being increasingly used in the analysis of three-dimensional two-phase flow for the safe operation of a nuclear power plant. In recent years, subcooled boiling in the rod bundle is one of the important phenomena observed in small break loss of coolant accidents (SBLOCA) for pressurized water reactor (PWR) and is also expected in the loss of heat sink accident of nuclear spent fuel pool (SFP) as a safety issue. Thus, experiments and measurements on multiphase turbulent flow in pipes or rod bundles have been conducted, and numerous fluid properties have been studied. In recent years, various simulations have been conducted in various papers [1; 2; 3; 4; 5]. Numerical analysis of subcooled two-phase flow at high pressure has confirmed that the Eulerian multiphase model including a Reynolds stress turbulence model is suitable for predicting the actual phenomenon [6; 7; 8; 9; 10]. Single-phase convective heat transfer and vapor-water two-phase boiling heat transfer experiments have been conducted in a vertical seven-rod bundle at a low flow rate near atmospheric pressure. In this case, bubble generation is dominantly determined by the enhancement factor of forced convection by the inlet mass flux and the suppression factor of nucleate boiling. Thus, it was necessary to react sensitively to the fluid velocity, and in the next step, excessive bubble prediction is prevented through a nuclear boiling suppression factor. By taking in account the enhancement factor of forced convection and the suppression factor of nucleate boiling, a new correlation in terms of boiling number, Reynolds number, and Martinelli number was found to be effective in predictingtwo-phase-flow boiling heat transfer coefficients [11]. Analytical studies using various turbulence models for rod bundles in various shapes have also been conducted. A \(3\times 3\) rod bundle flow was analyzed using the k-\(\epsilon\) model, shear stress transport k-\(\omega\) model, and Reynolds stress model; and the pressure rate, flow velocity, and temperature were predicted [12]. For supercritical pressure conditions, the importance of the turbulence mixing factor was studied thermal hydraulics characteristics of \(2\times 2\) rod bundle by numerical analyzing [13]. In total of 2000 mm length of the \(4\times 4\) rod bundle with a mixing grid installed at 2/3 height, turbulence was processed in a Reynolds-averaged Navier-Stokes method using a nonlinear and viscous model to simulate CFD [14]. A study on the effect of spacer grid distance considered the influence of spacer grid blockage ratio and spacer grid spacing on heat transfer in a vertical rod bundle [15]. Reynolds number of the working fluid in this paper is 24,800 to 30,100 to investigate the flow effect on heat transfer. As such, this study was carried out with reference to other studies on the two-phase model, turbulence model, and various rod bundle shapes and grids. The present study addresses the local flow inside the rod bundle. As the inside of a bundle of nuclear fuel rods has a very narrow flow path, a two-phase flow study using three-dimensional heat transfer analysis and a boiling model is required for reactor safety design and analysis. For two-phase flow analysis, the bubble size is a very important factor for accurately predicting bubble behavior in the flow path. In line with this focus, this study analyzes the results of the numerical analysis of the volume fraction of vapor (VF) and interfacial area concentration (IAC) compared with experimental values. VF is the volume of vapor in volume rate and IAC is defined by VF and interaction length scale. It focuses on the analysis of coolant boiling in a narrow flow path inside a nuclear fuel reactor \(4\times 4\) rod bundle. This phenomenon is solved by the Eulerian multiphase model as a segregated flow. The important elements to accurately predict the behavior of vapor are the interfacial momentum transfer force and wall boiling model of heat and mass transfer force.