Large-Scale Simulation of Rod Bundles: Coherent Structure Recognition and Stability Analysis

Large Eddy Simulation of a 5 × 5 rod bundle: Impacts of a central control rod thimble tube

Rod bundle flow is an important research field related to reactor cooling in nuclear power plants. Owing to the rapid development of computerized performance assessments, interest in coolant flow analysis using computational fluid dynamics has garnered research interest. Rod bundle flow research data compared with experimental results under various conditions are thus needed. To address this, a boiling model verification study was conducted with reference to experiments. This study adopts the Reynolds-averaged Navier–Stokes equation, a practical analysis method compared to direct numerical simulation and large eddy simulation, including turbulence modeling, to predict the flow of coolant inside a rod bundle. This study also investigates void behavior in low-pressure subcooled flow boiling using a Eulerian approach (two-fluid model). The rod bundle has a length of 0.59 m and a hydraulic diameter of approximately 14.01 mm. At the cross-section at a height of 0.58 m, near the exit, numerical results were compared with the experimental values of the volume fraction of vapor and interfacial area concentration. The simulation results showed good agreement with the experimental data for six different initial conditions with constant density.

Prediction of rod film thickness of vertical upward co-current adiabatic flow in rod bundle

Thermal-hydraulic analysis of LBE flow in rod bundles with wire spacers based on heat transfer framework of OpenFOAM

Turbulent flow of water in a 3×3 rod bundle – Numerical results from RANS, URANS and LES

Large eddy simulation on the turbulent mixing phenomena in 3×3 bare tight lattice rod bundle using spectral element method

Multi-dimensional modeling of two-phase flow in rod bundles and interpretation of velocities measured in BWRs by the cross-correlation technique

The results of measurements of shear stress distribution around wetted surfaces and velocity fields of turbulent flow in rod bundles with geometrical disturbances are presented. Experiments were performed on an aerodynamic model composed of 19 rods in a hexagonal arrangement 6 m long with an outside diameter of 120 mm and a pitch-to-diameter ratio of 1.17. The geometric disturbances were modeled by the displacement of the central rod from its regular position. An important influence of secondary flows on the velocity fields was revealed. From the experimental data, empirical formulas for shear stress distribution were derived.

Numerical approach for quantification of selfwastage phenomena in sodium-cooled fast reactor

As part of an ongoing integrated research project (IRP), detailed investigations of the flow characteristics of an 84-pin hexagonal rod bundle representing a potential modular gas-cooled fast reactor design have been performed. The rod bundle features a pitch-to-diameter ratio of 1.5 and a large central guide tube, along with simple spacer grids and an outer enclosure. The primary focus of the current work is modeling and simulation of this geometry using multiple computational techniques. These include high-fidelity large eddy simulation (LES) and advanced Reynolds-averaged Navier-Stokes (RANS) approaches. The flow predictions are validated at a Reynolds number of 12 000 using matched index of refraction particle image velocimetry experimental data that were also generated by Texas A&M University as part of the IRP. The comparison data include velocity magnitude and turbulent kinetic energy (TKE) at different lateral locations and at multiple distances downstream of the spacer grid. Many characteristics of the experimental flow are replicated in the LES and RANS runs, and a general agreement between simulation and experiment is shown. Except for the immediate vicinity of the grid, LES is able to capture the velocity magnitude within a 10% error and the TKE to within the experimental uncertainty. The LES shows notably better agreement with the experimental data than RANS, particularly regarding the TKE decay behavior downstream of the grid. Both methods show significant departures from the experiment in their predictions close to the central guide tube. The experimental and high-fidelity data will be used to inform closures for novel multiscale methodologies. The long-term goal of the work is to use the high-fidelity data to yield improved multiscale thermal analysis techniques for solving fuel performance problems of direct relevance to industry.

Drift-flux modeling of void fraction for boiling two-phase flow in a tight-lattice rod bundle

Traditionally, the flow regime in two phase flow in rod bundles are considered in a global sense by the visualization. However, a sub-channel flow regime is required to understand and model the two phase flow structures in rod bundles. In this work, a sub-channel impedance meter was designed to get the dynamic feature in the sub-channels, which was applied to identify the sub-channel flow regime by the fast search and finding peaks of the cumulative probability distribution functions (CPDFs) objectively. In the present study, five flow regimes, namely bubbly flow, quasi-cap bubbly flow, quasi-slug flow, cap-turbulent flow and churn-turbulent flow were defined and recognized. The sub-channel flow regimes at the same cross section were compared to each other, which show similar feature with the local flow regimes in pipe. It is possible to identify nine different global flow regime configurations by combining the corner, side and inner sub-channel flow regime at the same cross section, which was drawn in the 2D sub-channel flow regime.

Thermal–hydraulic performance of a 5 × 5 rod bundle with spacer grid in a nuclear reactor

Pressure drop performance of rod bundles in hexagonal arrangements

A review of CFD studies on thermal hydraulic analysis of coolant flow through fuel rod bundles in nuclear reactor