CFD modelling of supercritical water flow and heat transfer in a 2 × 2 fuel rod bundle

Numerical Verifications on Heat Transfer to Supercritical Water Flowing Upward in a 4-m Long Bare Vertical Tube

Thermal efficiency and safety of generation-IV nuclear-power-reactor concept supercritical water-cooled reactor (SCWR) are largely dependent on the coupled supercritical water (SCW) thermophysical properties and heat transfer performance in the supercritical region. This paper presents the numerical investigation of the heat-transfer characteristics of SCW flow in a 4-m long circular tube (ID = 10 mm) based on computational fluid dynamics. Numerical model for SCW was established in this analysis and forced-convection heat transfer was studied at different operating conditions. The data were collected at pressure of about 24 MPa, inlet temperatures from 320 to 350 °C, mass flux from 1000 to 1500 kg/m2·s, and heat flux up to 1500 kW/m2. Results of numerical simulation predict the experimental data with reasonable accuracy. A dimensional analysis was conducted to derive the general form of an empirical supercritical water heat-transfer correlation. The decrease of turbulent viscosity due to the decrease of density leads to a lower turbulent diffusion and turbulent kinetic energy, which inhibits heat transfer. The increased wall temperature and localized heat transfer deterioration (HTD) would occur as the liquid in the core of the tube is isolated for the low-density fluid adheres to the near-wall region, which is characterized by low thermal capacity.

Further Details of a Numerical Analysis on the Thermal Hydraulic Effect of Wrapped Wire Spacers in Fuel Bundle

The application of relatively simple and cheap wrapped wire spacer in the European supercritical water-cooled reactor (SCWR) (high-performance light water reactor (HPLWR)) has been proposed in order to provide enhanced heat transfer in the fuel assembly without unacceptable penalty in pressure loss. The wires cause twisting flow in the fuel assembly, which means the coolant not only flows straight in the axial direction but also has a significant transverse velocity component, and strong mixing between neighboring subchannels occurs. The aim of this ongoing research is to numerically investigate the effect of wrapped wire spacers on thermal hydraulics of the turbulent coolant flow and its heat transfer in a small bundle of four fuel rods. One bare and six-wired geometries with varying wire pitches (1–6 turn(s) of wires) have been studied. It was found that the wires generate significant amount of transverse velocity, decrease the wall temperature, and increase the heat transfer coefficient mostly in corner subchannels which were the hottest in bare geometry. Thus, the presence of wires enhances heat transfer where it is most needed. Temperature hot spots with moderate values have been identified on the cladding wall of fuel rods. Based on the results, a technically optimal choice of number of wire turns from thermal hydraulic sense has been proposed.

A Numerical Study of Turbulent Upward Flow of Super Critical Water in a 2 × 2 Rod Bundle With Nonuniform Heating

This work is part of a benchmarking exercise organized by an IAEA in supercritical water-cooled reactor (SCWR) thermal-hydraulics aimed at improving the understanding and prediction accuracy of the thermal-hydraulic phenomena relevant to SCWRs. An experiment carried out using a 2 × 2 SCWR bundle at University of Wisconsin-Madison was modeled using an open-source computational fluid dynamics (CFD) code—Code_Saturne. The k–ω shear stress transport (SST) model was used to account for the buoyancy-aided turbulent flow in the fuel channel. Significant heat transfer deterioration (HTD) was observed in the boundary layer, which is commonly expected to occur in buoyancy-aided flows. For comparison, simulations were also conducted using ansysfluent with similar model setups.

Computational Fluid Dynamic Simulations of Heat Transfer From a 2 × 2 Wire-Wrapped Fuel Rod Bundle to Supercritical Pressure Water

Within the Generation-IV International Forum, Canadian Nuclear Laboratories (CNL) led the conceptual fuel bundle design effort for the Canadian supercritical water cooled reactor (SCWR). The proposed fuel rod assembly for the Canadian SCWR design comprised of 64-elements with spacing between elements maintained using the wire-wrap spacers. Experimental data and correlations are not available for the fuel-assembly concept of the Canadian SCWR. To analyze the thermalhydraulic performance of the new bundle design, CNL is using computational fluid dynamics (CFD) as well as the subchannel approach. Simulations of wire-wrapped bundles can benefit from the increased fidelity and resolution of a CFD approach due to its ability to resolve the boundary layer phenomena. Prior to the application, the CFD tool has been assessed against experimental heat transfer data obtained with bundle subassemblies to identify the appropriate turbulence model to use in the analyses. In the present paper, assessment of CFD predictions was made with the wire-wrapped bundle experiments performed at Xi'an Jiaotong University (XJTU) in China. A three-dimensional CFD study of the fluid flow and heat transfer at supercritical pressures for the rod-bundle geometries was performed with the key parameter being the fuel rod wall temperature. This investigation used Reynolds-averaged Navier–Stokes turbulence models with wall functions to investigate the behavior of flow through the wire-wrapped fuel rod bundles with water subjected to a supercritical pressure of 25 MPa. Along with the selection of turbulence models, CFD results were found to be dependent on the value of turbulent Prandtl number used in simulating the experimental test conditions for the wire-wrapped fuel rod configuration. It was found that the CFD simulation tends to overpredict the fuel wall temperature, and the predicted location of peak temperature differs from the measurement by up to 65 deg.

Thermal-Hydraulics Program in Support of Canadian SCWR Concept Development

The thermal-hydraulics program in support of the development of the Canadian supercritical water-cooled reactor (SCWR) concept has undergone several phases. It focused on key parameters such as heat transfer, critical flow, and stability of fluids at supercritical pressures. Heat-transfer experiments were performed with tubes, annuli, and bundles in water, carbon dioxide (CO2), or refrigerant flows. Data from these experiments have led to enhancement in understanding of the phenomena, improved prediction methods, and verified analytical tools. In addition, these experiments facilitated the investigation of separate effects on heat transfer (such as geometry, diameter, spacing device, and transient). Chocking flow characteristics were studied experimentally with sharp-edged nozzles of two different sizes of opening. Experimental data have been applied in improving the critical-flow correlation in support of accident analyses. A one-dimensional (1D) analytical model for instability phenomena has been developed and assessed against the latest experimental data for quantifying the prediction capability and applicability.