Heat transfer characteristics of supercritical water in a 2 × 2 rod bundle – numerical simulation and experimental validation

The present paper is devoted to clarify the effect of buoyancy on the flow and heat transfer of supercritical pressure water flowing in horizontal pipes at supercritical pressures. A series of experiments have been designed and carried out in Xi’an Jiaotong University, Xi’an, China to obtain data in relation to flow and heat transfer of supercritical pressure water in pipes with different arrangements. The experimental parameters are as follows: pressures ranging from 23 to 28MPa, heat flux being up to 600 kW/m2, and the fluid mass fluxes being in the range from 100 to 1000kg/(m2s). In this study, distributions of the local wall temperatures and the local heat transfer coefficients around the circumference of the tube are measured at different cross-sections along the flowing direction. On the basis of the experimental data obtained in the study, some criteria available in open literatures, including Gr/Re2.7, Gr/Re2, and Grq/Grth, are employed to estimate the magnitude of buoyancy and the effect of buoyancy on the flow and heat transfer behavior of the supercritical fluid. It is showed that buoyancy is of particular importance for horizontal flows, but play significantly different role in different regions having different characteristics of the specific heat capacity. Strong buoyancy effect exists in the large specific heat region, but in the enthalpy region which is far away from the LSHR, the discrepancy between the temperature of the top wall and that of the bottom wall is small, indicating that the buoyancy effect can be negligible. Based on the present study, it was found that the criteria Grq/Grth is better than others in terms of the capability of evaluating the effect of the buoyancy on the flow and heat transfer of supercritical water.

Experimental and numerical investigation on heat transfer of supercritical water flowing upward in 2 × 2 rod bundles

Supercritical water reactor (SCWR) is one of the most promising nuclear reactor system among generation IV reactors thanks to its high thermal efficiency and simplicity. One of the main features of supercritical water is the strong variation of thermal-physical properties in the vicinity of the pseudo-critical temperature, which makes it very hard to predict the thermal-hydraulic behavior near this point. In this paper, CFD is used to investigate heat transfer of supercritical water in a 2×2 rod bundle with SST k-ε turbulence model. Two steady-state and one transient cases are simulated. The results show that there is strong non-uniform temperature distribution around the circumferential direction. Heat transfer deterioration (HTD) is found in the front of the heated section along the axial direction when the bulk temperature is near the pseudo-critical point due to secondary flow. Comparision of transient and steady-state flow shows that when the mass flux is less than 700 kg/m2s, the temperature in transient state is smaller than that in the steady-state, especially when the mass flux is 400 kg/m2s, the temperature difference is more than 10 °C.

Large eddy simulation on the heat transfer of supercritical pressure water in a circular pipe

Numerical analysis on supercritical water heat transfer in a 2 × 2 rod bundle

The supercritical water-cooled reactor was proposed as one of the Generation IV nuclear systems. Although many research works on the fluid flow and heat transfer of supercritical water in circular channels have been conducted, there is still lack of research on the fluid flow and heat transfer process in fuel bundles used in supercritical water-cooled nuclear reactors. Besides, fuel bundles have multiple fuel rods, the flow is an external flow, not internal flow as that in circle channels, which will cause the difference in the fluid flow phenomenon and heat transfer on the fuel rod cladding surface. In this work, the heat transfer and fluid flow characteristics of the supercritical water in the single-rod channel and the multi-rod channel are simulated numerically. The results show that there are secondary flows in both channels. The circumferential cladding surface temperature variation is large and should be considered in the future fuel rod design. With the same flow rate and heat flux input, the maximum cladding surface temperature in the multi-rod channel is much higher than that in the single-rod channel. Since the maximum cladding surface temperature is an important parameter for the safety of the nuclear reactor operation, it is recommended to use the multi-rod channel model to conduct numerical simulations for the fluid flow and heat transfer of the supercritical water in the Canadian SCWR.

ABSTRACTThe turbulent mixed convection heat transfer of supercritical water flowing in a vertical tube roughened by V-shaped grooves has been numerically investigated in this paper. The turbulent supercritical water flow characteristics within different grooves are obtained using a validated low-Reynolds number κ-ε turbulence model. The effects of groove angle, groove depth, groove pitch-to-depth ratio, and thermophysical properties on turbulent flow and heat transfer of supercritical water are discussed. The results show that a groove angle γ = 120° presents the best heat transfer performance among the three groove angles. The lower groove depth and higher groove pitch-to-depth ratio suppress the enhancement of heat transfer. Heat transfer performance is significantly decreased due to the strong buoyancy force at Tb = 650.6 K, and heat transfer deterioration occurs in the roughened tube with γ = 120°, e = 0.5 mm, and p/e = 8 in the present simulation. The results also show that the rapid variation in the supercritical water property in the region near the pseudo-critical temperature results in a significant enhancement of heat transfer performance.

Heat transfer experiments of supercritical water in 2×2 rod bundles were conducted based on Chinese Supercritical Water Cooled Reactor(CSR1000) fuel assembly design. The experimental parameters were as follows: the system pressure 23~25 MPa, the mass flow flux 680~1400 kg/(m2?s), and the heat flux 174~968 kW/m2. The experimental results indicated that the heat transfer performance of the bundles reduced as heat flux increasing and mass flow flux decreasing while kept insensitive to the pressure as which varied from 23 MPa to 25 MPa. Moreover, it was found that the heat transfer performance of 2×2 rod bundles was influenced by both the differences from bulk flow to boundary layer in thermal physical properties and the non-uniform of flow and heat transfer in different sub-channels. Based on the experimental data, a satisfactory heat transfer correlation of 2×2 rod bundles was obtained, and about 88.9% of the experimental points had deviations from the correlation within ±25%.

Effects of rib geometries and property variations on heat transfer to supercritical water in internally ribbed tubes

Numerical investigation on deteriorated heat transfer of supercritical water flowing upward in tubes with variable cross-sectional geometries

Analysis of entropy generation behavior of supercritical water flow in a hexagon rod bundle

Experimental study on heat transfer of supercritical water flowing upward and downward in 2 × 2 rod bundle with wrapped wire

Nonuniform heat transfer of supercritical water in a tight rod bundle – Assessment of correlations

Numerical investigation of diameter effect on heat transfer of supercritical water flows in horizontal round tubes

Numerical investigation of heat transfer in upward flows of supercritical water in circular tubes and tight fuel rod bundles