The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube–type, heavy water–moderated, boiling light water–cooled, natural-circulation–based reactor. The fuel bundle of AHWR contains 54 fuel rods arranged in three concentric rings of 12, 18, and 24 fuel rods. This fuel bundle is divided into a number of imaginary interacting flow passages called subchannels. A single-phase-flow condition exists in the reactor rod bundle during the start-up condition and up to a certain length of rod bundle when it is operating at full power. Predicting the thermal margin of the reactor during the start-up condition has necessitated the determination of the turbulent mixing rate of the coolant among these subchannels. Thus, it is vital to evaluate the turbulent mixing between the subchannels of the AHWR rod bundle. In this paper, experiments were carried out to determine the turbulent mixing rate in the simulated subchannels of the reactor. The size of the rod and the pitch in the test were the same as th...

The present study investigates the behaviour of thermal streaks on a heated foil which is cooled with turbulent flow in a square duct channel. Real-time infrared thermography is used to visualize and measure the spacing between the thermal streaks. A stainless-steel foil with a thickness of 25 microns is cooled by water. The experiments were performed in a range of Reynolds numbers from 5000 to 20000 and Prandtl numbers from 3 to 7. The mean temperature, root-mean-square of temperature and autocorrelation function have been calculated and used to measure the average thermal streak spacing and power spectra in the spanwise and streamwise directions. The root mean square temperature was 0.3 °C to 0.5 °C which corresponds to roughly 10% of the mean temperature difference between foil and water. The uncertainty in mean temperature difference and root mean square temperature was around 5% and 10%, respectively. The measured thermal streak spacing was 100 wall unit to 180 wall unit under the present experimental range. The uncertainty in measured thermal streak spacing was around 2.5%. The effects of Reynolds number, Prandtl number and heat flux on the thermal streak spacing and also on the statistics of the temperature field have been presented and discussed in this paper. A new correlation has been proposed to predict the dimensionless thermal streak spacing. The error in the prediction is estimated within ± 15 %.

The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube–type, heavy water–moderated, and boiling light water–cooled natural-circulation–based reactor. The fuel bundle of AHWR contains 54 fuel rods arranged in three concentric rings of 12, 18, and 24 fuel rods. This fuel bundle is divided into a number of imaginary interacting flow passages called subchannels. Transition from a single-phase-flow condition to a two-phase-flow condition occurs in the reactor rod bundle with increase in power. Prediction of the thermal margin of the reactor has necessitated the determination of intersubchannel mixing due to void drift. Void drift is due to redistribution of the non-equilibrium void fraction to attain an equilibrium void fraction. This redistribution occurs in the reactor rod bundle until it reaches the state of equilibrium void fraction. Hence, it is vital to evaluate void drift between subchannels of AHWR rod bundles. In this paper, experiments were carried out to investigate the void drift phenomena in simulated subchannels of AHWR. The size of the rod and the pitch in the test section were the same as those of the actual rod bundle in the prototype. Three subchannels are considered in 1/12th of the cross section of the rod bundle. Water and air were used as the working fluid, and the experiments were carried out at atmospheric condition without the addition of heat. The void fraction in the simulated subchannels was varied from 0 to 0.8 under various ranges of superficial liquid velocities. The void drift between the subchannels was measured. The test data were compared with existing models in the literature. It was found that the existing models could predict the measured equilibrium void fraction in the rod bundle of the reactor within the range +8% to −14%.