Conference Agenda

Flashing induced instabilities in vertical systems are one of the most common phenomena that can take place under low-pressure and low-power conditions. Typically, the vaporization process is trigged in the adiabatic riser due to the drop in the hydrostatic pressure. The physics of the flashing induced oscillations have been widely studied experimentally and numerically. However, most of the studied systems have been operated under natural circulation conditions which imposed restrictions in isolating the effect of the flow velocity in the process. Hence, the effect of the flow velocity during forced convection remains uncharted.

In this work, we study flashing induced instabilities under low pressure and low heat flux in a vertical pipe. The tests are conducted in a test loop consisting of a single horizontal heated channel followed a 5 m vertical inverted U-tube section. Sinusoidal flashing induced instabilities have been detected as the flow transitioned from stable single-phase to stable two-phase state. At low power, the oscillations are triggered by flashing and enhanced by subcooled boiling. As the power increases, boiling and flashing coexist. The oscillations amplitude and characteristics as a function of applied power are presented and explained physically.

High-Temperature Gas-cooled Reactor (HTGR) has the advantages of inherent safety and supplying high-temperature process heat. Two-phase flow instability may occur within once-through steam generator (OTSG). Two-phase flow model plays a critical role in predicting the stability boundary. There are several two-phase flow models, including homogeneous equilibrium model (HEM), drift-flux model (DFM) and two fluid model (TFM), etc. DFM is much more precise than HEM when there is a significant velocity difference between the liquid and gas phases, while TFM is very complicated. Consequently, DFM is adopted to deal with the velocity of two-phase mixture region. According to the differences in friction factor and heat transfer factor, the convective heat transfer process in the secondary side of OTSG can be divided into three regions. These three regions are the subcooled water region, the two-phase mixture region and the superheated vapor region. A frequency-domain method is adopted for stability analysis. The essence of DFM is the derivation of void fraction from gas-phase mass conservation equations. The expression of void fraction subsequently leads to the formulation of expressions for mixture density, mixture mass flux and quality of two-phase mixture region and superheated boundary. The transfer function for pressure drop and inlet velocity is derived from momentum conservation equations of three regions using integral and small perturbation method. OTSG can operate stably at the designed power level. Compared to DFM, the stability boundary of superheated evaporation systems predicted by HEM is more conservative. The superheated evaporation systems become stable when distribution parameter or drift velocity increases.

Steam Generator Tube Rupture (SGTR) accident in lead-cooled fast reactor would result in high-pressure subcooled water jetting into the high-temperature melt pool in primary circuit. The intense phase change could trigger a steam explosion, seriously threatening the structural integrity within the reactor. However, there is still a lack of the theoretical study of key physical processes involved in the initial stage of the accident, limiting the development of the safety analysis programs. The study focuses on the three components which are water, steam, and liquid lead-bismuth and establishes theoretical models of jet breakup and phase change steam expansion. The characteristics of jet instability and the variation patterns of key parameters such as breakup time, breakup length and droplet diameter are analyzed. Additionally, the pressure impact on the melt pool generated by the phase change heat transfer and steam expansion is calculated. A comparison with existing research validates the model’s rationality. Results indicate that an increase in jet velocity reduces both the breakup length and droplet diameter, while jet radius has a limited effect on the characteristic parameters of jet. Further, the steam expansion causes the temperature drop and pressure summit in the melt pool. This research provides valuable guidance for assessments of the risk of steam explosion in SGTR accident and for the development of related safety analysis models.

This work describes a transient computational code for prediction of the heat transfer regimes and the channel wall temperature during a transient heating of the channel. The modelling includes the single phase and two-phase heat transfer regimes and post Critical Heat Flux film boiling calculations. The code is based on conservation equations and correlations from literature. A transient heating experiment of water flow inside a stainless-steel tube (1 m length and 8.3 mm in diameter) was used for validation of the model. The flow velocity inside the channel was about 3 m/s, the heating power was increased up to 38 kW and the exit pressure was almost atmospheric. During the experiment, the flow rate, the channel power and the local outer wall temperature of the channel were continuously measured. In the experiment, the channel was connected in parallel to a large bypass, and during the power increase, redistribution of the flow was obtained. Based on the continuously measured values, the model uses a suitable correlation for each regime to calculate the channel thermal parameters (coolant temperature and quality, and the wall temperature). In the single-phase regime, an over-prediction of the experimentally measured wall temperature was obtained, partially due to inaccurate temperature measurement. In the two-phase regime, a good agreement was obtained between the measured temperature values, the temperature trend in time and the model. A new correlation was proposed for the post-CHF regime, based on the calculated void fraction in that zone.

This article focuses on defining an axial void fraction from axial pressure profiles in two-phase critical flow discharges made with CriticAl Flow Test facility (CRAFTY), a novel separate-effect-test facility located in LUT University in Lappeenranta, Finland. Additionally, a method to approximate the flashing delay from the same axial pressure profiles is included. Axial temperature profile is used for detecting local superheat before flashing occurs along the axis of the discharge tube. The discharge tube in CRAFTY resembles the VVER-440 steam generator tube with inner diameter of 13 mm and has length-to-diameter (L/D) ratio of 350 for the used cases. A high subcooling case (ΔTsub~60 °C, pup ~ 8 MPa) and near saturation case (ΔTsub~5 °C, pup ~ 5 MPa) is used in the article. Both cases are rerun as well for increased certainty for analysis.

The axial pressure profiles can offer insights where the continuous liquid phase disperses into non-continous mist/droplet flow. At this transition zone lies the two-phase choke plane as the pressure information from downstream cannot travel into upstream anymore. In one-phase critical flow, the flow velocity reaches Mach 1 making the pressure signal stalled, unable to travel upstream. For two-phase critical flow this analogy is not correct. The two-phase sonic velocities are order of magnitude lower than either the liquid or gas phase.

The working fluid leakage accident in a supercritical carbon dioxide (SCO₂) Brayton cycle system of a reactor will result in the depressurization and discharge of SCO2 from the tube into the atmospheric environment, accompanied by critical flow phenomena of gas-liquid two-phase, which seriously threaten the heat transfer characteristics of the reactor core. The gas-liquid two-phase flow characteristics within the tube determine the size of the critical flow rate at the break. To accurately predict the critical flow rate, a testing technique must be developed to quantitatively measure the gas void fraction of the gas-liquid two-phase flow. In this paper, a phase state measurement method for gas-liquid two-phase flow during the depressurization and flash evaporation process of low conductivity SCO₂ is developed based on a wire mesh sensor assembly. Combined with the established SCO₂ depressurization and discharge experimental setup, the effects of factors such as temperature and pressure on the gas-liquid two-phase flow characteristics within the tube during the SCO₂ depressurization and flash evaporation process are explored. The distribution patterns of gas and liquid phases (bubble size, shape, and velocity, etc.) are analyzed to obtain typical flow patterns of the gas-liquid two-phase flow. Finally, based on the experimental results and key dimensionless numbers, a flow pattern map is plotted, and a prediction model for flow pattern transition boundaries is proposed. This lays the foundation for the study of critical flow phenomena.