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In the published experimental results, it has been observed that when high-speed steam spurt into the subcooled waterflow, the total pressure along the axial direction at trailing edge of the steam plume shows a pressure-lift. To reveal the mechanism behind this phenomenon, this study utilizes a particle model to investigate the pressure profile of steam jet condensation in subcooled water flow in a narrow pipe. A numerical model based on the Eulerian–Eulerian multiphase model has been developed to accurately simulate the characteristics of pressure profile along the axial direction. The model's validity is established by comparing the steam plume shape and temperature profiles with the experimental data. By analyzing the total pressure profile of the axis and the contours of gas volume fraction, it is found that there exists a pressure-lift phenomenon at trailing edge of the steam plume. The dynamic pressure of the water also shows a pressure-lift at this position, so it can be inferred that the dynamic pressure of the water is the main factor of the total pressure-lift. [ABSTRACT FROM AUTHOR]

A centered water nozzle steam–water injector is driven by cold water to pump steam at a low pressure and to produce a high outlet water pressure. It can be used as a safety pump in a light water reactor to inject cooling water into the reactor core with no power supply in case of an accident. In this study, a modified calculation method for a centered water nozzle steam–water injector is proposed and verified by experimental data in the literature. The calculation method consists of a water nozzle model, a steam nozzle model, a mixing section model, and a shock wave model. Comparisons between the calculated results and the experimental results under different inlet steam pressures, inlet water pressures, and back pressures are conducted, and the calculated results show good agreement with the experimental results. The calculated results with different back pressures show that no shock wave occurs in the mixing section when the back pressure is small, but with the back pressure increasing, the pressure undergoes a dramatic increase in the throat tube, and the shock wave position moves towards the inlet of the mixing section. Due to the complexity of shock wave characteristics, it is necessary to conduct a more in-depth study of shock wave characteristics in the mixing section to determine more detailed boundary conditions for shock wave generation. [ABSTRACT FROM AUTHOR]

The lead-bismuth eutectic(LBE)-cooled nuclear systems have significant natural circulation due to its large thermal expansion. Therefore, the passive safety of the reactor in postulated accidents is greatly improved. Moreover, the flow resistance and heat transfer characteristics are important to natural circulation ability of LBE-cooled system. In this paper, the natural circulation capability, the flow resistance and heat transfer characteristics were experimentally studied in the N atural C irculation C apability L oop(NCCL) facility. The research mainly includes four aspects: 1) start-up process of natural circulation; 2) capability of natural circulation; 3) flow resistance of LBE in an annular channel and 4) heat transfer in an annular channel. The results showed that a significant natural convection flow was generated in the experiment. Due to high thermal expansion of LBE, the natural circulation is quickly established and stabilized. The flow rate of natural circulation depends on the loop resistance and the temperature between heating section and cooling section. In our experiment, the maximum natural circulation flow rate was measured to be ∼0.81 kg/s. A new friction coefficient correlation was fitted and the results is slightly larger than the value calculated by the Blasius formula. For the convection heat transfer of liquid LBE in an annular channel, the heat conduction term is dominant, and the Nusselt number increases with the increase of Peclet number. Fitted experimental correlation of convection heat transfer of liquid LBE in annular channel is given. • The flow resistance and heat transfer experiments were conducted in a lead bismuth eutectic loop. • The start-up process of natural circulation and capability of natural circulation were studied in this experiment. • Fitted experimental correlations of friction coefficient and convection heat transfer of LBE in annular channel were given. [ABSTRACT FROM AUTHOR]

The Steam-Water Co-Current Flow (SWCOCF) in the AP1000 Pressurizer Surge Line (PZR SL) are important phenomena during the depressurization stage of the Small Break Loss-of-Coolant Accidents (SB-LOCAs), which affect the predictions of SWCOCF pressure drop in PZR SL and safety analyses. To investigate the pressure drop characteristics of SWCOCF in the PZR SL of AP1000, a PZR SL test model made of stainless steel is designed and a test loop is also established. The test section is mainly composed of a long inclined tube with a short acrylic visual section, a whole surge line tube and a vertical section. Two experiments were conducted: the single phase water experiment and the SWCOCF experiment. The total friction factor and the full-liquid two-phase friction multiplier as well as the liquid-only two-phase friction multiplier of the PZR SL tube were obtained, separately. The research results show that: the whole surge line two-phase friction multiplier increases with the increasing of mass quality, and decreases with the increasing of pressure. Two prediction models of two-phase friction multiplier for the surge line were developed based on the Martinelli-Nelson and Lockhart-Martinelli method, respectively. The fitted full-liquid two-phase friction multiplier model and the liquid-only two-phase friction multiplier model can predict the two-phase friction pressure drop in PZR SL very well. This paper may provide contributions to the design and safety analysis of AP1000 surge line. [ABSTRACT FROM AUTHOR]

To investigate Counter-Current Flow (CCF) characteristics in the Pressurizer Surge Line (PZR SL) assembly of the large advanced passive nuclear reactors, steam-water CCF experiments are carried out on a stainless steel test section, which is a 1:4 scaled-down model of the AP1000 PZR SL assembly. The inner diameter of the test PZR SL pipe is 90 mm, which also belongs to the large-diameter pipes, as same as the prototype PZR SL pipe, in the field of Counter-Current Flow Limitation (CCFL) researches. The present steam-water CCFL experiments are conducted under normal pressure and saturated temperature, with the PZR simulator collapse water level ranging from 350 to 900 mm. CCFL becomes severer at higher steam flow rate, and the most limiting CCFL effect locates between the PZR simulator bottom head and the vertical part of PZR SL pipe. The onset of CCFL and zero liquid penetration (ZP) are two critical conditions in the CCF development process, dividing the process into three stages: Before-CCFL, Partial-CCFL, and CCFL-ZP. According to the local CCFL conditions, the development of the CCFL process is also divided into four-regions since onset of CCFL. The present steam-water CCF data are well normalized in terms of dimensionless Kutateladze (Ku) numbers, and a Ku-type empirical partial CCFL correlation is developed. The comparisons of the present CCF data and partial CCFL correlation with the CCF data and correlations of former experiment researches validate that the present empirical partial CCFL correlation is conservative to predict steam-water partial CCFL in the prototype PZR SL assembly of the large advanced passive nuclear reactors. [ABSTRACT FROM AUTHOR]