In the hypothetical liquid metal reactor SGTR (Steam Generator Tube Rupture) accident, the high-pressure water in steam generator tube will be injected into the high-temperature liquid metal. During the water-liquid metal interaction, water vaporized and vapor bubbles migrated into liquid metal, which could be resulting a sharp pressure raise and threatening the safety of the reactor. This multi-phase system can be abstracted as a system in which a high-pressure subcooled fluid (overheating state at atmospheric pressure) is injected into another higher temperature fluid (normal pressure) to form a two-component (water and liquid metal) system in which multiphase (fluid and vapor) were existed. There are two effect could cause bubbles expansion which lead to overall system pressure increase, they are flash evaporation of pressurized water since the depressurization and high-temperature glycerin heating the water/vapor. In this paper, the theoretical analysis of the single bubble expansion model is used to evaluate the effect of these two phenomena on the system pressure rise. First of all, this paper uses the force balance point of view to derive the single bubble expansion equation inside the inviscid fluid, and obtain the relationship between bubble expansion speed and bubble internal pressure. Secondly, through conservative assumptions, the bubble expansion speed caused by the heating effect of the high temperature fluid on the bubble is evaluated. Finally, this paper compares the strength of these two effects in the expansion of a single bubble, and theoretically explains the driving force of the pressure increase in the simulation experiment and the potential flow pattern.

Bubbly flow or foam flow could prevail in the subcooled flow boiling of seawater in an annulus. Compared with the bubbly flow of de-ionized water in the subcooled flow boiling, the bubble size of seawater is smaller. Further, bubbly flow with small bubbles or foam flow in a subcooled flow boiling of seawater would be affected by different heat fluxes, mass fluxes and inlet subcooling. The change in the boiling behaviors due to different conditions can be reflected in terms of variations of bubble characteristics. However, details of the characteristics of bubbles in seawater generated by the two-phase subcooled flow boiling are still limited. In the present study, bubble characteristics are measured and detailed by an image analytic method in the software Image J. The subcooled flow boiling experiments with different heat fluxes are conducted in a vertical upward annulus with artificial seawater at the inlet temperature of 90°C and mass flux of 1200 kg/m 2 s. The two-phase flow patterns are recorded by the high-speed camera in 6000 frame/s. To conduct measurements of bubble characteristics using Image J, images are selected as one set of measurement from the videos recording under different heat fluxes. After preprocessing of images, measurements of bubbles and fitting procedures with ellipses are conducted. Averaged values of descriptors of bubbles are obtained, including areas, perimeters, lengths of major axes, lengths of minor axes, circularities, angles, solidities, and aspect ratios, etc. It provides a possible way to investigate the details of bubbly flow or foam flow in the subcooled flow boiling of seawater.

Critical heat flux (CHF) is one of the most significant thermal criteria for nuclear reactor design and safety in subcooled flow boiling. The accurate prediction capabilities of the characteristic size of departure bubbles are crucial for predicting the net vapor generation point (NVG) and CHF. An experimental research facility was designed to determine the bubble departure diameter and subcooling at the net vapor generation point, not only for vertical flow boiling but also in any orientation between vertical and downward-facing horizontal. An improved force-balanced model is proposed to calculate the bubble departure diameter at the net vapor generation point to predict CHF in vertical subcooled flow boiling. The force-balance equation was solved numerically through iteration to calculate the bubble departure diameter using a novel MATLAB script. The net vapor generation point was modified using the bubble departure diameter obtained from the proposed model. The model agrees well with the experimental data. Finally, the modified departure diameter and NVG were applied to the liquid sublayer dryout model to predict the CHF on upward subcooled flow boiling. It was found that the model predicts the experimental CHF data with an average relative error of 7.17%.

Direct contact condensation (DCC) is a common physical phenomenon appearing in the nuclear power plants and other industrial applications. The current research on DCC focuses on steam-water counterflow or jet flow under forced flow conditions. For some natural-circulation passive safety systems in floating nuclear power plants, the heated section is connected with the heat sink by the horizontal pipes. The heat sink is usually the ocean. In such natural circulation systems (NCSs), the steam produced in the heated section will inevitably contact the subcooled water in the ocean and result in DCC event. In this paper, the fundamental characteristics of two-phase flow were described in detail. In addition, the influences of the subcooled water temperature on the flow rate, outlet temperature, steam bubble behaviors, and pressure surge were emphatically discussed. The experimental results revealed that the subcooled water in the water tank could reversely flow into the pipe, which was able to cause condensation induced water hammer (CIWH) event and flow oscillation. Further research showed that the subcooled water was evidently prevented from reversely flowing into the pipe with the increase in the subcooled water temperature. The position where the DCC event occurs transferred from the pipe to the water tank, and the pressure peak obviously reduced, even disappear when the subcooled water temperature is larger than 61 °C.

The effects of rolling condition on the flow instability characteristics of natural circulation (NC) in rod bundle channel were experimentally studied. A 3 × 3 rod bundle channel is used as the testing section. The experimental system pressure range is 0.2 to 0.6 MPa, and the range of inlet subcooling is 10 to 70 °C. The ranges of rolling motion amplitude and period are 10 ∼ 20° and 10 ∼ 30s, respectively. Two typical two-phase flow instabilities in rod bundle channel under rolling condition were found in experiments: (a) the trough-type oscillation caused by the vapor generation at the minimum point of flow fluctuation and (b) the compound oscillation formed by the superposition of the trough-type oscillation and DWO I . Experimental results show that the rolling motion can reduce the threshold heating power of trough-type oscillation and cause the occurrence of NC flow instability in advance. But the rolling motion cannot affect the dimensionless boundary of DWO I in rod bundle channel.

Quenching is an important phenomenon in the evaluation of an emergency core cooling system following a hypothetical loss of coolant accident (LOCA) in a nuclear reactor. In the present study, an experimental apparatus is designed and constructed with the purpose of conducting high-temperature transient pool boiling quenching experiments for zirconium (Zr-4) cylindrical test samples. Three thermocouples are inserted in the test sample to investigate the effect of axial distance on the minimum film boiling temperature. The Zr-4 rodlet is heated up to a temperature well above the minimum film boiling temperature (up to 600°C), and then plunged vertically in a quiescent pool of subcooled water. A data acquisition system is used to record the temperature of the embedded thermocouples with time. Data reduction is performed by an inverse heat conduction code to calculate the surface temperature and corresponding surface heat flux. A visualization study with a high-speed camera is conducted to record the quenching behavior on the test sample. It is found that the minimum film boiling temperature decreases with the axial distance, while the CHF temperature is relatively insensitive to the axial distance. The film boiling heat transfer coefficient decreases with surface temperature, and seems to be independent of axial distance. The quench front is observed to originate from the bottom and move upwards. It is found that the quench front velocity remains nearly constant in the lower region of the test sample, and significantly increases in the upper region.

Single bubble growth processes under subcooled boiling condition in a vertical rectangular channel with a gap of 2.8 mm have been visually studied. A high-speed camera was used to observe and record the bubble growth processes at a rate of 6000 frames per second. Four kinds of bubbles with different equivalent radius change trends near the ONB (Onset of Nucleate Boiling) point were observed. The bubble equivalent radius change trends were fitted by traditional empirical formula R ( t ) = k · t n and found that the value of empirical parameter n was in the range of 0.11 to 0.53 which was smaller than that in literatures, and the value of empirical parameter k had a positive correlation with Ja number. The bubbles generated at the same nucleation point under different heating powers, inlet fluid temperatures and mass flow rates were compared and found that within the experimental range the changes of heating power and inlet fluid temperature had a significant effect on bubble growth rectangular channel which may be used in research nuclear reactor and engineering test reactor. Due to the shape and azimuth of the channel, the bubble behaviors in vertical rectangular channel may be different from those in conventional large-size circular channel or horizontal channel. The investigations reported in the literature generally involved multiple bubbles with complex interactions between the bubbles, and the bubble parameters were usually obtained by averaging the parameters of bubbles generated at different nucleation sites. However, even under the same working condition, bubbles generated by two adjacent nucleation sites may also have big differences. In this experiment, the steady single bubble growth processes in vertical rectangular channel with deionized water as the working fluid were visually investigated. Four kinds of bubbles with different bubble growth curves were observed and the equivalent bubble radius change trends at different heating powers, inlet fluid temperatures and mass flow rates were compared.

In the nuclear reactor design, the critical heat flux (CHF) is one of the most important parameters for the reactor safety analysis. The occurrence of CHF will cause a sharp increase in the fuel rod surface temperature, which will result in the failure of fuel claddings and damage of the core. The CHF depends on the local flow conditions and the geometry of the flow channels, which makes the prediction of CHF in a fuel assembly more difficult when considering the cross flow between neighboring channels, spacer grids and mixing vanes. In this paper, the departure from nucleate boiling (DNB) type CHF in rod bundles under motion conditions is investigated based on the coupled analysis of the subchannel method and a CHF mechanism model, namely the liquid sublayer dryout model. The liquid sublayer dryout model assumes that there is a thin liquid sublayer underneath a vapor blanket formed by the coalescence of small bubbles near the heated wall. The dryout of this sublayer is considered as the CHF occurrence. In the liquid sublayer dryout model, sublayer thickness, velocity and length of the vapor blanket are three crucial parameters. In present research, the subchannel code calculates the local flow conditions for the rod bundle and provides input parameters for the liquid sublayer dryout model to predict CHF. In order to verify the method above, the predicted results are compared with the CHF Look-Up Table 2006 (LUT-2006) and a reasonable agreement can be achieved. In addition, the effects of rod bundle inlet coolant mass flow rate, subcooling and motion conditions on the CHF are analyzed.

Similar to conventional nuclear power plant, condensate water subcooling is a common problem in secondary coolant of floating nuclear power plant, which is caused by many reasons. In this article, RELAP5 is used to simulate the phenomenon of condensate water subcooling caused by noncondensable gas. The influence of noncondensable gas to condenser pressure, subcooling temperature, heat transfer rate, terminal temperature difference, cooling water temperature rise is presented. The results obtained through this study have shown that the model with non-condensable gas in steam can simulate condensate water subcooling, and reveal the discipline of condenser heat transfer characteristics as a function of noncondensable gas content.

Steam submerged jetting is an important process in depressurization tank and condenser deaerator tank of nuclear power plant. As the steam contact the liquid water directly, some complicated behaviors such as strong turbulence and phase transition would happens. Especially when the sub-cooling degree is low, the condensation may cause vigorous pressure pulsation and radiation noise, which not only causes noise damage to workers but also affect the safety of the heat exchanger tubes bundle because of vibration transmission. An experiment is proposed to study the complex evolutionary behavior and vibration and noise characteristics of gas-water two-phase flow. The experimental results show that in the case of low subcooling, the mass flow rate of steam has a great influence on gas plume, and, as the mass flow rate increases, the main contribution frequency of noise is gradually increasing from low frequency to high frequency. The researches in this paper can provide the technical basis for the design of the deoxygenation system of condenser in onshore and ship nuclear power plant.

Series of experiments are conducted in a single microchannel, where subcooled water flows upward inside a transparent and vertical microchannel. The cross section of the channel is rectangle with the hydraulic diameter of 2.8mm and the aspect ratio of 20. The working fluid is 3–15K subcooled and surface heat flux on the channel is between 0–3.64 kW/m 2 , among which two-phase instability at low vapor quantity may occur. By using a novel transparent heating technique and a high-speed camera, visualization results are obtained. The parameters are acquired with a National Instruments Data Acquisition card. In the experiments, long-period oscillation and short-period oscillation are observed as the primary types of instability in a microchannel. Instability characteristics represented from signals correspond well with the flow pattern. Moreover, effects of several parameters are investigated. The results indicate that the oscillating period generally increases with the heat flux density and decreases with inlet subcooling, while the effects of inlet resistance are more complex.

The non-condensable gas is used to keep the pressure stable in the steam-gas pressurizer. The processes of heat and mass transfer during steam condensation in the presence of non-condensable gas play an important role and the thermal hydraulic characteristics in the pressurizer is particularly complicated due to the non-condensable gas. The effects of non-condensable gas on the process of heat and mass transfer during steam condensation were experimental investigated. A steam condensation experimental system under high pressure and natural convection was built and nitrogen was chosen in the experiments. The steam and nitrogen were considered in thermal equilibrium and shared the same temperature in the vessel under natural convection. In the experiments, the factors, for instance, pressure, mass fraction of nitrogen, subcooling of wall and the distribution of nitrogen in the steam, had been taken into account. The rate of heat transfer of steam condensation on the vertical wall with nitrogen was obtained and the heat transfer coefficients were also calculated. The characteristics curve of heat and mass transfer during steam condensation with non-condensable gas under high pressure were obtained and an empirical correlation was introduced to calculated to heat transfer coefficient of steam condensation with nitrogen which the calculation results showed great agreement with the experimental data.

The flow rate can fluctuate in offshore nuclear power systems which are exposed to wind and waves, as well as in loops where flow instabilities occur, resulting in different thermal-hydraulic characteristics compared with that under steady flow. Among the thermal-hydraulic characteristics, onset of nucleate boiling (ONB) model determines whether the fluid is boiling, and boiling heat transfer is crucial to equipment performance and safety, both being key issues in subcooled flow boiling. Therefore, an experimental study was conducted to investigate how an imposed periodic flow oscillation affects the boiling inception and heat transfer of subcooled flow boiling of water in a vertical tube. The experiments were conducted under atmospheric pressure with the average flow rate ranging from 96kg/m2s to 287kg/m2s and heat flux ranging from 10kW/m2 to 197kW/m2. The relative pulsatile amplitude range is 0.1–0.3 and pulsatile period range is 10s-30s. Photographic images and thermal parameters such as temperatures and flow rate were recorded. The lack of nucleation site on the heated surface of the test section results in high wall superheat at ONB. The effects of pulsatile amplitude and period on superheat at boiling onset and average heat transfer were analyzed. The results show that the superheat at boiling inception is decreased when the average heat flux is lower than the heat flux at boiling inception of the corresponding steady flow, and the superheat at boiling onset is increased when the average heat flux is higher than the heat flux at boiling onset of the corresponding steady flow. The above effect of flow rate pulsation on superheat increases with increasing amplitude and decreasing period, and the mechanism can be explained by boiling nucleation theory. The lack of large active nucleation site also affects the boiling heat transfer. By comparing the contribution of nucleate boiling to heat transfer with the widely used Cooper’s pool boiling correlation, the subcooled flow boiling was found suppressed by convection. The average heat transfer of both the intermittent flow boiling and the single phase flow is influenced by flow oscillation.

An experimental investigation was conducted in a natural circulation (NC) loop to study the characteristics of two-phase flow instability under low pressure condition. A 3 × 3 rod bundle channel was used as the test section. The effects of heating power, inlet subcooling degree and system pressure on the two-phase NC flow instability types and stable boundaries were studied. The experimental results show that three typical flow conditions can occur in rod bundle channel under NC condition, which are single-phase NC flow, subcooled boiling NC flow oscillation and density wave oscillations (DWO). The oscillation amplitude and period of DWO can be enlarged by increasing the heat flux. Increasing the inlet subcooling degree can increase the marginal heating power of flow instability in NC system. The occurrence of DWO can be suppressed by increasing the system pressure. The flow instability boundary presented by the subcooling number and phase change number was also obtained in present work.

Experimental investigation has been conducted to explore the effect of double nozzle on the characteristics of steam jet condensation. The condensation process and jet noise are recorded by high speed camera and hydrophone, respectively. Five typical condensation regimes are identified based on the shape of steam plume. A condensation map suitable for both single nozzle and double nozzle is developed based on steam mass velocity and water temperature. The maximum expansion ratio of steam plume increases with the rise of steam mass velocity and water temperature. The maximum expansion ratio of double nozzle becomes larger than single nozzle as the water temperature increases. The equivalent diameter of steam bubble in condensation oscillation regime increases with the rise of water temperature. The forming frequency of steam bubble decreases with the rise of water temperature. There is no remarkable difference of the steam bubble equivalent diameter between single and double nozzle. However, the steam bubble forming frequency of double nozzle experiments is smaller than single nozzle experiments. The first dominant frequency decreases with the rise of water temperature and it is very close to the steam bubble forming frequency in condensation oscillation regime. The double nozzle can reduce the first dominant frequency about 5% at the steam mass velocity of 400 kg/m 2 s and 15% at the steam mass velocity of 800 kg/m 2 s.

Mechanical non-homogeneous and thermal non-equilibrium phenomenon exists in two-phase critical flow compared with single phase flow. A one-dimensional two-fluid critical flow model is developed for initially subcooled water flowing in pipe or orifices. The model accounts for thermal nonequilibrium between the liquid and vapor bubbles and for interphase relative motion. In this model, an improved correlation to calculate flashing inception location and surperheat is proposed. The model consists of six conservation equations as well as a seventh equation representing bubble growth in bubbly flow. Closure of the set of governing equations is performed with constitutive relationships which determine the interfacial momentum terms due to mass exchange, wall to liquid and wall to vapour frictional forces, liquid to gas interfacial force and interfacial heat transfer rate. The model considers the development of three flow regimes, namely, bubbly, churn and annular flow regimes. Model predictions compare favorably with experimental data over a wide range of pressures and pipe diameters and lengths.

In predicting the void fraction in subcooled flow boiling, accurate evaluation of single bubble behaviors is of considerable importance. In particular, bubble lift-off velocity affects the void fraction significantly since the bubble disappear quickly due to heat transfer with subcooled liquid if the lift-off velocity is high. In this study, the process of bubble lift-off was experimentally investigated to develop mechanistic correlations for the bubble lift-off velocity. In the development of the correlations, it was assumed that the bubble lift-off velocity in the horizontal direction is proportional to the bubble growth rate and that in the vertical direction is determined primarily by the local liquid velocity evaluated at the bubble center position. Then, impact of the bubble lift-off velocity on the void fraction was explored through numerical simulations. In the simulations, the bubble lift-off velocity in the lateral direction was parametrically changed. It was shown that the mean void fraction decreases with an increase in the lateral bubble lift-off velocity since the bubble condensation is enhanced. It was therefore confirmed that accurate evaluation of the bubble lift-off velocity is important for high accuracy prediction of the void fraction.

Single-phase natural circulation experiments were conducted to study the flow resistance and heat transfer characteristics in a 3 × 3 rod bundle channel with the ratio of rod pitch and rod outer diameter ( P / D ) 1.38. The range of inlet subcooling degree is 30∼90K and the heating power is 1∼20kW. The rods are heated with constant heat flux. According to the experimental results, the flow regime under natural circulation condition is divided and the transition Reynolds number is considered as 800. The flow transition is recognized by the slope change of friction factor curve since the flow transition in the rod bundle channel is not as obvious as that in round pipe. Simultaneously, the flow transition in the rod bundle is much earlier and the upper critical Reynolds number is much larger compared to regular channel like round pipe and rectangular channel. Two correlations for laminar and transition regime are fitted to calculate the friction factor. As for the grid spacer local resistance coefficient, there is slight change at Reynolds number 800 and similarly two correlations are fitted to calculate the local resistance coefficient. The Nusselt number tendency changes at around Reynolds number 4000 but keep unchanged at transition point, which means the flow transition has no obvious effect to the heat transfer. The heat transfer results are compared with different single-phase convective heat transfer correlations. D-B and Gnielinski correlations are not suitable for the heat transfer prediction in rod bundle channel and the relative deviation is more than 20%. Weisman, Presser and Markoczy correlations predict relatively well in high Reynolds number region, and Markoczy correlation is the best of them. In low Reynolds number region, most experimental results are larger than the correlations. D-B correlation based methods may be unsuitable for the heat transfer prediction in rod bundle channel and a new correlation needs to be proposed.

The heat transfer coefficient of the curved surface reflects the external cooling capacity of the reactor pressure vessel under serious accident. This topic used Ansys fluent software 14.5 as simulation tool, mesh generation of the flow area is carried out after the geometric model of the lower head, and then the heat transfer coefficient of the curved surface is numerically simulated. In this paper, the heat transfer coefficient formulas are fitted according to the laminar and turbulent flow, and the influence of the cavity on the heat transfer coefficient is discussed. Calculation results show that, for single-phase flow both laminar and turbulent flow, the heat transfer coefficients increase with the increase of Re, and increase with the increase of Pr, decrease with the increase of theta; According to the comparison of the heat transfer coefficients under different radii of curved surfaces, reducing the radius is beneficial to increase the heat transfer capability; In subcooled boiling flow, the void fraction is less than 21%, the heat transfer coefficient increases with the increases of the void fraction, the void fraction is conducive to heat transfer; but when the void fraction is higher than 21%, it is opposite.

The solid wall convection melting physics is computed with coupling external boiling physics based on the volume-of-fluid model in the STAR-CCM+ code. A three-dimensional rectangular computational sample domain is used with 5.0 cm height, 4.0 cm width and 1.0 cm depth. This domain is separated by carbon steel representing a reactor vessel section in contact with water channel representing the gap between the reactor vessel outer surface and the insulation. The flow rate of the bottom inlet flow in the water flow channel is 2 × 10 −2 kg/sec with temperature at 92 °C. The concentrated heat flux of 2.1 MW/m 2 obtained from a typical APR1400 reactor condition is imposed at the center of the left wall of the solid region. At the same time, subcooled water entering the right channel boils and the vapor film covers the surface which expedites melting process in the solid heated region. Important finding is that solid wall melt-through is not expected as far as for the present analysis conditions and there is a minimum water flow rate below which the rate of solid melting is saturated.