Based on the external force method, the present study integrates the nonlinear dynamic model developed previously by the authors with vertical seismic accelerations to investigate the seismic-induced effect on the single nuclear-coupled boiling channel natural circulation loop. The natural frequencies of the states in the stable region are widely explored through nonlinear analysis. The results indicate that the natural frequency of initial state tends to increase as the increase in the phase change number (operating power) or as the decrease in the subcooling number (inlet subcooling). As supposing that a real seismic acceleration is directly imposed on the system, some parametric effects on the seismic-induced oscillations are performed in the present natural circulation loop system. The seismic-induced oscillations are found to be consistent with the combined results of the inherent system stability characteristics and the impact of external seismic acceleration. The system with a larger outlet loss coefficient of the heated channel, or a longer heated channel, would destabilize the seismic-induced oscillations, while the inlet loss coefficient of the heated channel has a stable effect on the system. A much stronger resonance oscillation could be induced by the increase in the core inlet subcooling. Notably, the enlargement and contraction of the heated channel diameter would move the natural circulation system closer to type-I and type-II instability regions, respectively. These both generate unstable effects on the seismic-induced oscillations due to the inherent stability characteristics of the initial states.

The jet impingement cooling is always preferred over the other cooling methods due to its high heat removal capability. However, rapid quenching may lead to the formation of cracks and poor ductility to the quenched surface. Mist jet impingement cooling offers an alternative method to uncontrolled rapid cooling, particularly in steel and electronic industries. In mist cooling, the droplets are atomized by compressed air. Experiments are performed under transient conditions using two full-cone spray nozzles ( Lechler Pneumatic atomizing nozzle 136.115.xx.A2 and 136.134.xx.A2) to study the effect of subcooling and nozzle diameter on surface heat flux. The hot surface used for the experiment is a stainless steel foil (AISI-304) of thickness 0.15mm. The initial surface temperature of the plate is maintained at 500±10°C with the help of an AC transformer. Infrared thermal image camera (A655sc, FLIR System) is used for data estimation. The IR camera and the nozzle are positioned on either side of the plate. The variation in surface temperature has been acquired at 8 different spatial points. It has been observed that that as we move away from the stagnation point then irrespective of air and water flow rates surface heat flux decreases. The maximum surface heat flux obtained at the stagnation point. With the increase in diameter surface heat flux increases irrespective of pressure values. The correlation between q m /q stag experimental and predicted values has been shown.

Nanofluid has been attracted great attention since it was proposed as a preeminent working fluid. Flow boiling is familiar in heat transfer system and the critical heat flux is a key parameter for the design of thermal hydraulic. In present work, the critical heat flux of nanofluid flow boiling is experimentally investigated in a vertical tube with the consideration of outlet pressure, mass flux, inlet subcooling, heating length and diameter. The results indicate that the critical heat flux of nanofluid flow boiling is enhanced compared with base fluid and the increasing radio is increased with increasing the mass flux, diameter and pressure, and with decreasing the heating length. In addition, the inlet subcooling and concentrations (0.1vol.%, 0.5vol.%) have almost no significant influence. Furthermore, a new mechanism for the enhancement of nanofluid flow boiling critical heat flux was proposed by the SEM images of nanopariticle deposition on the heating surface.

Subcooled boiling has been investigated by using the RPI wall boiling models in the last two decades. High accuracy of such models has been achieved by improving the submodels for interphase actions or tuning the model parameters. However, the reliabilities of the models are still suspicious due to the limited model validation and the experimental data based model calibration. The applicability of calibrated model parameters in the new experiment data can not be assured. The effects of model parameters need to be calibrated were treated as the uncertainties of these parameters. The critical parameters that dominate the prediction of subcooled boiling were selected by using the hierarchy analysis. After that, the input samples for uncertainty analysis were obtained by an efficient Monte-Carlo sampling technology — Latin Hypercube Sampling based on the hypothetic normal distribution. Then, the samples were transferred into the FLUENT code for CFD calculations. Results from CFD code were extracted for statistical analysis. Besides, the uncertainties from boundary conditions were also analyzed to quantify the effects of experimental uncertainties. The dependency of the predicted subcooling parameters and the input parameters can be obtained. A PIRT table can be drawn from the generated correlation coefficients between inputs and outputs to quantify the importance of model parameters on subcooled boiling.

Two-fluid model is a common method to simulate the subcooled flow boiling heat transfer, in which the wall boiling model is mainly used for the partition of wall heat flux and the mass transfer between two phases on the wall. The model determines the amount of vapor phase and predicts the cross-sectional void fraction in the channel, nucleate site density and bubble departure diameter play an important role in the accurate prediction of wall boiling model. Eulerian two-fluid model coupled with Rensselaer Polytechnic Institute (RPI) wall boiling model is employed to simulate the heat transfer characteristics and boiling phenomena in vertical narrow rectangular channels by using FLUENT code. Based on the experimental data of subcooled boiling in vertical narrow rectangular channel, different combinations of nucleate site density and bubble departure diameter correlations are used to calculate under different conditions of heat flux and inlet subcooling. Comparing the calculated heat transfer coefficients along the vertical height with experimental results, it can be found that these two parameters have a significant effect on the subcooled boiling heat transfer in narrow rectangular channels. Different parameter combinations lead to differences in wall heat flux distribution, different heat flux and inlet subcooling also have different effects on these models, which eventually lead to different evaporative heat flux, thus affecting the prediction of void fraction.

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.

The melt jet breakup phenomenon in a pre-flooded reactor cavity during a severe accident is related to the debris bed coolability. It is important to predict the jet breakup length for the evaluation of the debris bed coolability. A large volume of works on the jet breakup length have been performed. However, the consistency between experiments and correlations was difficult to achieve. Some data follow the Saito correlation (include Froude number in the correlation) and others follow the Epstein correlation (doesn’t include Froude number). The separation of the jet breakup length correlation along the water subcooling was reported based on the experimental data using the low melting temperature materials in our previous works. Since the previous experiments show an unclear jet shape before entering the water pool which could be an uncertainty factor, a slide gate system for a clear jet shape was additionally installed. Experiments were conducted with the similar condition of previous work and different initial condition of melt jet. With a clear jet shape, the jet breakup length in the subcooled water show different tendency following the Saito correlation. To figure out the effect of the entry condition of the melt jet, the jet diameter and the method of estimating the jet breakup length are revisited. Our previo0us experiments show large uncertainties on the jet diameter, leading to the large discrepancy of the dimensionless jet breakup length. Also, early broken jet core is reported in subcooled water cases. Thus, the uncertain characteristics of the jet breakup length analysis is discussed in this paper including the jet diameter and the method to estimate the jet breakup length.

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.