We investigated a single droplet impinging on an inclined heated wall with different inclination angles. A high-speed camera was used to observe this impinging process at 10000 frames / second. The phenomena of the droplet spreading, shrinking, rebounding, boiling, break-up and splashing were observed. The effects of the wall temperature (T w , 40–262 °C), the Weber number of the droplet (We d , 0.66–589) and the wall inclination angle (α, 0–45.6°) on the spreading behavior of the droplet after impinging on the wall were analyzed. Energy conservation equation was used to analyze experimental results. The results show that increasing the inclination angle of the wall is beneficial to the forward spreading of the droplet along the wall, but not to the downward spreading. When α > 45.6°, the droplet will break through the flow resistance of the wall and slide down all the time. The increase of the wall temperature and the Weber number of the droplet is beneficial to the backward and forward spreading of the droplet. Increasing α, increasing T w and decreasing We d will promote the shrinking of the droplet after spreading. In addition, the experimental phenomenon also shows that the larger the wall inclination and the higher the wall temperature, the easier the droplet will break away from the wall.

Water is the most common working fluid using as the coolant for most of nuclear power plants due to its low cost, wide availability, and high heat capacity. In the Fukushima Daiichi accident caused by the tsunami following a powerful earthquake on 11th March 2011, the loss of heat sink occurred. Seawater, which had been considered as an alternative coolant before, was eventually injected into the nuclear reactor for more than one week as an emergency measure. The studies about the characteristics of heat transfer with the seawater as the working fluid have received much more attentions after the Fukushima Daiichi accident. The differences of thermophysical properties such as density, dynamic viscosity, and surface tension, between deionized water and seawater play an important role in the heat transfer. In addition, different boiling behaviors exhibit in the boiling process for two types of working fluid. Compared to pure water, the diameter of bubbles in seawater during pool boiling could be smaller than in water. To improve the safety of a nuclear reactor power plant, mechanisms of the heat transfer using seawater as an alternative emergency coolant should be studied thoroughly. In the present research, experiments of pool boiling in an annulus gap with a riser section after the sudden shutdown of the pump are investigated. Heat transfer characteristic of the heated surface using DI water and 3.5 wt% artificial seawater are described. Boiling behaviors at each fluid are compared and studied.

The design of Pressurized Water Reactor (PWR) fuel has to rely on costly prototypical experimental campaigns, due to the inherent limitations of legacy lumped-parameter thermal-hydraulics codes. By resolving the fine spatial scales, computational fluid dynamics (CFD) methods offer the opportunity of delivering improved fuel design, taking full advantage of three-dimensional mixing and turbulence control, but must extend their predictive capabilities to multiphase conditions up to DNB, where the maturity and accuracy of the methods is still developing. In this paper, the closures developed within the Eulerian-Eulerian two-fluid model framework are assembled and extended to a general boiling formulation to deliver prediction of DNB at reactor conditions. The classic heat flux partitioning approach is advanced through selection of optimal closures for interaction length scale, nucleation site density, and bubble departure diameter, and further extended though the addition of a consistent DNB detection criterion, to reflect the near wall two phase characteristics. Assessment is performed at 138 bar against the experimental dataset for vertical pipe, where the inlet subcooling ranges from 1 to 150 K and the mass flux varies from 600 kg/(m 2 .s) to 2650 kg/(m 2 .s). The presented closures demonstrate consistent agreement with the experimental data, in particular in light of the calibration free application, while further work is ongoing to close the remaining gap and support the introduction of further modeling improvements.

Onset of nucleate boiling — ONB — and critical heat flux — CHF — on subcooled flow boiling under vibration condition, focusing on liquid velocity, amplitude and frequency of vibration, were investigated experimentally. Experiments were conducted using a copper thin-film and subcooled water in a range of the liquid velocity from 0.27 to 4.07 m/s at 0.10MPa. The liquid subcooling was 20K. The test section was set on a vibration table. The acceleration was 1.3 and 4.2 m/s 2 , respectively; the frequency was 2, 4 and 20 Hz, respectively. The test section was arranged for horizontal position facing upward and for vertical position, respectively: the vibration directions were horizontal and vertical to orientation of the heater. The present experimental results showed that critical heat fluxes under the vibration conditions of both vibration directions, i.e., horizontal and vertical vibrations to the heater, were higher than those for steady flow. The CHF under the vibration condition was increased with an increasing of acceleration of vibration. According to present observations, coalesced bubble on the heater was frequently released by vibration of the test heater. This behavior causes the CHF to become higher under the vibration condition.

Boiling heat transfer experiments were performed by using ethanol as test fluid. A heat transfer surface was at the bottom wall of the flow channel. Two types of experiments were conducted; horizontal-rectangular-narrow flow channel experiments and horizontal-narrow-flat space experiments. Experiments were conducted at the pool condition and at 0.1 MPa. In the rectangular-narrow flow channel experiments, the width of the flow channel was varied in the range of 1.0 ∼ 2.0 mm and the height of the flow channel was varied in the range of 0.5 mm ∼ 8 mm. The length of the flow channel was 30 mm. The diameter of the heat transfer surface was the same as the width of the flow channel. Both ends of the flow channel were opened to wide space filled with liquid. In the experiments of the narrow-flat space experiments, the heat transfer surface was at the center of the flat space of 20 mm × 20 mm or 30 mm × 30 mm. The size of the heat transfer surface diameter was 3.0 mm. The space was varied in the range of 0.5 mm ∼ 8 mm. Pool-type experiments of no flat space were also conducted for comparison. The circumference of the narrow-flat space was opened to wide space filled with liquid. In the rectangular-narrow flow channel experiments, a large bubble periodically left from the heat transfer surface and moved to the both ends of the flow channel. As the heat flux was increased, the critical heat flux condition was finally reached. The critical heat flux decreased with a decrease in the flow channel height. When the flow channel height was large, the critical heat flux was close to the pool boiling value. In the experiments of the narrow-flat space experiments, when the space was wide, bubbles generated on the heat transfer surface left freely from the flat space. The heat transfer characteristics were close to those of the pool boiling. As the space became narrow, a large bubble sat on the heat transfer surface, which resulted in the critical heat flux condition. When the space was narrow, the initiation of the bubble generation on the heat transfer surface immediately resulted in the critical heat flux condition.

Onset of nucleate boiling — ONB — and critical heat flux — CHF — on subcooled flow boiling under oscillatory flow and vibration conditions, focusing on liquid velocity, amplitude and frequency of oscillation and vibration were investigated experimentally. Experiments were conducted using a copper thin-film and subcooled water in a range of the liquid velocity from 0.27 to 4.07 m/s at 0.10MPa. The liquid subcooling was 20K. Frequency of oscillatory flow was 2, 4 and 6 Hz, respectively; amplitude of oscillatory flow was 25 and 50% in a ratio of main flow rate, respectively. For the vibration condition, the test section was set on a vibration table. The acceleration was 1.3 and 4.2 m/s 2 , respectively; the frequency was 2, 4 and 20 Hz, respectively. The present experimental results showed that temperature at ONB and critical heat flux for oscillatory flow were lower than those for steady flow. The decreasing of liquid velocity by oscillatory caused the ONB and the CHF to decrease. Critical heat fluxes under the vibration conditions were higher than those for steady flow. The CHF under the vibration condition was increased with an increasing of acceleration of vibration. According to present observations, coalesced bubble on the heater was frequently released by vibration of the test heater. This behavior causes the CHF to become higher under the vibration condition. Furthermore, the effects of oscillation and vibration on velocity profile in a channel were estimated to discuss behaviors of a coalesced bubble on a heater.

Water single-phase and nucleate boiling heat transfer were experimentally investigated in vertical annuli with narrow gaps. The experimental data about water single-phase flow and boiling two-phase flow heat transfer in narrow annular channel were accumulated by two test sections with the narrow gaps of 1.0mm and 1.5mm. Empirical correlations to predict the heat transfer of the single-phase flow and boiling two-phase flow in the narrow annular channel were obtained, which were arranged in the forms of the Dittus-Boelter for heat transfer coefficients in a single-phase flow and the Jens-Lottes formula for a boiling two-phase flow in normal tubes, respectively. The mechanism of the difference between the normal channel and narrow annular channel were also explored. From experimental results, it was found that the turbulent heat transfer coefficients in narrow gaps are nearly the same to the normal channel in the experimental range, and the transition Reynolds number from a laminar flow to a turbulent flow in narrow annuli was much lower than that in normal channel, whereas the boiling heat transfer in narrow annular gap was greatly enhanced compared with the normal channel.

This paper expands the principal of coolant-cushioning in Nuclear Steam Generators whereby the two-phase coolant, especially the bubble film on the tube surface, moderates the vibration of coolant tubes against their supports. This thesis was advanced by the author in JSME ICONE-11 paper #36443 “Fretting in Nuclear Steam Generators – A New Approach”. The current paper addresses tube bundle and AVB geometry issues; examines the tube bundle-coolant-AVB interfaces and examines implications for recirculation flow, AVB design and boiler size. The need for research to confirm empirical findings is reiterated.