Experimental Study on Liquid Film Dryout Under Oscillatory Flow Conditions

Author(s):  
Yosuke Yamagoe ◽  
Taisuke Goto ◽  
Tomio Okawa

The use of high power density core is one of the promising ways to improve economic efficiency of advanced boiling water reactors. It is however known that in boiling two-phase flows, an increase in power density commonly reduces the margin to the onset of unanticipated flow instability. Hence, in the development of a boiling water reactor of high power density core, ability to predict the occurrence of boiling transition is considered indispensable even when the coolant flow rate is not in the steady state. In the present work, sinusoidal oscillation was applied to the inlet mass flux and the experimental measurement of the critical heat flux was carried out under flow oscillation conditions. It was shown that the critical heat flux decreases monotonically with increased values of oscillation amplitude and oscillation period. These results are consistent with experimental data reported by previous investigators. A simple theory was then proposed to estimate the critical heat flux in oscillatory flow condition. Considering the application to the advanced boiling water reactors, the triggering mechanism of the critical heat flux condition is supposed to be the liquid film dryout in annular two-phase flow regime of high vapor quality. Under the flow oscillation condition, it is expected that long waves are formed on a liquid film due to the time variation of inlet mass flux. Assuming that the wave evolution within a boiling channel is influential in the occurrence of the local dryout of a liquid film, an available nonlinear wave theory was applied to the estimation of critical heat flux under the flow oscillation condition. It was demonstrated that the critical heat fluxes measured under the oscillatory conditions agree with the proposed theory fairly well.

Kerntechnik ◽  
2017 ◽  
Vol 82 (2) ◽  
pp. 148-160
Author(s):  
T. Kaliatka ◽  
A. Kaliatka ◽  
E. Uspuras ◽  
M. Vaisnoras ◽  
H. Mochizuki ◽  
...  

2009 ◽  
Vol 10 (3) ◽  
pp. 398-417 ◽  
Author(s):  
Bo Jiao ◽  
Li-min Qiu ◽  
Jun-liang Lu ◽  
Zhi-hua Gan

Author(s):  
Chang-Lung Hsieh ◽  
Guan-Yu Chen ◽  
Chunkuan Shih ◽  
Jong-Rong Wang ◽  
Hao-Tzu Lin

Under certain conditions, boiling water reactors (BWRs) would be susceptible to couple neutron-thermalhydraulic instability. It is important to predict such potential problems as early as possible and prevent the core instability from happening. In each BWR reload core design, fuel vendors are required to provide instability boundaries on power/flow map to assure safety operation of the nuclear reactor. In Taiwan, a LAPUR5.2 methodology had been adapted to build up the remarkable analysis mode for different types BWRs to verify vendor’s results. However, with upgrading nuclear safety technology, most of boiling water reactors has been adopting partial length fuel assemblies to reduce two-phase pressure drop and void fraction, to improve reactor stability. The question is that LAPUR5.2 methodology cannot precisely analysis stability characteristics from the variation of flow area in fuel assemblies. From the reasons of upgrading stability analysis, a LAPUR6.0 methodology had built to do the related researches. This research was based on a comparison study between LAPUR5.2 and LAPUR6.0 to realize the major differences and their effects on stability characteristics. According to the comparison results for Kuosheng Nuclear Power Plant Unit 2 Cycle 21 reload design, it shows that LAPUR6.0 can completely present pressure drop, void fraction and density reactivity coefficient from the changing of flow area and fuel spacers.


2011 ◽  
Vol 2011 (0) ◽  
pp. 371-372
Author(s):  
Ayaka Fujiwara ◽  
Takuya Suzuki ◽  
Takeyuki Ami ◽  
Hisashi Umekawa ◽  
Mamoru Ozawa

1964 ◽  
Vol 86 (1) ◽  
pp. 23-33 ◽  
Author(s):  
F. E. Tippets

Based on the two-phase flow patterns shown in high-speed motion pictures of the process, a general working equation is derived which relates the critical heat flux for high-pressure bulk boiling water in forced convection to the significant local flow parameters and fluid properties. The equation is applied to a representative selection of several hundred data points from the major available sources for the purpose of investigating trends in the data and to test the validity of the equation.


2021 ◽  
Vol 7 (26) ◽  
pp. eabg4537
Author(s):  
Jiaqi Li ◽  
Daniel Kang ◽  
Kazi Fazle Rabbi ◽  
Wuchen Fu ◽  
Xiao Yan ◽  
...  

Enhancing critical heat flux (CHF) during boiling with structured surfaces has received much attention because of its important implications for two-phase flow. The role of surface structures on bubble evolution and CHF enhancement remains unclear because of the lack of direct visualization of the liquid- and solid-vapor interfaces. Here, we use high-magnification in-liquid endoscopy to directly probe bubble behavior during boiling. We report the previously unidentified coexistence of two distinct three-phase contact lines underneath growing bubbles on structured surfaces, resulting in retention of a thin liquid film within the structures between the two contact lines due to their disparate advancing velocities. This finding sheds light on a previously unidentified mechanism governing bubble evolution on structured surfaces, which has notable implications for a variety of real systems using bubble formation, such as thermal management, microfluidics, and electrochemical reactors.


CORROSION ◽  
1967 ◽  
Vol 23 (3) ◽  
pp. 57-64 ◽  
Author(s):  
L. E. LESURF ◽  
P. E. C. BRYANT ◽  
M. G. TANNER

Abstract Radiolysis of the coolant in nuclear reactors cooled by boiling water results in oxygen in the steam and recirculated water phases. This has dictated the use of stainless steels as the major circuit materials for these reactors. It is shown that ammonia additions to the coolant eliminate oxygen production, permitting the use of mild steel for circuit construction with consequent savings in capital cost Corrosion data are presented for various out-reactor materials (carbon steel, low alloy steels, stainless steels, Monel alloy 400, Inconel alloy 600) exposed out-of-flux to the coolant of two phase in-reactor loops when operated neutral and with ammonia addition. Activation of corrosion products from different origins is discussed. Intergranular attack is related to the presence of oxygen or oxidizing radiolytic species in the water.


Author(s):  
R.W. Morse ◽  
T.A. Moreira ◽  
J. Chan ◽  
K.M. Dressler ◽  
G. Ribatski ◽  
...  

Author(s):  
Tomio Okawa ◽  
Saeyun Kim ◽  
Isao Kataoka ◽  
Masanori Naitoh

Incorporating the recently developed correlations for the rates of droplet deposition and entrainment with the film flow model, critical heat flux due to liquid film dryout in steam-water annular flow was predicted. In the present calculations, the predicted critical heat flux is affected not only by the deposition and entrainment rates but also by the entrainment fraction at the transition to annular flow (Transition quality was estimated by Wallis’s correlation). For this reason, the entrainment fraction at the transition was correlated in terms of dimensionless transition length to annular flow and density ratio. A simple correlation for the occurrence of dryout was also proposed. It was demonstrated by numerical simulations that the present method is to predict the available 1,340 data of dryout heat flux with the mean square relative error of 0.068.


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