Experimental Study of Steady-State Boiling of Sodium Flowing in a Single-Pin Annular Channel

Kikuchi, Yoshihiro; Haga, Kazuo; Takahashi, Tadao

doi:10.1080/18811248.1975.9733074

Cited by 11 publications

(4 citation statements)

References 8 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…In addition, the liquid droplets were observed to deposit to the wall from a series of snapshots of X ray images [60]. Another experimental study on sodium boiling in a single-pin annular channel showed a similar flow pattern transitions from bubbly flow to slug flow and then to annular flow [61]. Therefore, it can be expected that the flow patterns inside a heat pipe for liquid metals and for low-temperature fluids will also be similar.…”

Section: High-temperature Working Fluidsmentioning

confidence: 99%

A two-phase three-field modeling framework for heat pipe application in nuclear reactors

Shi

Liu

Yilgor

et al. 2022

Annals of Nuclear Energy

View full text Add to dashboard Cite

Section: High-temperature Working Fluidsmentioning

confidence: 99%

A two-phase three-field modeling framework for heat pipe application in nuclear reactors

Shi

Liu

Yilgor

et al. 2022

Annals of Nuclear Energy

View full text Add to dashboard Cite

“…To investigate the onset and propagation of sodium boiling within pin bundles, various experiments have been carried out by different authors. Some of previous experimental investigations of sodium boiling are Kikuchi et al, [1], Khafizov et al [2], Qiu et al [3], Zicheng et al [4] and Katsuro et al [5]. But since sodium experiments are expensive and involve safety concerns, large volumes of numerical simulations were also reported on sodium boiling.…”

Section: Introductionmentioning

confidence: 99%

CFD Simulation of Sodium Boiling in Heated Pipe using RPI Model

Raj¹,

Velusamy²

2017

World Congress on Momentum, Heat and Mass Transfer

View full text Add to dashboard Cite

CFD prediction of sodium flow boiling in a 2-D asymmetric heated pipe is carried out using RPI wall heat flux partitioning model available in the commercial CFD code FLUENT. The two phase models are validated with published pressurized water and R12 flow boiling experimental data. The influence of radial mesh density on validation parameter is investigated. Void fraction, wall temperature and liquid temperature at different axial section are taken as the validation parameter for water experiment. Among the various parameters predicted, the void fraction of water at the outlet gave the largest deviation of 35% with experimental data. But its value is in close comparison near the location of boiling initiation. Further the void fraction calculated based on heat balance compares well with the CFD results. Radial distributions of void fraction, vapour velocity, bubble diameter at the end of heated section are the parameters compared with R12 experimental data. Except bubble diameter, all other parameters gave reasonably good comparison. But when invoking breakup and coalescence model the predicted value of bubble diameter value approaches experimental data. Finally, sodium boiling simulations were carried out with the chosen models. It was found that, unlike water and R12, in the case of sodium, stable boiling was not achieved and boiling is highly localized to the end of heated section. Periodic growth and re-condensation of sodium vapour is found. Further, temporal oscillation in the heat transfer coefficient up to 8% was observed in the boiling zone. During sodium bubble expansion the evaporative heat flux dominates and once the bubble detaches, the quenching heat flux dominates. Unlike pressurised water evaporative and quenching heat flux have almost equal participation in boiling region.

show abstract

“…The upper boundary of gas volume fraction has been chosen according to work [1,2]. To calculate the interphase friction coefficients, the following relation are used:…”

Section: Description Of the Models To Calculate Processes In Two-phasmentioning

confidence: 99%

“…So it is important to provide numerical and experimental investigations of the motion of bubbles in liquid metals. During the sodium boiling in core or during the leakage in steam generator slug regime is the main two-phase [1,2]. The feature of slug flow regime is the existence of sharp boundary between phases that make some difficulties to numerical simulating especially in one dimensional approximation.…”

Section: Introductionmentioning

confidence: 99%

Numerical simulation of gas volume motion during the gas injection into liquid metal coolant

Usov¹,

Лобанов²,

Прибатурин³

et al. 2017

J. Phys.: Conf. Ser.

View full text Add to dashboard Cite

Abstract. Detailed description of relations and numerical approaches to simulate transport of gas phase in a vertical liquid column is presented in a current paper. These approaches are important to calculate phenomena that take place during steam generator tube rapture in fast reactors with liquid metal coolant. Presented relations determine interphase friction between gas and fluid in different flow regimes of two-phase flow. It is shown that correct definition of interphase friction coefficients determines the correct value of bubble velocity that is very important to simulate two-phase flow in steam generator and reactor core. The paper also contains numerical algorithm to calculate motion of gas volume in fluid flow. Especial attention is paid to describe the algorithms for simulating two-phase flow with sharp edges between phases that are character for slug flow regime. Also some experimental results are presented in the paper. Comparison between experimental data and calculation results has been provided. IntroductionThe active studies of processes in two-phase flow of liquid metals been providing in the middle of the 20th century due to the development of project of fast reactors. Unfortunately, due to the inherent technological problems in operation with liquid metals, many countries refused to further develop this technology, with the exception of Russia, which is still developing projects for liquid metal-cooled reactors. There are projects of lead, lead-bismuth and sodium cooled reactors. Accidents caused by the steam generator tubes rupture and the coolant boiling due to insufficient of heat removal from the core is determined by the behaviour of the gaseous phase in liquid. The motion of bubbles in the core, pulsations of theirs volume strongly influence to the behaviour of the reactor due to the presence of a positive feedback between the core neutron power and the gas volume. Knowledge about the specific of gas bubble motions into liquid metal coolant is also important for metallurgy to calculate of boiling or blown steel-melting baths. There is a lack of information about the processes in two-phase liquid metal flow. So it is important to provide numerical and experimental investigations of the motion of bubbles in liquid metals. During the sodium boiling in core or during the leakage in steam generator slug regime is the main two-phase [1,2]. The feature of slug flow regime is the existence of sharp boundary between phases that make some difficulties to numerical simulating especially in one dimensional approximation.

show abstract

Experimental Study of Steady-State Boiling of Sodium Flowing in a Single-Pin Annular Channel

Cited by 11 publications

References 8 publications

A two-phase three-field modeling framework for heat pipe application in nuclear reactors

A two-phase three-field modeling framework for heat pipe application in nuclear reactors

CFD Simulation of Sodium Boiling in Heated Pipe using RPI Model

Numerical simulation of gas volume motion during the gas injection into liquid metal coolant

Contact Info

Product

Resources

About