Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology

Chengdu, China

Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology

Chengdu, China

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Peng C.,Shanghai JiaoTong University | Liu J.,ShanDong Electrical Power Engineering Consulting Institute Corporation | Yan X.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Cao X.,Shanghai JiaoTong University
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015

Thermal fragmentation process plays a key role in Fuel Coolant Interaction (FCI) during NPP's severe accidents, which significantly affects the heat transfer and determines the ratio of heat transferred to mechanical energy. Although various thermal fragmentation models have been raised, the phenomenon is not well understood due to its complicated process. Unstable film boiling is one of the mechanisms that lead to thermal fragmentation of the melt. In this paper, thermal fragmentation process induced by unstable film boiling condition is discussed based on theoretical analysis, including a momentum equation for vapor film dynamics, an energy equation for each phase involved and some appropriate boundary conditions. The effects of the initial melt temperature, coolant temperature and ambient pressure on thermal fragmentation process are also investigated. In order to evaluate this fragmentation model, a set of experiments on typical simulant materials are introduced, which give the fragmentation time and the evolution of mixture region. The evaluation shows that the main results calculated from the model are consistent with the experimental data and reflect the fragmentation process well. Copyright © 2015 by JSME.


Wang S.,Xi'an University of Science and Technology | Shan J.,Xi'an University of Science and Technology | Zhang B.,Xi'an University of Science and Technology | Lang X.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015 | Year: 2015

Heat transfer is maximized at the CHF (critical heat flux) point and sharply reduced after that. CHF has a significant effect on the integrity, safety or economic efficiency of components and systems. Therefore, it's of vital importance to predict CHF accurately for given conditions. An accurate prediction of CHF requires a good understanding of physical mechanisms, parametric trends, as well as adequate experimental databases. A study on parametric trends of CHF has been carried out for water flow upward in vertical uniform heated round tubes. The study is based on the UO (University of Ottawa) database which is one of the largest tube CHF databases containing 78 data sets compiled from worldwide sources, more than 30,000 data points in all. The data in the range of conditions of practical interest is selected and unreliable data are removed according to the slice method, heat balance requirement and other screening criteria. The parametric trends of normalized CHF are explored with respect to variations in pressure, mass velocity and local quality. In addition, more than 2500 CHF experiment data points of uniform heated 5x5 bundle are picked out from Columbia University bundle CHF databases, which are used in comparison analysis of CHF trends in tube and bundle with respect to pressure, mass velocity and local quality obtained from subchannel code COBRA. It turns out that the two are largely consistent in terms of parametric trends. The research lay a solid foundation of better understanding, accurately predicting and enhancing CHF in tubes and bundle.


Huang Z.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Xiao Z.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Yan X.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Zan Y.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | And 2 more authors.
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015

In this work, theory analysis about K-H instability of film interface in swirling flow field in cyclone separator was performed. The momentum equations and continuity equations of each phase fluid were linearized by substitution of potential function firstly. Then the dynamic boundary condition and kinematic boundary condition were obtained based on stress analysis of film interface. According to the linearized equation and boundary conditions, the dispersion relation was established. Additionally, the motion law of film was obtained based on principle of stress equilibrium. Then the criterion for interface K-H instability was obtained. A computer program was developed according to theory model. Then simulation and analysis was performed on interface K-H instability under different conditions. It was found that centrifugal force of film could constrain interface K-H instability of film but the centrifugal force of steam could cause K-H instability. Additionally, film interface tended to unstable state with the increase in film thickness. It was also found that the increase in steam velocity could constrain interface instability when rise angle of swirl-vane of separator was less than a certain value. But the increase in steam velocity could cause instability when rise angle exceeded a certain value. Copyright © 2015 by JSME.


Bao W.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Chen B.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Xu J.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Xie T.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | And 2 more authors.
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015

Subcooled flow boiling in inclined tube occurs widely in many industries and applications. An experimental study on the local interfacial characteristics of subcooled flow boiling was carried out under vertical and inclined conditions. The test section is a circular tube (i.d. 24mm) of which axial heated length is 1m. A double-sensor optical probe was used to investigate the radial distribution of interfacial parameters including local void fraction, bubble frequency, interfacial velocity, Sauter bubble diameter and Interfacial Area Concentration (IAC). The range of heat flux and mass flux are 140-900kW/m2 and 200-800kg/m2s, respectively. The inclination angles are 5°, 10°, 20°, 30°, and the polar angle between measurement direction and inclination direction in the probe measurement cross-section are 0°,45°, 90°. From the test, the local interfacial parameters were measured at 15 radial locations at probe elevation. Based on these data obtained in the previous test loop, the influence of flow condition and inclination angle on the profiles of local interfacial parameters was discussed. Copyright © 2015 by JSME.


Chen B.,North China Electrical Power University | Zhou T.,North China Electrical Power University | Li J.,North China Electrical Power University | Song M.,North China Electrical Power University | Huang Y.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015

By the method of numerical simulation and the geometric modeling based on existing experiment table, the heat transfer characteristics of water, sodium and lead-bismuth in a 2mm vertical rectangular channel were studied. The effect of different boundary conditions on the thermal characteristics was also studied. The effect of inlet velocity and heat flux density on the heat transfer coefficient and the pressure drop of different fluids in the rectangular channel were studied. It provides a reference for selection of working fluid in the narrow rectangular channel. Copyright © 2015 by JSME.


Song M.,North China Electrical Power University | Zhou T.,North China Electrical Power University | Chen B.,North China Electrical Power University | Fang X.,North China Electrical Power University | And 2 more authors.
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015

Based on a 2mm vertical rectangular channel which was designed by North China Electric Power University, experiments were conducted on the experiment table to research heat transfer deterioration in natural circulation. The result is that, heat transfer deterioration can happen in natural circulation, at which heating section temperature increase rapidly and heat transfer coefficient decrease rapidly. The increase of subcooling degree and mass velocity contribute to delay heat transfer deterioration. Because of low fluid velocity in natural circulation and secondary flow influence in narrow rectangular channel, liquid film can be evaporated easily and heat transfer deterioration happens. Futuremore, liquid film can be squeezed by narrow rectangular channel, so the critical heat flux in narrow rectangular channel is lower than that in others. Copyright © 2015 by JSME.


Liu L.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Xiao Z.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Yan X.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Zeng X.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Huang Y.,Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
Annals of Nuclear Energy | Year: 2013

As there is no phase change, the risk of boiling crisis can be excluded in Super Critical Water-cooled Reactor (SCWR). The heat transfer deterioration (HTD) phenomena, which result in an abnormal increase of wall temperature, however, may occur under some conditions. Therefore, it is important to understand the mechanism of HTD for keeping the temperature integrity of the fuel cladding. In this study, mechanisms of the influence of turbulence structure and fluid properties on HTD phenomena in annular channels at low mass flux and high mass flux were both investigated. The results showed that HTD at low mass flux is mainly caused by buoyancy effect and the variations of fluid properties, while at high mass flux HTD is due to the acceleration effect and the variations of fluid properties. Heat flux and the operating pressure also play important parts in HTD both at low mass flux and high mass flux. © 2012 Elsevier Ltd. All rights reserved.

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