CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology

Chengdu, China

CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology

Chengdu, China
SEARCH FILTERS
Time filter
Source Type

Hong G.,Shanghai JiaoTong University | Hong G.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Yan X.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Yang Y.-H.,Shanghai JiaoTong University | Xiao Z.-J.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
International Journal of Thermal Sciences | Year: 2012

Subcooled flow boiling is a commonly applied technique for achieving efficient heat transfer. Although numerous works have been done on two-phase flow phenomena, most of these works focus on land-based channels. Therefore, for identifying the heat transfer phenomena in the barge-mounted system, it is necessary to know the local flow condition in an oscillating acceleration field. In this study, forced convection subcooled water boiling experiments are conducted in narrow rectangular channels at low frequency oscillations. The bubble size, bubble velocity and bubble number density have been statistically analyzed under different heaving conditions and at different flow rates. The results of the bubble size distribution have been presented as cumulative distribution functions, which exhibit in reality a very wide spread of bubble size. The results show that an increase of the oscillation frequency causes an increase fluctuation of bubble size, bubble velocity and bubble number density. Under the same heaving condition, an increase of mass flow rate leads to a decrease fluctuation of bubble size and bubble velocity. A correlation has been sought for the fluctuation of bubble diameter due to heaving motion. The proposed model agrees well with the experimental data within the averaged relative deviation of ±19.2%. © 2011 Elsevier Masson SAS. All rights reserved.


Hong G.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Hong G.,Shanghai JiaoTong University | Yan X.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Yang Y.-H.,Shanghai JiaoTong University | And 2 more authors.
Nuclear Engineering and Design | Year: 2012

A visual study of bubble departure size in forced convective subcooled boiling flow under static and heaving conditions was presented. High-speed digital images of flow boiling phenomena were obtained, which were used to measure bubble departure diameter. Experiments were conducted at atmosphere pressure in a narrow rectangular channel, with mass flux ranging from 300 to 710 kg/m 2 s, heat flux ranging from 65 to 298 kW/m 2 and inlet subcooling ranging from 20 to 40 K. The heaving frequency, which is generated by a six degrees-of-freedom platform, ranged from 0.2 to 0.61 Hz. The results indicated that decreasing mass flux and increasing heat flux had a tendency to increase bubble departure diameter under static condition. In heaving motion, bubble departure size was affected by additional heaving acceleration and flow rate fluctuation. A bubble departure model was proposed to predict the bubble departure diameter under static and heaving conditions by considering the additional acceleration and flow rate fluctuation. The proposed model agreed well with the experimental data within the averaged relative deviation of ±17.5%. © 2012 Elsevier B.V. All rights reserved.


Hong G.,Shanghai JiaoTong University | Hong G.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Yan X.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Yang Y.-H.,Shanghai JiaoTong University | And 2 more authors.
Annals of Nuclear Energy | Year: 2012

An experimental study on onset of nucleate boiling (ONB) in narrow rectangular channel under static and heaving conditions was presented. Flow direction in the channel was vertical upward. The test runs were performed at atmosphere pressure. The mass flux ranged from 298 to 840 kg/m2 s. The heat flux ranged from 33 to 184 kW/m2. The inlet subcooling ranged from 28 to 55 K. The heaving motion was carried out by a six degrees-of-freedom platform. The heaving frequency ranged from 0.2 to 0.61 Hz. Under static conditions, the experimental results indicate that the heat flux and wall superheat, which were needed to initiate the nucleate boiling in narrow rectangular channel, increased with increasing mass flux and inlet subcooling. The classical correlations for conventional channels were not suitable for the present narrow rectangular channel. A new correlation was developed to predict the ONB in narrow rectangular channel under static conditions. The proposed correlation predictions agreed well with the experimental data. Under heaving motion, the fluctuation of mass flux was associated with heaving conditions. The heat flux and the wall superheat for ONB decreased with increasing heaving frequency. © 2011 Elsevier Ltd. All rights reserved.


Xu X.Y.,Xi'an Jiaotong University | Zeng M.,Xi'an Jiaotong University | Zhu H.B.,Xi'an Jiaotong University | Wang Q.W.,Xi'an Jiaotong University | Yan X.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
Progress in Computational Fluid Dynamics | Year: 2013

In this study, the heat transfer of supercritical water in uniformly heated vertical tube is numerically investigated with three different turbulence models, i.e., the renormalisation group (RNG) k-ε model, the shear stress transport (SST) k-ω model and the Reynolds stress model (RSM). In order to find suitable turbulence models for engineering applications, for example, the supercritical water-cooled reactor (SCWR), the results are compared with experimental results by Yamagata et al. (1972) firstly. It has been found that, as the physical properties of supercritical water change rapidly with temperature in pseudo-critical region, these turbulence models, which are simplified, based on isotropic flow and constant physical properties, do not agree well with the experimental results, and even fail in some cases. The numerical results of RSM are also not good enough as expected near critical point. To further explore the flow and heat transfer mechanism in supercritical water, sub-cooled nucleate boiling under subcritical pressure is also numerically investigated in order to find the similarities and differences with those under supercritical pressure. Copyright © 2013 Inderscience Enterprises Ltd.


Liu H.,Chongqing University | Pan L.,Chongqing University | Deng J.,Chongqing University | Yuan D.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Huang Y.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2016

Based on the model of VOF (Volume of Fluid) and UDS (User Defined Scalar), this paper builds the simulation model of using conductivity probe measuring two-phase flow parameters. The process of using double-sensor probe measuring two phase flow was simulated. The electric field distribution was obtained when the probe pierced the bubbles. The results show that the probe piercing the bubbles cause the distinct change of the distribution of current and voltage. As the process of simulation is not influenced by noise signal, bubble shape variation and signal response delay, the ideal signal is obtained such as the square wave signal of current and voltage. The simulation results truly reflect the basic process of the measurement by using the double-sensor probe for gas liquid two-phase flow. © 2016, Yuan Zi Neng Chuban She. All right reserved.


Li J.,North China Electrical Power University | Zhou T.,North China Electrical Power University | Ju Z.,North China Electrical Power University | Huo Q.,North China Electrical Power University | Xiao Z.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
Annals of Nuclear Energy | Year: 2014

Construct the predicting model of CHF based on BP neural network. The sensitivity coefficients of different parameters could be calculated by solving partial differential of the predicting model. With the method of neural network connection weight sensitivity analysis and the data from other researchers' experiments, the sensitivity of different factors to the critical heat flux (CHF) is analyzed. The result shows that, ΔGmax/G0 has the largest sensitivity coefficients to CHF and the inlet temperature has the smallest sensitivity coefficients in the test range. The sensitivity of ΔGmax/G0 could be 20 times of that of the inlet temperature. The BP predictions of CHF fit well with the experimental data, and the errors fall in the margin of 5%. The BP predictions of the influences of ΔGmax/G0 and τ to CFm fit well with Kim's formula, and the largest error is 12.5%. © 2014 Elsevier Ltd. All rights reserved.


Gong H.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Xi Z.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Zan Y.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Huang Y.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
International Congress on Advances in Nuclear Power Plants, ICAPP 2016 | Year: 2016

The second side passive residual heat removal system (PRHRS) is one of innovation designs of Chinese ACP1000 NPP, and it consists of the second side of steam generator, C-type heat exchanger and high accident cooling water tank. This system can remove decay heat passively through the steam generators (SGs) in case of non-LOCA accidents, such as SBO (station black-out) etc. NPIC in China performed steady-state and transient experiments in the ESPRIT facility to validate the capability of prototype PRHRS and to produce data to validate the current thermal-hydraulic safety codes. The experimental results indicated that PRHRS can removal reactor core residual heat successfully in 72 hours during SBO accident. As a candidate analysis tool, RELAP5/MOD3.2 was preliminary evaluated with steadystate and transient experiments. The discrepancies between the calculation and experiment were identified in condensation temperature as well as EST injection flow rate.


Huang J.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Huang Y.-P.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Wang Y.-L.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2013

Based on 170 density wave oscillation experimental data from parallel round tube and narrow rectangular channel, the experiment method, identification method of oscillation and analysis method of experimental data have be uniformed, and the oscillation boundary of round tube and narrow rectangular channel have be analyzed. The investigation results show that the oscillation boundary is not affected by the channel section forms with identical equivalent diameter with pressure 1.0~19.2 MPa, mass flux 101.9~1200.0 kg·m-2·s-1 and inlet sub cooling 18.0~85.2°C.


Niu M.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Huang Z.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Wang J.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Yan X.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2015

According to reasonable simplifying forces exerted on a droplet, the dynamics equation of a droplet at the swirler area in Swirler Pattern Separator is given. Numerical simulation, which is verified by the conclusions from the experiment, is used to discuss the variable percentage of droplets passing throw the swirler area, droplets impacting the swirlers and the barrel wall in accordance with droplets average diameter, air inlet-velocity and swirler angle. The percentage of droplets impacting swirlers appears to increase when droplets average diameter or air inlet-velocity or swirler angle increases. In addition, swirler angle has little effect on the percentage. The more the rate of drag force to inertial force accumulating in droplets moving time, the more difficult for droplets to impact swirlers. © 2015, Yuan Zi Neng Chuban She. All right reserved.


Gong H.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Xi Z.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Zhuo W.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology | Huang Y.,CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology
International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015 | Year: 2015

The second side passive residual heat removal system is one of innovation designs of Chinese ACPI000 NPP, and it consists of the second side of steam generator, C-type heat exchanger and high accident cooling water tank. The purpose of this paper is to evaluate the capability of RELAP5/MOD3.2 code to simulate thermal-hydraulics behavior associated with experiments PRS-ST1 and PRS-TTl. The PRS experiments were simulations of the operation of the second side passive residual heat removal system when station block-out accident happened. The trends of RELAP5/MOD3.2 calculation were very similar to those observed in experiments, and the calculation results were in good agreement with experimental data. The discrepancies between the calculation and experiment were also identified in the water level of accident cooling water tank as well as EST injection flow rate. Much more attention should be paid to the nodalization of ACWT to model three-dimensional natural convection, as well as to modify flow pattern and heat transfer coefficient of direct-contact steam condensation in the future.

Loading CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology collaborators
Loading CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology collaborators