State Nuclear Power Technology Randnter

Beijing, China

State Nuclear Power Technology Randnter

Beijing, China

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Deng C.,Huazhong University of Science and Technology | Chang H.,Tsinghua University | Chang H.,State Nuclear Power Technology Randnter | Qin B.,Tsinghua University | Wu Q.,Oregon State University
Annals of Nuclear Energy | Year: 2016

In the scaled-down test facilities that simulate the accident transient process of the prototype nuclear power plant, the stored energy release in the metal structures has an important influence on the accuracy and effectiveness of the experimental data. Three methods of stored energy analysis are developed, and the mechanism behind stored energy distortion in the test facilities is revealed. Moreover, the application of stored energy analysis is demonstrated for the ACME test facility newly built in China. The results show that the similarity requirements of three methods analyzing the stored energy release decrease gradually. The physical mechanism of stored energy release process can be characterized by the dimensionless numbers including Stanton number, Fourier number and Biot number. Under the premise of satisfying the overall similarity of natural circulation, the stored energy release process in the scale-down test facilities cannot maintain exact similarity. The results of the application of stored energy analysis illustrate that both the transient release process and integral total stored energy of the reactor pressure vessel wall of CAP1400 power plant can be well reproduced in the ACME test facility. © 2016 Elsevier Ltd.


Xiang Y.,Xi'an University of Science and Technology | Wu Y.W.,Xi'an University of Science and Technology | Sun D.C.,Xi'an University of Science and Technology | Sun D.C.,Nuclear Power Institute of China | And 4 more authors.
Progress in Nuclear Energy | Year: 2016

The fourth stage Automatic Depressurization System is an important passive safety feature in Westinghouse AP1000 which enables controlled depressurization of reactor coolant system in small break LOCA. However, large amount of coolant can be carried to the containment via the ADS-4 branch entrainment and the upper plenum entrainment during the depressurization process, which poses great threats to core uncovering and melting. The automatic Depressurization and Entrainment TEst Loop (ADETEL) modeled after AP1000 with a scaling ratio of 1:5.6 was constructed to investigate the entrainment and depressurization behavior after the actuation of ADS-4 valves. The entrainment and depressurization features were investigated under different initial pressure, mixture liquid level in the pressure vessel and heating power. The entrainment deposition effect of the reactor internals was also investigated. The test data reveals that large amount of water are entrained through the ADS-4 branch line within a short period of time. The liquid entrainment rate and the reduced rate of the mixture liquid level in the pressure vessel increase with the initial system pressure. It is notable that the core uncovery was experienced when the initial pressure was set to 0.5 MPa in current experimental conditions. The reactor internals have little effect on the entrained mass and the mixture liquid levels in the pressure vessel. © 2016 Elsevier Ltd. All rights reserved.


Sun D.C.,Xi'an University of Science and Technology | Zhang J.,Xi'an University of Science and Technology | Xiang Y.,Xi'an University of Science and Technology | Tian W.X.,Xi'an University of Science and Technology | And 3 more authors.
Annals of Nuclear Energy | Year: 2015

Tee branches are widely used in nuclear power plants for varying purposes. The tee branch is adopted by the fourth stage Automatic Depressurization System (ADS-4) in Westinghouse AP600/AP1000 to depressurize the primary loop during the small break loss of coolant accident (SBLOCA). However, large amount of coolant will be entrained simultaneously through the ADS-4 branch which poses the threat of core uncovering and melting. Visualization experiments with double-end gas inlets were conducted to investigate the ADS-4 tee branch entrainment phenomena in AP1000. Entrainment process were recorded by high speed camera and analyzed in detail. The onset of liquid entrainment and entrainment rate data were obtained and compared with existing data and correlations, and discrepancies were found in the comparison due to the difference of test section geometric configuration. The gas flow rate has little effect on the branch quality at the same dimensionless gas chamber height in entrainment rate tests. The entrainment frequency was also studied. The test data reveal that the entrainment period decreases rapidly with the increase of entrainment rate in low range of the entrainment rate, and gradually stabilizes in high range of the entrainment rate. © 2014 Elsevier Ltd.


Sun L.,State Nuclear Power Technology Randnter | Han W.,CAS Institute of Engineering Thermophysics | Jin H.,CAS Institute of Engineering Thermophysics
Applied Thermal Engineering | Year: 2015

This study proposes a hybrid refrigeration system activated by mid/low-temperature sensible heat source with ammonia-water (NH3-H2O) binary mixture as working fluid. The heat source is utilized in cascade in the hybrid system. This heat source is used to generate a superheated NH3-H2O mixture vapor, which is used successively in the power generation and compression subsystem and in the absorption and rectification subsystem to produce refrigerant vapor. Energy and exergy analysis results show that the Coefficient Of Performance, System Coefficient Of Performance, and exergy efficiency of the proposed system in the base case are 0.722, 0.485, and 23.1%, respectively. Effects of turbine inlet pressure, turbine outlet pressure, and solution concentration in the power generation subsystem on system performance are examined. This study provides a new refrigeration method to effectively utilize mid/low-temperature sensible heat. © 2015 Elsevier Ltd. All rights reserved.


Sun D.C.,Xi'an University of Science and Technology | Tian W.X.,Xi'an University of Science and Technology | Qiu S.Z.,Xi'an University of Science and Technology | Su G.H.,Xi'an University of Science and Technology | And 3 more authors.
Progress in Nuclear Energy | Year: 2014

ADS-4 is an important passive safety feature in AP1000 design which provides a controlled depressurization. In this paper, reduced diameter and height scaling analysis with identical fluid properties was conducted on AP1000 ADS-4 blowdown and depressurization process. The scaling analysis consisted of ADS-4 branch line entrainment scaling, system depressurization scaling and upper plenum entrainment scaling. Reasonable dimensionless criteria of related thermal hydraulic phenomena were chosen and developed by analyzing conservation equations. Experimental geometric dimensions and operating conditions for the scaled test facility were obtained. © 2014 Elsevier Ltd. All rights reserved.


Zhang H.,North China Electrical Power University | Zhang H.,China Nuclear Power Engineering Co. | Niu F.,North China Electrical Power University | Yu Y.,North China Electrical Power University | And 3 more authors.
Annals of Nuclear Energy | Year: 2015

Mixing and thermal stratification often occur in passive containment cooling systems. Currently, most of reactor system analysis codes do not use the thermal stratification to simplify the calculation. The 2-D or 3-D CFD methods require very fine grids and long running times. In this paper, a new code based on thermal stratification is used to solve heat transfer problems in large enclosures which can give good results in a short time without complex meshing. The models in the code provide the capability to simulate the containment. At the same time, a series of small scale model experiments with air injection are conducted to simulate the LOCA accidents. Simple variable method is adopted in the experiments to study the effects of four different factors on the flow field. The results of the experiments and the codes are compared to verify the validity of the code. Several typical conditions were calculated by a CFD code, and comparison between the running times of these two codes shows the advantage of the model used in the new code. © 2015 Elsevier Ltd. All rights reserved.


Zhang P.,State Nuclear Power Technology Randnter | Chen P.,Circle Technology | Li W.,State Nuclear Power Technology Randnter | Di Z.,State Nuclear Power Technology Randnter | And 3 more authors.
Progress in Nuclear Energy | Year: 2016

Pool entrainment is an important phenomenon in LOCA transient in both conventional and passive PWR plants. Kataoka and Ishii's pool entrainment model is widely used in small break LOCA and reactor transient analysis, but has limited validation against experimental data in high gas flux region. This paper is to report an experimental study of air-water pool entrainment with prototypic gas flux conditions of AP1000. The test section is 2200 mm long and 380 mm in diameter, and built by transparent material for visualization. The range of air flow superficial velocity is 0.98-5.41 m/s. The two phase mixture level is measured by Guided Wave Radar (GWR), and the water entrainment is recorded by weight measure after it is separated from entrainment separator. A correlation of pool entrainment in high gas flux region is proposed based on experimental data obtained in this study. The phenomenon of entrainment saturation is found in high gas flux region. The visualization indicates that pool entrainment in high gas flux region is in transition to jet/fountain flow, which might have strong exit effect. © 2016 Elsevier Ltd. All rights reserved.


Lin Y.,State Nuclear Power Technology Randnter | Liyong H.,State Nuclear Power Technology Randnter
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2013

To maintain the containment within the design limits of pressure and temperature, the advanced pressurized water reactor (APWR) designed by Westinghouse uses a passive safety system to transfer the heat from inner containment to outside. The passive containment cooling system (PCCS) includes many natural phenomena mechanisms. Steam condensation is one of the most important phenomena. Most heat is removed by steam condensation on inside surface of the containment during the postulated design basic accidents (DBA). It is very significant for engineering designing and code developing to study the mechanism of steam condensation on cold surface. There was an experiment made by University of Wisconsin on it. In this paper, the structure pressure of the pressured equipment is calculated and the tightness is also analyzed. Copyright © 2013 by ASME.


Chang L.,State Nuclear Power Technology Randnter | Yang Y.,State Nuclear Power Technology Randnter
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2013

Here we present an unsteady two-dimensional numerical study on drops migrating in a matrix fluid. A computer code is developed to simulate the temperature distribution and velocity field both inside and outside drops. The governing momentum and energy equations are solved by finite-difference method and the deformable interface of drops is captured by fronttracking technique. The moving interface advection in fronttracking method is improved and mass conservation is maintained during the entire course of drop migration. Simulation results show that our improved model is consistent with previous numerical work on this subject. With the numerical model, two-dimensional thermocapillary motion of drops is studied. Copyright © 2013 by ASME.


Sun D.C.,Xi'an University of Science and Technology | Xiang Y.,Xi'an University of Science and Technology | Tian W.X.,Xi'an University of Science and Technology | Liu J.C.,Xi'an University of Science and Technology | And 3 more authors.
Progress in Nuclear Energy | Year: 2015

Upper plenum liquid entrainment can reduce the coolant inventory in the reactor pressure vessel and poses great threats to core uncovering and melting in the accident conditions. To make up for the de ficiency of upper plenum entrainment database, the Automatic Depressurization and Entrainment TEst Loop (ADETEL) was constructed and upper plenum entrainment experiments were conducted. The test matrix was designed with various variables to investigate the upper plenum entrainment comprehensively. Entrainment phenomena and test data were obtained and analyzed. The entrainment deposition effect of reactor internals was also investigated. The test results indicate that the steam flow rate has little on the entrainment rate under stable experimental conditions. The entrainment rate decreases dramatically with the mixture liquid level in the pressure vessel when the liquid level drops below the hot leg elevation. The entrainment rate is promoted after the installation of reactor internals due to the increase of steam velocity in the upper plenum. Huge discrepancy exists between the test data and Kataoka and Ishii's (1984) pool entrainment model in the near surface region suggesting that the upper plenum entrainment is mainly dominated by side branch entrainment mechanism which is different from the pool entrainment mechanism. © 2014 Elsevier Ltd. All rights reserved.

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