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Zhao Y.,State Nuclear Power Technology R and nter Beijing | Zhang M.,State Nuclear Power Technology R and nter Beijing | Hou F.,Tsinghua University | Gao T.,State Nuclear Power Technology R and nter Beijing | Chen P.,State Power Investment Group Corporation
Nuclear Engineering and Design | Year: 2016

In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40-60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to effectively remove big vapor mushrooms on the heating surface. © 2016 Elsevier B.V. All rights reserved.


Deng C.-C.,Tsinghua University | Chang H.-J.,Tsinghua University | Chang H.-J.,State Nuclear Power Technology R and nter Beijing | Qin B.-K.,Tsinghua University | And 2 more authors.
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2012

During small break loss of coolant accident (SBLOCA) of AP1000 nuclear plant, the behavior of pressurizer surge line has an important effect on the operation of ADS valves and the initial injection of IRWST, which may happen at a time when the reactor core reaches its minimum inventory. Therefore, scaling analysis of the PRZ surge line in nuclear plant integral test facilities is important. Four scaling criteria of surge line are developed, which respectively focus on two-phase flow pattern transitions, counter-current flow limitation (CCFL) behavior, periodic draining and filling and maintaining system inventory. The relationship between the four scaling criteria is discussed and comparative analysis of various scaling results is performed for different length scale ratios of test facilities. The results show that CCFL phenomenon and periodic draining and filling behavior are relatively more important processes and the surge line diameter ratios obtained by the two processes' scaling criteria are close to each other. Thus, an optimal scaling analysis considering both CCFL phenomenon and periodic draining and filling process of PRZ surge line is given, which is used to provide a practical reference to choose appropriate scale of the surge line for the ACME (Advanced Core-cooling Mechanism Experiment) test facility now being built in China. Copyright © 2012 by ASME.


Zhao G.,Dalian University of Technology | Yu J.,Dalian University of Technology | Tian F.,Dalian University of Technology | Liu Y.,Dalian University of Technology | And 2 more authors.
Proceedings of 2011 International Conference on Electronic and Mechanical Engineering and Information Technology, EMEIT 2011 | Year: 2011

The Radial fluid exciting force model and its dynamic characteristics is derived through the flow field simulation of AP-1000 nuclear main pump. The result indicates that complicated harmonics frequency vibrations is existed under the operating of nuclear main pump, and the coupling resonance of blades and vanes should be taken into serious consideration under the designing of nuclear main pump. © 2011 IEEE.


Yang P.-Y.,State Nuclear Power Technology R and nter Beijing | Wang X.-W.,State Nuclear Power Technology R and nter Beijing | Zhang J.-L.,State Nuclear Power Technology R and nter Beijing | Ji S.,State Nuclear Power Technology R and nter Beijing
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2013

Diluting the core melt with oxide sacrificial material (OSM) during in-vessel retention (IVR) is a newly proposed severe-accident-management strategy of advanced LWR. When severe accident occurs, the OSM is melted by the relocated core melt, resulting in the formation of a ternary liquid mixture of core melt and OSM in the corium. To select OSM and evaluate the heat flux on the vessel outer surface, and to assess the feasibility of the dilution design scheme, the thermophysical properties of the formed multi-component mixture should be obtained first. In this paper, the thermophysical properties of Fe3O4, TiO2 and Al2O3 were compared. Density, specific heat, thermal conductivity and viscosity of the molten ternary mixture UO2-ZrO2-OSM were also calculated. The results show that to ensure the inverse stratification to occur, implying that the oxide layer locates on top of the metallic layer, the required minimum mass of Fe3O4 should be larger than that of TiO2 and Al2O3. The specific heat and thermal conductivity of the ternary mixture increase with OSM mass, while the viscosity decreases as the mass of OSM increases. Moreover at a given temperature, the molten mixture with a lower melting point also has a smaller viscosity.

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