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Wang T.,Nuclear Power Institute of China | Wang T.,National Energy Pressurized Water Reactor Technology Research Center | Wang J.,Nuclear Power Institute of China | Wang X.-J.,Nuclear Power Institute of China
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2012

Two-phase boiling experiments have been performed in narrow rectangular channel at middle-low pressure. Based on the measured pressure drop and temperature across the test section, the effect of pressure and mass flow rate has been analyzed. Saturated boiling heat transfer was studied. And a new correlation has been developed for saturated boiling heat transfer under middle-low pressure condition. Additionally, based on the predigested one-dimensioned model of two-phase boiling flow, a new correlation has been developed for two-phase flow pressure drop under middle-low pressure condition. Furthermore, F parameter has been defined to expresses proportion of moistening perimeter between steam-phase and wall in this new correlation. The calculation results of new correlation predict well the experimental results. The average saturated boiling heat transfer coefficient is great impacted by system pressure, mass flow and heat flux. Source


Wang Y.,Nuclear Power Institute of China | Wang Y.,National Energy Pressurized Water Reactor Technology Research Center
Proceedings of the 2012 24th Chinese Control and Decision Conference, CCDC 2012 | Year: 2012

Here, the robustness of the nuclear reactor control system is discussed by its computer simulation results which are obtained respectively from the analysis without the external control system and the analysis with the external control system. After this, the simulation tests both with the classical PID control and with the advanced PID control are also given. And then, the comparative analysis on the robustness of the reactor control system is shown. Finally, the useful approach improving the robustness of the nuclear reactor control system is provided, too. © 2012 IEEE. Source


Jiang N.-B.,Nuclear Power Institute of China | Zang F.-G.,National Energy Pressurized Water Reactor Technology Research Center | Zhang Y.-X.,National Energy Pressurized Water Reactor Technology Research Center | Guan H.,National Energy Pressurized Water Reactor Technology Research Center
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2013

Compared with the void fraction models of two-phase flow in tubes, the research on void fraction models for vertically upward two-phase cross-flow in horizontal tube bundle are not enough. With experimental data, the existing void fraction models of vertically upward two-phase cross-flow in horizontal tube bundle are compared. Moreover, two existing void fraction models are modified. The modified void fraction models are validated by some other experimental data. The result shows the modified void fraction models give better prediction than the original models. Source


Liu Y.,Nuclear Power Institute of China | Liu Y.,National Energy Pressurized Water Reactor Technology Research Center | Li F.,Nuclear Power Institute of China | Li F.,National Energy Pressurized Water Reactor Technology Research Center | And 4 more authors.
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2012

On the basis of RELAP5, COBRA-IV and NLSANMT codes, the three dimensional neutronics and thermal-hydraulics coupled system RECON was developed by means of the parallel processing model and parallel virtual machine technology. The ways of coupling were flexible since different combinations of codes can be selected for various coupled analyses. Validations of the coupled system were completed with a series of benchmarks. Specially in the MSLB benchmark, the comparison of results with other coupled codes demonstrates that RECON has good accuracy and is suitable for the analysis of reactivity insertion accident. Source


Wang X.,Nuclear Power Institute of China | Wang X.,National Energy Pressurized Water Reactor Technology Research Center | Zhang Z.,Nuclear Power Institute of China | Zhang Z.,National Energy Pressurized Water Reactor Technology Research Center | And 6 more authors.
ASME 2011 Small Modular Reactors Symposium, SMR 2011 | Year: 2011

In the concept of BWR-PB with a relatively low power level, the core region is filled with a large number of coated particles, which are directly cooled by boiling light water. It's significant to understand the two-phase flow characteristics in such a complicated pebble-bed structure. A visualization experiment was carried out to investigate the flow phenomena of steam-water two-phase flow in a pseudo-three-dimensional pebble bed using a high-speed video camera. The pebble bed in the experiment was constructed of hundreds of glass beads and a specially designed stainless heating plate, which was used to simulate the heat-generation of solid particles. Based on our observation, five typical flow regimes were identified to distinguish different phase distribution characteristics: bubbly flow, bubbly-slug flow, slug flow, slug-annular flow and pure annular flow. System pressure, mass flow rate and inlet subcooling were considered as the key influence factors for flow regime transition in the experiment. A flow pattern map for low pressure and low inlet subcooling condition was obtained from the experimental data. Copyright © 2011 by ASME. Source

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