ICONE15-10822 DEVELOPMENT OF TECHNOLOGIES ON INNOVATIVE-SIMPLIFIED NUCLEAR POWER PLANT USING HIGH-EFFICIENCY STEAM INJECTORS : (13)STUDY ON HEAT TRANSFER OF DIRECT CONDENSATION OF STEAM ON SUBCOOLED WATER JET

Steam injector (SI) is a passive jet pump which is driven by high-performance steam condensation onto water jet and it is expected to be active at severe accident of nuclear power plant with no electricity. SI is mainly consists of convergent-divergent nozzle. Supersonic steam flow condenses onto water jet in the mixing nozzle and mass, momentum, and energy of steam is transferred to water in the mixing nozzle. Condensed water jet is accelerated at the throat and kinetic energy is converted into pressure in the diffuser, which produces higher pressure than inlet steam pressure. It is easy to apply the SI to nuclear power plant since SI has quite simple and compact structures. The objectives of the present study are to clarify the mechanism of heat and momentum transfer in the mixing nozzle and to determine operating range of SI for practical use. A transparent test section is adopted to conduct visualization of the flow structure with a high-speed video camera as well as measurement of pressure distribution in mixing nozzle, throat, and diffuser with changing back pressure. Fundamental parameters change between operative and inoperative state of the injector were evaluated by measuring pressure and temperature distribution along axial direction of the test section. Discharge pressure as one of operating characteristics of the injector was also measured in changing back pressure by decreasing the opening ratio of the back pressure valve attached downstream of the test section. It was confirmed that discharge pressure increased and the injector became inoperative unsteadily with decreasing opening ratio of the back pressure valve just after it produced the maximum discharge pressure. In the present investigation, this maximum discharge pressure is evaluated as the operation limit of the injector. Furthermore, discharge pressure from diffuser, which is one of the indicators of operating performance as well as operating limit is predicted from inlet condition adopting one-dimensional analysis model proposed previously. By comparing analytical result with experimental data, as well as visualization of flow structure in throat and diffuser, physics model including two-phase flow structure with shock wave which was observed at throat and diffuser are discussed in order to predict injector’s operation with high accuracy.

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The Simulation Research of Noncondensable Gas to Condensation in Secondary Side Condenser of Floating Nuclear Power Plant Based on RELAP5

Volume 6B: Thermal-Hydraulics and Safety Analyses ◽

10.1115/icone26-82222 ◽

2018 ◽

Author(s):

Si-wei Yan ◽

Chun-mei Li ◽

Tie-bo Liang ◽

Jing Zhao ◽

Cheng-ming Hao ◽

...

Keyword(s):

Heat Transfer ◽

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Cooling Water ◽

Gas Content ◽

Simulation Research ◽

Cooling Water Temperature ◽

Noncondensable Gas ◽

Secondary Side

Similar to conventional nuclear power plant, condensate water subcooling is a common problem in secondary coolant of floating nuclear power plant, which is caused by many reasons. In this article, RELAP5 is used to simulate the phenomenon of condensate water subcooling caused by noncondensable gas. The influence of noncondensable gas to condenser pressure, subcooling temperature, heat transfer rate, terminal temperature difference, cooling water temperature rise is presented. The results obtained through this study have shown that the model with non-condensable gas in steam can simulate condensate water subcooling, and reveal the discipline of condenser heat transfer characteristics as a function of noncondensable gas content.

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Evaluation of the In-Vessel Heat Transfer Characteristics for the Debris Removal of the Fukushima Daiichi Nuclear Power Plant

The Proceedings of the National Symposium on Power and Energy Systems ◽

10.1299/jsmepes.2016.21.b215 ◽

2016 ◽

Vol 2016.21 (0) ◽

pp. B215 ◽

Cited By ~ 1

Author(s):

Hikaru Mitsuda ◽

Kai Yamada ◽

Shuichiro Miwa ◽

Hiroto Sakashita ◽

Michitsugu Mori

Keyword(s):

Heat Transfer ◽

Power Plant ◽

Nuclear Power Plant ◽

Nuclear Power ◽

Heat Transfer Characteristics ◽

Transfer Characteristics ◽

Debris Removal ◽

Fukushima Daiichi

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Hydraulic Characteristics Research on SG Under Tube Plugging Operations Using FLUENT

Volume 9: Student Paper Competition ◽

10.1115/icone26-81641 ◽

2018 ◽

Author(s):

Xiaohan Zhao ◽

Mingjun Wang ◽

Wenxi Tian ◽

G. H. Su ◽

Suizheng Qiu

Keyword(s):

Heat Transfer ◽

Power Plant ◽

Nuclear Power Plant ◽

Pressure Drop ◽

Nuclear Power ◽

Reactor System ◽

Economic Losses ◽

Maximum Pressure ◽

Hydraulic Characteristics ◽

Tube Plugging

Steam Generator (SG) is a critical equipment in the nuclear power plant, it is the huge heat exchanger in reactor system which can achieve removing fission energy from the reactor system effectively to ensure safety of the whole nuclear system. It is located between the primary and the secondary loop in reactor system act as the intermediate hub of energy and the security barrier in nuclear power plant. Generally, there are numerous of U-shaped heat transfer tubes in SG it is one of the weakest structures throughout the primary loop system. So the integrity of the SG especially its heat transfer tubes is important to the safety of reactor operation. The degradation problem of heat transfer tubes together with ruptures accidents often occur under suffer environments in reactors, which include thermal stress, mechanical stress and so on, it is noteworthy that this kind of accidents is inevitable due to the limited properties of existing materials. The performance of the SG is seriously affected by the number of failure tubes. Plugging operations through various mechanical means is the most common method to solve the tubes ruptures problems which can reduce the economic losses to the utmost extent. However, plugging operations will make huge impact on the thermal hydraulic performances of both sides of SG. It’s meaningful to research the characteristics of the plugging affects under different operations. In this paper the hydraulic characteristics of primary side in AP1000 SG under a certain fraction of heat transfer tube plugging conditions is researched. Three dimensional hydraulic characteristics of primary side coolant in SG under different plugging conditions are obtained by using the thermal hydraulic software FLUENT. The typical plugging fraction in this simulation model is 10 percent, and the effect of plugging locations also be considered through changing the plugging positions using the zone marking method. The results shows that the pressure drop under the structure integrated SG is 358.01MPa which is accordance with the results from Westinghouse 343KPa. The pressure drop values varies when changing positions of the plugging tubes under the same plugging fraction condition. The flow fields in bottom head also change meanwhile and the maximum pressure drop can reach up to 388.05KPa when the plugging fraction is 10%. The growth rate become significant when tube plugging fraction larger than 5%, and differences between maximum and minimum values of total pressure drop under different plugging positions become larger gradually. Finally the local resistance coefficients and flow field distributions of primary side in SG under various plugging conditions are obtained which is meaningful for the reactor safety and it can be a good reference for the maintenance of SG.

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