Flow instability analysis of supercritical water-cooled reactor CSR1000 based on frequency domain

2012 ◽  
Vol 49 ◽  
pp. 70-80 ◽  
Author(s):  
Xiaoyan Tian ◽  
Wenxi Tian ◽  
Dahuan Zhu ◽  
Suizheng Qiu ◽  
Guanghui Su ◽  
...  
2019 ◽  
Vol 63 (2) ◽  
pp. 328-332 ◽  
Author(s):  
Ákos Horváth ◽  
Attila R. Imre ◽  
György Jákli

The Supercritical Water Cooled Reactor (SCWR) is one of the Generation IV reactor types, which has improved safety and economics, compared to the present fleet of pressurized water reactors. For nuclear applications, most of the traditional materials used for power plants are not applicable, therefore new types of materials have to be developed. For this purpose corrosion tests were designed and performed in a supercritical pressure autoclave in order to get data for the design of an in-pile high temperature and high-pressure corrosion loop. Here, we are presenting some results, related to corrosion resistance of some potential structural and fuel cladding materials.


2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Wang Lianjie ◽  
Lu Di ◽  
Zhao Wenbo

Transient performance of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWel (CSR1000) core during some typical transients, such as control rod (CR) ejection and uncontrolled CR withdrawal, is analyzed and evaluated with the coupled three-dimensional neutronics and thermal-hydraulics SCWR transient analysis code. The 3D transient analysis shows that the maximum cladding surface temperature (MCST) retains lower than safety criteria 1260 °C during the process of CR ejection accident, and the MCST retains lower than safety criteria 850 °C during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.


2018 ◽  
Vol 913 ◽  
pp. 237-246 ◽  
Author(s):  
Yan Xia Yu ◽  
Li Ping Guo ◽  
Zheng Yu Shen ◽  
Yun Xiang Long ◽  
Zhong Cheng Zheng ◽  
...  

The average size and density evolution of dislocation loops in AL-6XN austenitic stainless steel, a candidate fuel cladding material for supercritical water-cooled reactor, under proton irradiation were simulated through a rate theory model. The simulation results exhibit relatively good agreement with the experimental results at 563 K. The size and density of defect clusters are calculated under irradiation temperature between 550 K and 900 K and irradiation doses up to 15 dpa which satisfies the working condition in supercritical water-cooled reactor. The fast nucleation between self-interstitials happens at the initial stage of irradiation. The average size of dislocation loops increases while the average density of these loops reduces with the increasing temperature, and the average density approaches to a constant when irradiated at higher irradiation doses. The mechanism is discussed based on the variation of rate constants of defect reactions and the variation of the diffusion coefficients of interstitials and dislocation loops with dose and temperature.


2014 ◽  
Vol 64 ◽  
pp. 295-300 ◽  
Author(s):  
Jianli Hao ◽  
Wenzhen Chen ◽  
Xi Chu ◽  
Shaoming Wang

2015 ◽  
Vol 44 ◽  
pp. 69-76 ◽  
Author(s):  
Francisco Gómez ◽  
José Miguel Pérez ◽  
Hugh M. Blackburn ◽  
Vassilis Theofilis

2010 ◽  
Vol 169 (2) ◽  
pp. 126-133 ◽  
Author(s):  
Suxia Hou ◽  
Fuyu Zhao ◽  
Yun Tai ◽  
Liu Cheng

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