Materials Reports  2020, Vol. 34 Issue (Z2): 437-440 |
| METALS AND METAL MATRIX COMPOSITES |
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| Research on General Corrosion of Alloy 690 in Primary and Secondary Loops of Nuclear Plants |
| LIN Zhenxia, DANG Ying, XU Qi, LI Dan
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| Nuclear Power Institute of China, Chengdu 610213, China |
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Abstract Alloy 690 is the most widely used material for steam generator (SG) heat transfer tubes of nuclear power plants. The failure of heat transfer tubes is caused mainly by corrosion. So it’s necessary to study the corrosion resistance of alloy 690 in nuclear power plant. The current experiment on general corrosion of alloy 690 can control the parameters of temperature, pressure and chemical component, but not yet consider the influence of flow. Actually, the primary and secondary water in contact with SG tubes flows at high speed, and the flow possibly has some promoting effect on general corrosion. The experimental method of this paper can control the flow rate for the first time by means of a newly developed set of agitation equipment.The experiments of general corrosion in this paper are conducted in simulated primary and secondary loops of nuclear plants, to acquire the corrosion rate and the metal corrosion product release rate of domestic alloy 690 tubes, which are compared with tubes from Valinox. In the experiments, water chemistry and flow rate are controlled accurately. The results indicate that, the general corrosion resistance of domestic alloy 690 tubes is almost equal to the imported tubes; erosion has some effect on primary general corrosion of alloy 690, but has no obvious effect on secondary general corrosion.
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Published: 08 January 2021
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| About author:: Zhenxia Lin received his M.S. degree in June 2014 from Unclear Power Institute of China (NPIC). He was a research assistant with the research field of reactor structural material. |
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