Chapter

Energy Materials 2017

Part of the series The Minerals, Metals & Materials Series pp 329-341

Effect of Steam Pressure on the Oxidation Behaviour of Alloy 625

  • Shengli JiangAffiliated withCAS Key Laboratory of Nuclear Materials and Safety Assessment, Institute of Metal Research, Chinese Academy of SciencesDepartment of Mechanical and Aerospace Engineering, Carleton University Email author 
  • , Xiao HuangAffiliated withDepartment of Mechanical and Aerospace Engineering, Carleton University
  • , Wenjing LiAffiliated withCanadian Nuclear Laboratories
  • , Pei LiuAffiliated withCANMET

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Abstract

The Canadian Supercritical Water-cooled Reactors (SCWR), among the Generation IV (Gen IV) reactors concepts, are currently being developed in Canada and many other countries. The preliminary design of the Canadian SCWR uses a coolant operating under a pressure of 25 MPa at 625 °C, reaching a peak cladding temperature as high as 800 °C. This presents challenges in materials selections due to limited data on material performance at such high temperatures and pressure. Ni-based alloys have been of particular interest for use in the Gen IV SCWRs, due to their ability to maintain high strength and toughness at elevated temperatures. In this work, corrosion resistance of nickel-based Alloy 625, SS 310 and SS 304 was assessed at 625°C for 1000 h after being exposed to supercritical water (SCW), subcritical water (Sub-CW), and superheated steam; i.e., under pressures of 29, 8 and 0.1 MPa, respectively. The samples showed very small amount of weight gains after the exposure at 29 and 0.1 MPa, and a slight weight loss at 8 MPa due to pitting formation. The surface morphology and cross-section microstructure were analyzed using a Scanning Electron Microscope (SEM). The Energy Dispersive X-Ray Spectrometry (EDS) examination of the compositions of the surface oxide, indicated similar oxide formation on the top surface after exposures at different pressures, likely NiO or/and Ni(Cr,Al)2O4 type spinel. The implications of these results are discussed.

Keywords

Alloy 625 Supercritical water-cooled reactor (SCWR) Corrosion mechanism