Characterization of TiO 2 -Doped Yttria-Stabilized Zirconia (YSZ) for Supercritical Water-Cooled Reactor Insulator Appli
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JTTEE5 22:734–743 DOI: 10.1007/s11666-013-9911-1 1059-9630/$19.00 ASM International
Characterization of TiO2-Doped Yttria-Stabilized Zirconia (YSZ) for Supercritical Water-Cooled Reactor Insulator Application F. Barrett, X. Huang, and Dave Guzonas (Submitted May 25, 2012; in revised form January 25, 2013) In this study, TiO2-doped YSZ samples were tested in supercritical water (SCW) to evaluate their corrosion behavior. The doped samples were produced by mechanically alloying standard 7 wt.% Y2O3ZrO2 with 5, 10, and 15 wt.% of TiO2 first. The bulk sample pieces were then obtained using plasma spraying of the alloyed powder materials followed by sintering. The results showed that the weight changes for 5TiYSZ and 10TiYSZ after 1000 h of exposure in SCW were negligible and the sample surfaces did not exhibit any indication of corrosion. In comparison to the reference materials (Al2O3 and 7YSZ) processed using the same method, the rate of weight change followed the order of Al2O3 > 7YSZ, 15TiYSZ > 10TiYSZ > 5TiYSZ. As several TiO2-doped 7SYZ compositions also display increased fracture toughness and reduced thermal conductivity, they may be considered as potential candidates for thermal insulation in a SCW-cooled nuclear reactor.
Keywords
ceramic materials, corrosion, material for supercritical water-cooled nuclear reactor (SCWR), plasma spraying
1. Introduction Supercritical water (Tc = 374 C and Pc = 22.1 MPa) is becoming increasingly important in recent years as a fluid for energy generation and the disposal of bio-hazardous materials. It can convert biomass into combustible gasses in a supercritical water gasification process (SCWG) (Ref 1) and is also capable of oxidizing various types of toxic bio-waste in a supercritical water oxidation process (SCWO) (Ref 2). Additionally, to increase the thermodynamic efficiency of conventional nuclear-powered steam turbines, the use of supercritical water as working fluid has been proposed in a new generation of supercritical water-cooled nuclear reactors (SCWR) (Ref 3). Canada is a member of an international consortium in the development of next generation (Gen IV) nuclear power reactors. Among the six reactor technologies selected for research and development under the Gen IV program, Canada has chosen to focus on the SCWR design. The proposed reactor outlet temperature for the F. Barrett and X. Huang, Department of Mechanical and Aerospace Engineering, Carleton University, Ottawa, ON Canada; and Dave Guzonas, Atomic Energy Canada Laboratory, Chalk River, ONCanada. Contact e-mail: xhuang@ mae.carleton.ca.
734—Volume 22(5) June 2013
Canadian SCWR is estimated to be at or over 625 C (Ref 4) under an operating pressure of 25 MPa. The use of high temperature supercritical water as working fluid allows a high thermodynamic efficiency that is close to 50%, compared to about 35% for an advanced Light Water Reactor (LWR). Due to the elevated temperature of SCWR coolant in the reactor core, an insulation material must be placed between the pressure tube and liner as shown
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