Oxidation Behavior of an Austenitic Stainless Steel Used in the UK Advanced Gas Cooled Reactors

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Product Number: 51315-5965-SG

ISBN: 5965 2015 CP

Author: Bo Chen

Publication Date: 2015

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The UK’s advanced gas cooled (AGR) nuclear reactors operate over the last 30 years primarily at temperatures ranging from 470 °C up to 650 °C. The coolant gases used in this type of nuclear system contain a mixture of carbon dioxide carbon monoxide hydrogen methane and water vapour. A high incidence of cracking has been reported in several 4 mm thick austenitic stainless steel tubes since 2006. The R5 high temperature structural integrity assessment implied that this type of cracking is unlikely to have been caused by a creep-fatigue mechanism. Following the comprehensive metallographic examinations of the cracked body of tubes it has been generally accepted that the mechanism underlying the initiation of cracks has been identified to be a consequence of carburisation associated with the presence of duplex oxide layer. In this paper several complementary microstructural evaluation techniques have been used to investigate the oxidation behaviour of Type 316H stainless steel in the simulated AGR environments. A primary focus was given to the effects of surface finish and the water vapour content on oxidation. The experimental results show that surface deformation promotes the formation of a thin oxide layer whereas a deformation-free surface leads to a formation of thick duplex oxide layers. Furthermore the presence of water vapour in the mixed gas environment accelerated the growth of the oxides. These results are discussed with respect to the diffusivities of solute elements as well as the roles of oxygen partial pressure and the water content on the oxidation driving force.