Corrosion and stress corrosion cracking in supercritical water G.S. Was a, * , P. Ampornrat a , G. Gupta a , S. Teysseyre a , E.A. West a , T.R. Allen b , K. Sridharan b , L. Tan b , Y. Chen b , X. Ren b , C. Pister b a University of Michigan, Nuclear Engineering, 1911 Cooley, Ann Arbor, MI 48109, United States b University of Wisconsin, United States Abstract Supercritical water (SCW) has attracted increasing attention since SCW boiler power plants were implemented to increase the eﬃciency of fossil-based power plants. The SCW reactor (SCWR) design has been selected as one of the Generation IV reactor concepts because of its higher thermal eﬃciency and plant simpliﬁcation as compared to current light water reactors (LWRs). Reactor operating conditions call for a core coolant temperature between 280 °C and 620 °C at a pressure of 25 MPa and maximum expected neutron damage levels to any replaceable or permanent core com- ponent of 15 dpa (thermal reactor design) and 100 dpa (fast reactor design). Irradiation-induced changes in microstructure (swelling, radiation-induced segregation (RIS), hardening, phase stability) and mechanical properties (strength, thermal and irradiation-induced creep, fatigue) are also major concerns. Throughout the core, corrosion, stress corrosion cracking, and the eﬀect of irradiation on these degradation modes are critical issues. This paper reviews the current understanding of the response of candidate materials for SCWR systems, focusing on the corrosion and stress corrosion cracking response, and highlights the design trade-oﬀs associated with certain alloy systems. Ferritic–martensitic steels generally have the best resistance to stress corrosion cracking, but suﬀer from the worst oxidation. Austenitic stainless steels and Ni-base alloys have better oxidation resistance but are more susceptible to stress corrosion cracking. The promise of grain boundary engineering and surface modiﬁcation in addressing corrosion and stress corrosion cracking performance is discussed. Ó 2007 Elsevier B.V. All rights reserved. 1. Introduction One of the most promising advanced reactor concepts for Generation IV nuclear reactors is the Supercritical Water Reactor (SCWR). Operating above the thermodynamic critical point of water (374 °C, 22.1 MPa), the SCWR oﬀers many advan- tages compared to state-of-the-art LWRs including the use of a single phase coolant with high enthalpy, the elimination of components such as steam gener- ators and steam separators and dryers, a low coolant mass inventory resulting in smaller compo- nents, and a much higher eﬃciency (45% vs. 33% in current LWRs). Overall, the design provides a simpliﬁed, reduced volume system with high thermal eﬃciency. The challenge is provided by the substantial increase in operating temperature and pressure as compared to current BWR and PWR designs. The reference design for the SCWR [1,2] calls for an operating pressure of 25 MPa and an outlet water temperature up to 620 °C, Fig. 1. Since supercritical water has never been used in nuclear power applications, there are numerous 0022-3115/$ - see front matter Ó 2007 Elsevier B.V. All rights reserved. doi:10.1016/j.jnucmat.2007.05.017 * Corresponding author. Fax: +1 734 763 4540. E-mail address: gsw@engin.umich.edu (G.S. Was). Journal of Nuclear Materials 371 (2007) 176–201 www.elsevier.com/locate/jnucmat