Corrosion and Cracking of Short Listed Cladding Materials for Supercritical Water-Cooled Reactors

Abstract

de. Supercritical water-cooled nuclear reactor (SCWR) concept has been selected by Generation IV Form (GIF), but scientists and designers are facing unprecedented difficulties in solving the problems in material, neutronics and thermo-hydraulics. The high temperature and pressure operation condition of a SCWR core is tough and challenging to the materials for fuel cladding, pressure vessel and reactor internals. After 20 years of testing, the list of candidate materials for SCWR become shorter with availability of testing data. Among them, it has been clear that Cr and Al concentration is crucial for corrosion resistance, surface finish affects the diffusion of Cr and therefore affects the corrosion rate significantly. And we also show that creep crack growth is the major factor to cracking of austenitic alloys, such as 310, 800H and 690. People has tried to innovate new alloys based on the known factors, using ODS to improve the creep resistance, and adding more Cr and Al element to austenitic alloys to make AFA (alumina forming austenitic) alloys. The development of alloys for accident tolerant fuel (ATF) in the past 10 years casts a light on the selection and development of novel materials for SCWRs. 20 years passed after the establishment of Generation IV International Forum (GIF), we are still in the shortage of reliable environmental reliability data of materials for SCWR application, especially the corrosion rate and environmental cracking performances of short- listed materials for the nuclear fuel cladding. The currently available data showed high degree of inconsistency. The reactor engineering materials group at SJTU has been deeply involved in R&D of SCWR for 15 years, and the present paper presents the test data, mechanism and proposed testing procedures to obtain consistent and reproducible environmental reliability data for the engineering design of a SCWR.