Corrosion Behaviors Of Candidate Materials For Supercritical Water Reactor

The supercritical water cooled 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 (LWR). The supercritical water is extremely corrosive to the metal materials. Zr alloy which is widely used as the fuel cladding material in the current LWR is no longer applicable for the SCWR due to the high corrosion rate and loss of strength. Evaluation of several candidate materials for the SCWR has been conducted on the major fire tube materials used in supercritical fossil fired plant, structural materials used in current nuclear reactor and materials for aeroengine blades. For the application of a structural material to a fuel cladding in the SCWR, the material should be evaluated in terms of its corrosion and stress corrosion cracking susceptibility.General corrosion behaviors of austenitic stainless steel 800H, HR3C,316Ti and Ni-base alloy 718 were investigated in the supercritical water at 650℃/25MPa for 3000h in this paper. The experimental results showed that the weight gain of HR3C,718,800H is less than 50mg/dm2. Effects of surface treatment on corrosion resistance of 9Cr,12Cr and modified 12Cr ferritic/martensitic (F/M) steel were evaluated in supercritical water at 550℃/25MPa for 1000h. The results showed that the QPQ complex salt bath treatment could not improve the corrosion resistance of F/M steel in the supercritical water environment while Cr coatings prepared by both electro-plating and the magnetron sputtering can greatly reduce the corrosion weight gain rate. The weight gains of modified 12Cr F/M steel treated by Cr electro-plating or magnetron sputtering, and the 9Cr F/M steel treated by magnetron sputtering are less than 50mg/dm2.Slow stress rate tests (SSRT) were used in this paper to investigate the stress corrosion cracking (SCC) behaviors of austenitic stainless steel AL-6XN (both unirradiated and irradiated),316Ti, HR3C, TP347HFG and Ni-base alloy 718 in the supercritical water at temperature of 500, 600 and 650℃and at pressure of 25MPa, with the strain rate of 1×10-6s-1. AL-6XN showed the best ductility among the five candidate materials. The dynamic strain aging (DSA) phenomenon related to the interaction of solute atoms with dislocations was observed in the stress-strain curve of AL-6XN at 550℃. As the test temperature increased from 550℃to 650℃, the yield strength and tensile strength of 316Ti decreased and the elongation increased while both the strength and elongation of HR3C and TP347HFG decreased. Ni-base alloy 718 showed the highest yield strength and tensile strength but the lowest elongation. Intergranular cracks were observed on the fractured surface of the AL-6XN (650℃), HR3C (550℃), TP347HFG (550℃) and 718 (both 550℃and 650℃)in the tests.