Corrosion Tests of Candidate Fuel Cladding and Reactor Internal Structural Materials

Abstract

Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor (SCWR) in static and flowing supercritical water (SCW) autoclave at the temperatures of 550, 600 and 650°C, pressure of about 25MPa, deaerated or saturated dissolved hydrogen (STP). Samples are nickel base alloy type Hastelloy C276, austenitic stainless steels type 304NG and AL-6XN, ferritic/martensitic (F/M) steel type P92, and oxide dispersion strengthened steel MA 956. This paper focuses on the formation and breakdown of corrosion oxide scales, and proposes the future trend for the development of SCWR fuel cladding materials.