This paper focuses on the characterization of an advanced steel which has been developed for use as a structural material within future nuclear fusion reactors, including in irradiated water coolant-facing locations. In the paper an experimental plan is described which would allow both the corrosion and stress corrosion cracking susceptibility of irradiated Eurofer-97 to be studied. Also included are early results from characterization of self-ion irradiated (using Fe ions) Eurofer-97 following high temperature corrosion experiments using electron microscopy techniques. Field emission gun scanning electron microscopy (FEG SEM) images of the surfaces of unirradiated Eurofer-97 are compared to conventional unirradiated stainless steel types following a high temperature corrosion experiment. In a different, longer, high temperature corrosion experiment using ion-irradiated Eurofer-97, a focused ion beam scanning electron microscope (FIB SEM) was used to prepare lamella for observation by high magnification transmission electron microscopy (TEM) and elemental analysis.
Nuclear energy currently contributes approximately 10 % of the worldwide energy mix.1 Nuclear energy generation is a form of low-carbon electricity, typically run as base-load, which alongside renewables can help nations toward climate change goals. Nuclear fission thermal reactors make up the majority of the reactors operating today. Nuclear fusion on the other hand is a promising alternative which produces less radioactive waste and does not have a reliance on the finite source of uranium fuel. Eurofer-97, a reduced activation ferritic-martensitic (RAFM) steel, will be used as a structural material for fusion reactors. The earliest literature reference to RAFM steels originated from 1994 by Abe et al.2 One option for the European demonstration fusion reactor (DEMO) is to use a water-cooled lead-lithium (PbLi) breeder blanket (WCLL BB) design for heat extraction. Breeder blankets will be used to generate a source of tritium, for the fusion reaction with deuterium. Such a BB design will expose Eurofer-97 to high temperature water on the inside of cooling channels, and liquid PbLi on the outside.3 Several authors have already published details on the corrosion of Eurofer-97 in contact with the PbLi liquid metal eutectic,4–7 though comparably little published literature is available for Eurofer-97 in contact with high temperature water. Although limited work has been published on the corrosion of Eurofer-97 in water, a small body of work has been conducted at lower temperatures (60 °C) by Martin et al.8 to consider the fusion-specific impact of a magnetic field. These researchers found indications that the presence of a magnetic field could affect corrosion in aggressive solution chemistries, which is notable as toroidal-type fusion reactors control the hot plasma by means of magnetic confinement. Other researchers have published details on the WCLL BB water chemistry, temperature and pressure working specification, which leverages fission operating experience (pressurized water reactor (PWR) water temperatures (295 °C inlet to 328 °C outlet) and 15.5 MPa pressure).3 Along with the current lack of corrosion experiments being reported in the open literature for Eurofer-97 in high temperature water, there is also a corresponding lack of stress corrosion cracking (SCC) experiments for the alloy detailed in the open literature. An exception to this is the work by Dyck and Bosch9 who studied neutron irradiated Eurofer-97 in high-temperature water using slow strain rate testing (SSRT). In their work they found no instances of SCC when the water chemistry was well controlled.
