Today, countries all over the world, faced with a global energy crisis and the effects of climate change, are looking for alternatives to fossil fuels [...]

Five Euratom projects launched since 2017 in support of the development of ESNII/Generation-IV reactor systems are briefly presented in the paper in terms of key objectives, results, and recommendations for the future. These projects focus on various aspects of the following ESNII/Generation-IV systems: Sodium Fast Reactor, Gas Cooled Fast Reactor, Supercritical Water Cooled Reactor, and Molten Salt Fast Reactor. The paper does not consider EU projects focused on the Gen-IV reactor technologies based on the use of heavy metals as a coolant because these projects are reviewed in a different paper.

Titanium stabilized stainless steel 08Ch18N10T is widely used in VVER nuclear power plants. The alloy 800H is a very promising structural material for application in future nuclear power plants working under more severe working conditions. Thus, both materials were exposed above the critical point of water at 395 °C, 25 MPa for 500 and 2700h. The effect of such exposures was evaluated based on analyses of microstructure via scanning electron microstructure involving chemical and crystallographic analyses. The thin oxide layers were further described by X-ray diffraction. The stainless steel 08Ch18N10T proved very good and comparable corrosion resistance with highly alloyed 800H at the given environment. In both cases, the oxide layer was very thin even after long exposure.

The main objectives of the project are to define the design requirements for the future SCW-SMR technology, to develop the pre-licensing study and guidelines for the demonstration of the safety in the further development stages of the SCW-SMR concept including the methodologies and tools to be used and to identify the key obstacles for the future SMR licencing and propose strategy for this process. The project consortium consists of laboratories from Europe, Canada and China. Beside the thermo-hydraulics, safety, neutronics, and reactor physics, assessment of the corrosion behaviour of cladding candidate materials (alloy 800H, stainless steel 310S, alumina forming austenitic alloy -AFA) in supercritical water (SCW) at 380/500 °C and 25 MPa is one of the key activities. For this purpose, experiments such as long-term (1000-8000 hours) exposure tests in autoclaves, neutron irradiation, electrochemistry, and radiolysis are involved. The paper describes the progress made on the ECC-SMART project with focus on microstructural characterization of candidate cladding materials. In addition, the description of the setup of corresponding experiments is included.