Metal Fuel–Cladding Interaction

Abstract

While developing advanced nuclear energy systems, one needs to carefully consider the performance of the fuel in a particular reactor. If a metallic fuel is used in a sodium fast reactor (SFR), many phenomena can affect the performance of the fuel, including the one that this article discusses, namely, fuel–cladding chemical interaction (FCCI). During irradiation, a metallic fuel swells and eventually contacts the cladding, at which time FCCI can occur. During this process, interdiffusion occurs between the fuel, cladding, and fission products, which can result in the formation of interaction zones on the inner surface of the cladding that can become brittle or may contain relatively low-melting phases. The result of this process can be cracking and failure of the cladding. Detailed information on FCCI in irradiated metallic SFR fuel is minimal in the literature. The metallic fuels with the most data include U–Zr and U–Pu–Zr alloys. The bulk of the irradiation data that is available was generated from the destructive examination of individual fuel elements irradiated in the Experimental Breeder Reactor-II over the course of a 30-year time frame. Three examination techniques were employed to characterize FCCI in these fuel elements: optical metallography, electron probe microanalysis, and scanning electron microscopy. The results of these examinations have led to a reasonable understanding of FCCI and how it affects fuel performance. The fuel elements that were characterized included U–Zr or U–Pu–Zr alloy fuels and HT-9, D9, or Type 316 stainless steel cladding. Out-of-pile testing has also been conducted using annealed diffusion couples comprised of alloy and cladding constituents to improve the understanding of fuel–cladding interactions. Furthermore, to investigate liquefaction in FCCI zones, actual irradiated fuel segments or complete fuel pins have been heated to high temperatures and then characterized to ascertain the development of liquid phases at the fuel–cladding interface. Finally, transient tests have been performed on irradiated fuel pins to investigate microstructural development at the fuel–cladding interface. This article describes the data that is available with respect to out-of-pile and in-pile testing that can be employed to develop an understanding of FCCI in irradiated metallic nuclear fuels.