An Assessment of Layer Development at the Fuel/Cladding Interface During Irradiation of Metallic SFR Fuel Elements
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An Assessment of Layer Development at the Fuel/Cladding Interface During Irradiation of Metallic SFR Fuel Elements Dennis D. Keiser, Jr. and James I. Cole Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83402-1625, U.S.A. ABSTRACT To investigate fuel cladding chemical interaction in irradiated metallic nuclear fuels, diffusion couple experiments have been performed using prototypic metallic fuel alloys with additions of noble metal and lanthanide fission product components mated against stainless steel claddings. The developed interdiffusion zones have been characterized using SEM/EDS/WDS to determine the development of phases and the interdiffusion behavior of specific fuel, cladding, and fission product components. The formed diffusion structures have been compared to actual interaction zones that form in irradiated metallic SFR fuels. This paper discusses how the structures compare between the diffusion couple-generated interdiffusion zones and those that develop in irradiated metallic nuclear fuels. It was found that similarities exist between the phase development and interdiffusion behavior in the annealed diffusion couples and the irradiated fuels. Nd, Mo, and Ru, which were added to a fuel alloy to represent fission products that are present in irradiated metallic nuclear fuels, were found to exhibit interdiffusion behavior in annealed diffusion couples that was similar to what has been observed in actual irradiated metallic fuels. This was also true for the original fuel components U, Pu, and Zr, along with the cladding constituent Fe, Ni, and Cr. INTRODUCTION During irradiation of a metallic nuclear fuel element, the fuel swells and eventually contacts the cladding. Interactions between the fuel and cladding may then occur resulting in the development of interaction zones that can impact the overall irradiation performance of a fuel pin [1,2]. Formed interaction zones can be potentially brittle or can contain phases that may be relatively low melting. To improve the understanding of fuel/cladding chemical interactions (FCCI), out-of-pile interdiffusion experiments have been performed using prototypic metallic fuel alloys and stainless steel claddings [3,4]. Of particular interest when performing these types of experiments is how they compare to what is actually observed in-reactor. This paper will compare the results of some recent interdiffusion experiments with what is observed in irradiated fuel. EXPERIMENTAL The diffusion couple experiments discussed in this paper employed a prototypic fuel alloy that imitated high-burnup fuel. The alloy was cast using an arc-melting process. The feedstock materials were first melted together in highly purified argon gas. The alloys were then
inverted and remelted. This process was repeated three times to enhance homogeneity. The ascast alloy was then annealed at 850˚C for 48 hours to further homogenize the alloy. The alloy was comprised of U, Pu, Zr, Nd, Mo, and Ru. The composition of this alloy was (in wt%): 65U-19Pu-9Zr-2Nd-2.5Mo-2.5Ru, whi
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