Fuel-Cladding Interaction Layers in Irradiated U-Zr and U-Pu Zr Fuel Elements

摘要

Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr and U-Pu-Zr alloy fuel elements irradiated in the Experimental Breeder Reactor-II (EBR-II). The electrometallurgical treatment process extracts usable uranium from irradiated fuel elements and places residual fission products, actinides, process Zr, and cladding hulls (small segments of tubing) into two waste forms--a ceramic and a metal alloy. The metal waste form will contain the cladding hulls, Zr, and noble metal fission products, and it will be disposed of in a geologic repository. As a result, the expected composition of the waste form will need to be well understood. This report deals with the condition of the cladding, which will make up a large fraction of the metal waste form, after irradiation in EBR-II and before insertion into the electrorefiner. Specifically, it looks at layers that can be found on the inner surface of the cladding due to in-reactor interactions between the alloy fuel and the stainless steel cladding that occurs after the fuel has swelled and contacted the cladding. Many detailed examinations of fuel elements irradiated in EBR-II have been completed and are discussed in the context of interaction layer formation in irradiated cladding. The composition and thickness of the developed interaction layers are identified, along with the irradiation conditions, cladding type, and axial location on fuel elements where the thickest interaction layers can be expected to develop. It has been found that the largest interaction zones are observed at combined high power and high temperature regions of fuel elements and for fuel elements with U-Pu-Zr alloy fuel and D9 stainless steel cladding. The most prevalent, non-cladding constituent observed in the developed interaction layers are the lanthanide fission products.