Interactions of Zircaloy Cladding with Gallium: Final Report
The U.S. Department of Energy has established a dual-track approach to the disposition of plutonium arising from the dismantling of nuclear weapons. Both immobilization and reactor-based mixed-oxide (MOX) fuel technologies are being evaluated. The reactor-based MOX fuel option requires assessment of the potential impact of concentrations of gallium (on the order of 1 to 10 ppm), not present in conventional MOX fhel, on cladding material performance. Three previous repmts"3 identified several compatibility issues relating to the presence of gallium in MOX fuel and its possible reaction with fiel cladding. Gallium initially present in weapons-grade (WG) plutonium is largely removed during processing to produce MOX fhel. After blending the plutonium with uranium, only 1 to 10 ppm gallium is expected in the sintered MOX fuel. Gallium present as gallium oxide (G~OJ could be evolved as the suboxide (G~O). Migration of the evolved G~O and diffusion of gallium in the MOX matrix along thermal gradients could lead to locally higher concentrations of G~03. Thus, while an extremely low concentration of gallium in MOX fiel almost ensures a lack of significant interaction of gallium whh Zircaloy fhel cladding, there remains a small probability that corrosion effects will not be negligible. General corrosion in the form of surface alloying resulting from formation of intermetallic compounds between Zircaloy and gallium should be ma& limited and, therefore, superficial because of the expected low ratio of gallium to the surface area or volume of the Zircaloy cladding. Although the expected concentration of gallium is low and there is very limited volubility of gallium in zirconium, especially at temperatures below 700 "C,4 grain boundary penetration and liquid metal embrittlement (LME) are forms of localized corrosion that were also considered. One fuel system darnage mechanism, pellet clad interaction, has led to some failure of the Zircaloy cladding in light-water reactors (LWRS). This has been attributed to stresses in the cladding and one or more aggressive fission products. Stress corrosion cracking by iodines' 6 and LME by cadmium7>8 have been reported, and it is known that Zircaloy can be embrittled by some low-melting metals, (e.g., mercury).g LME is a form of environmentally induced embrittlement that can induce cracking or loss of ductility. LME requties wetting and a tensile stress, but it does not require corrosion penetration. Experimentally, it has been demonstrated that gallium can cause embrittlement of some alloys (e.g., aluminum) at low temperatures,'"' ] ] but experiments relative to LME of zirconium by gallium have been limited and inconclusive.*2 This report describes a series of tests designed to establish the effects of low levels of residual gallium in WG-MOX fhel on its compatibility with Zircaloy. In addition, to establish damage mechanisms it was important to understand types of cladding interactions and available stiety margins with respect to gallium concentration.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Oak Ridge, TN
- Sponsoring Organization:
- USDOE Office of Nuclear Energy
- DOE Contract Number:
- AC05-96OR22464
- OSTI ID:
- 1303
- Report Number(s):
- ORNL/TM-13684; GA 01 02 01 4; ON: DE00001303
- Country of Publication:
- United States
- Language:
- English
Similar Records
Interactions of zircaloy cladding with gallium -- 1997 status
Interactions of Zircaloy cladding with gallium: 1998 midyear status