Specimen geometries have been developed to determine the mechanical properties of irradiated Zircaloy cladding subjected to the mechanical conditions and temperatures associated with reactivity-initiated accidents (RIA) and loss-of-coolant accidents (LOCA). Miniature ring-stretch specimens were designed to induce both uniaxial and plane-strain states of stress in the transverse (hoop) direction of the cladding. Also, longitudinal tube specimens were also designed to determine the constitutive properties in the axial direction. Finite-element analysis (FEA) and experimental parameters and results were closely coupled to optimize an accurate determination of the stress-strain response and to induce fracture behavior representative of accident conditions. To determine the constitutive properties, a procedure was utilized to transform measured values of load and displacement to a stress-strain response under complex loading states. Additionally, methods have been developed to measure true plastic strains in the gauge section and the initiation of failure using real-time data analysis software. Strain rates and heating conditions have been selected based on their relevance to the mechanical response and temperatures of the cladding during the accidents.

Zirconium hydrides precipitate in fuel cladding alloys as a result of hydrogen uptake from the high-temperature corrosion environment of light water reactors. Synchrotron X-ray diffraction was performed at room temperature on stress-relieved Zircaloy-4 cladding with two distributions of hydrides – (1) uniformly distributed hydrides across the entire cladding wall and (2) hydride rim next to the outer surface. The δ-hydride phase was found to be the predominant hydride phase to precipitate for hydrogen contents up to 1250 weight parts per million (wtppm). At a higher content, about 3000wtppm, although δ-hydride is still the majority phase, a significant amount of γ-hydride is also observed. At even higher hydrogen contents, in excess of approximately 6000wtppm, such as can occur in a highly dense hydride layer, peaks associated with the ε-hydride phase are also observed in the diffraction pattern. The volume fraction of hydrides was estimated as a function of hydrogen content using the integrated intensities of select diffraction peaks corresponding to the α-Zr matrix and the hydride phases. These estimated values agree well with calculated values from the independently measured concentrations. The results of this study indicate that hydride precipitation in Zircaloy-4 is a complex process of evolving hydride phases with increasing local hydrogen content.

The fracture behavior Of unirradiated Zircaloy-4 containing either solid hydride blisters or hydrided rims has been examined for the contrasting conditions of equal-biaxial and plane-strain tensile deformation at three temperatures (25 degrees, 300 degrees, and 375 degrees C). Cold-worked and stress-relieved Zircaloy-4 sheet containing hydride blisters shows nearly identical failure strains in equal-biaxial and plane-strain tensile deformation for a wide range of blister or rim depths. In all cases, failure strains decrease rapidly with increasing hydride blister or rim thickness, especially in the <= 100 mu m range. Test temperature has a significant effect on ductility with failure strains at 300 degrees and 375 degrees C being much greater than at room temperature. The results indicate that the ductility of material containing hydride rims/blisters greater than approximate to 30-40 mu m deep is limited by crack growth, which occurs in a mode 1 manner at 25 degrees C but in a mixed mode I/II manner at >= 300 degrees C (and at higher failure Strain levels).

The overall objective of this 3 year research program has been to support the aging management programs for LWR reactors through the extended 60-year lifetimes by adding to the knowledge base of irradiation-induced effects on the mechanical properties and cracking resistance of stainless steel (SS) core components. Over the course of this project, efforts have focused on enhancing the understanding of the link between deformation and fracture behavior in work-hardened and irradiated stainless steels. This understanding is achieved through a combination of mechanical testing, microstructural characterization, and development of models to describe the observed behavior. The understanding gained provides a foundation for evaluating the aging behavior of components during in-core service in terms of radiation embrittlement and alloy susceptibility to irradiation assisted stress corrosion cracking (IASCC).

During operation of nuclear power reactors, irradiated Zircaloy-4 cladding tubes contain circumferentially oriented hydrides concentrated in a layer near the outer surface of the cladding. This study has investigated the effect of a hydride layer or “rim” located near the outer surface of the cladding tube on the failure of unirradiated Zircaloy-4 cladding tubes. Utilizing plane-strain ring-stretch tests with the maximum principal stress along the circumferential or hoop direction, we examined the influence of a hydride rim on the failure of unirradiated Zircaloy-4 cladding at both room temperature and 300°C. Fracture is found to be sensitive to hydride-rim thickness such that cladding tubes with a hydride-rim thickness >140 μm (≈700 wppm total hydrogen) exhibit brittle behavior, while cladding tubes with a rim thickness 95 μm and >110 μm under uniaxial and biaxial hoop tension, respectively. These results suggest that a ductile-to-brittle transition occurs in the deformation behavior of cladding tubes containing a a finite hydride rim thickness. Additionally, previous studies [18,19] have investigated the effect of a hydride rim on the failure path of Zircaloy-4 cladding tubes irradiated to high fuel burnups and subjected to RIAsimulation tests. Although not quantitative, according to post-test examinations, Garde and coworkers [18] and Fuketa and co-workers [19] suggest that fracture initiation occurs in the hydride rim by brittle crack growth and, depending on temperature and loading path, followed by either ductile or brittle fracture of the remaining cladding ligament. The purpose of this study is to explore the response of unirradiated Zircaloy-4 cladding tubes that contain hydrides concentrated in the form of a thin layer near the outer surface, as is typical of high-burnup cladding [19]. Using “ring-stretch” specimen geometries in order to impose multi-axial stresses and near plane-strain tension in the hoop direction of the cladding tube, we examine the influence of thin layers of hydrides on the ductility of Zircaloy-4 cladding subject to stress states relevant to potential in-service accidents such as the RIA. Experimental Procedures Material As in previous studies [12,15,17], Zircaloy-4 cladding tubes were obtained from Sandvik Metals and Westinghouse Electric Corporation in a CWSR condition with outer diameter of